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Sample records for austenitic alloys irradiated

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  4. Induced effects in Fe-Ni-Cr austenitic alloys by electron irradiation

    International Nuclear Information System (INIS)

    Materials behaviour under high energetic particles exposure has to be know for technological aspects, but also for microscopic material state physics. Large macroscopic investigations have been developed but reliability with theoretical calculations or fundamental physics measurements is not clear. We present four experimental procedures in order to characterize austenitic Fe-Ni-Cr synthetic alloys in the atomic scale. First, results obtained about vacancy and interstitial, after electrical resistivity measurements and monoenergetical or classical positron annihilation process, are discussed. Then, defects clustering and microstructural evolution is investigated using positron lifetime measurements and high resolution electronic microscopy. In this study, special care has been taken to understand the composition effect as a function of the irradiation conditions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  6. Effect of Mn addition on decrease of Cr depletion at grain boundary in austenitic alloys irradiated with electrons

    International Nuclear Information System (INIS)

    Radiation-induced Cr depletion at a grain boundary (GB) is known as one of the major factors to degrade corrosion resistance of austenitic stainless steel. The effect of 10% Mn addition on prevention of the Cr depletion was investigated from a viewpoint of volume size factor (VSF) of Cr in the austenitic alloys irradiated with 1 MeV electrons. VSF of Cr in solution-annealed 316L steel added with 10 wt% Mn was +0.8%, decreased by 4% compared with 316L. Radiation-induced Cr depletion at GB of 316L+10%Mn was smaller than that of 316L at 723 and 773 K. Decrease of radiation-induced Cr depletion in 316LF+10%Mn is thought to be derived mainly from the suppression of vacancy-Cr atom interaction. (orig.)

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

    Science.gov (United States)

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

    2003-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-16

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    NARCIS (Netherlands)

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

    2003-01-01

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

  15. The influence of pre-irradiation heat treatments on thermal non-equilibrium and radiation-induced segregation behavior in model austenitic stainless steel alloys

    International Nuclear Information System (INIS)

    The effect of pre-irradiation heat treatments on thermal non-equilibrium grain boundary segregation (TNES) and subsequent radiation-induced grain boundary segregation (RIS) is studied in a series of model austenitic stainless steels. The alloys used for this study are based on AISI 316 stainless steel and have the following nominal compositions: Fe-16Cr-13Ni-1.25Mn (base 316), Fe-16Cr-13Ni-1.25Mn-2.0Mo (316+ Mo) and Fe-16Cr-13Ni-1.25Mn-2.0Mo-0.07P (316+ Mo+ P). Samples were heat treated at temperatures ranging from 1100 to 1300 C and cooled at 4 different rates (salt brine quench, water quench, air cool and furnace cool) to evaluate the effect of annealing temperature and quench rate on TNES. The alloys were than processed with the treatment (temperature and cooling rate) that resulted in the maximum Cr enrichment. Alloys with and without the heat treatment to enrich the grain boundaries with Cr were characterized following irradiation to 1 dpa at 400 C with high-energy protons in order to understand the influence of alloying additions and pre-irradiation grain boundary chemistry on irradiation-induced elemental enrichment and depletion profiles. Various mechanistic models will be examined to explain the observed behavior

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  17. Investigation of joining techniques for advanced austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Influence of phosphorus on point defects in an austenitic alloy

    International Nuclear Information System (INIS)

    The influence of phosphorus on points defects clusters has been studied in an austenitic alloy (Fe/19% at. Cr/13% at. Ni). Clusters are observed by transmission electron microscopy. After quenching and annealing, five types of clusters produced by vacancies or phosphorus-vacancies complexes are observed whose presence depends on cooling-speed. Vacancy concentration (with 3.6 10-3 at. P) in clusters is about 10-5 and apparent vacancy migration is 2± 0.1 eV. These observations suggest the formation of metastable small clusters during cooling which dissociate during annealing and migrate to create the observed clusters. With phosphorus, the unfrequent formation of vacancy loops has been observed during electron irradiation. Ions irradiations show that phosphorus does not favour nucleation of interstitial loops but slowers their growth. It reduces swelling by decreasing voids diameter. Phosphorus forms vacancy complexes whose role is to increase the recombination rate and to slow vacancy migration

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

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

    Science.gov (United States)

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

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-21

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  10. Study of irradiation damage structures in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, Shozo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-08-01

    The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

  14. Manifestations of DSA in austenitic stainless steels and inconel alloys

    International Nuclear Information System (INIS)

    The aim of the investigation was to examine and compare different types of DSA (Dynamic Strain Aging) manifestations in AISI 316 austenitic stainless steel (SS) and Inconel 600 and Inconel 690 alloys by means of slow strain rate tensile testing, mechanical loss spectrometry (internal friction) and transmission electron microscopy (TEM). Another aim was to determine differences in the resulting dislocation structures and internal friction response of materials showing and not showing DSA behaviour

  15. Alkaline stress corrosion of iron-nickel-chromium austenitic alloys

    International Nuclear Information System (INIS)

    This research thesis reports the study of the behaviour in stress corrosion of austenitic iron-nickel-chromium alloys by means of tensile tests at imposed strain rate, in a soda solution at 50 pc in water and 350 degrees C. The author shows that the mechanical-chemical model allows the experimental curves to be found again, provided the adjustment of characteristic parameters, on the one hand, of corrosion kinetics, and on the other hand, of deformation kinetics. A classification of the studied alloys is proposed

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

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.

    1996-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Microstructural evolution in solution-annealed Japanese-PCA (JPCA-SA) and four other austenitic stainless steels, irradiated at 400 deg. 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 1.0x1022 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 deg. C

  1. High temperature irradiation creep in austenitic steels

    International Nuclear Information System (INIS)

    An analysis has been made of the in-reactor and ex-reactor creep at 700 - 7500C of various Ti - and Nb - stabilised steels. Above a critical transition stress that depended on steel composition and thermomechanical treatment, the stress dependence of the creep rate was high but there was little influence of irradiation on the kinetics. At lower stresses the stress dependence was small and the creep rate varied as the inverse cube of the grain size. In-reactor creep rates were about ten times faster than those ex-reactor, the in-reactor rates approaching the magnitude of the Coble grain boundary diffusion creep process. A mechanism is proposed to explain the enhanced creep rates in-reactor based on the idea that SIPA irradiation creep of carbide particles occurs at the grain boundary vacancy sinks during diffusion creep. This limits the stress redistribution at the grain boundary and the generation of high stresses at the particles which, in the ex-reactor tests, can markedly inhibit the diffusion creep process. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Etienne, A., E-mail: auriane.etienne@etu.univ-rouen.f [Groupe de Physique des Materiaux, Universite et INSA de Rouen, UMR CNRS 6634, BP 12, 76 801 Saint Etienne du Rouvray Cedex (France); Hernandez-Mayoral, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Genevois, C.; Radiguet, B.; Pareige, P. [Groupe de Physique des Materiaux, Universite et INSA de Rouen, UMR CNRS 6634, BP 12, 76 801 Saint Etienne du Rouvray Cedex (France)

    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 deg. 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.

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

    Science.gov (United States)

    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.

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

    Directory of Open Access Journals (Sweden)

    Rouxel Baptiste

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  6. Microstructure and properties of laser surface alloyed PM austenitic stainless steel

    OpenAIRE

    Z. Brytan; M. Bonek; L.A. Dobrzański

    2010-01-01

    Purpose: The purpose of this paper is to analyse the effect of laser surface alloying with chromium on the microstructural changes and properties of vacuum sintered austenitic stainless steel type AISI 316L (EN 1.4404).Design/methodology/approach: Surface modification of AISI 316L sintered austenitic stainless steel was carried out by laser surface alloying with chromium powder using high power diode laser (HPDL). The influence of laser alloying conditions, both laser beam power (between 0.7 ...

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

    Institute of Scientific and Technical Information of China (English)

    Selva Büyükakkas; H Aktas; S Akturk

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, K.M.

    1979-05-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Jie HUANG; Zhou XU; Xin XING

    2003-01-01

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

  10. MODELING OF AUSTENITE GRAIN SIZE IN LOW-ALLOY STEEL WELD METAL

    Institute of Scientific and Technical Information of China (English)

    A.G.Huang; Y.S.Wang; Z.Y.Li; J.G.Xiong; Q.Hu

    2004-01-01

    The size of austenite grain hassignificant effects on components and proportions of various ferrites in low-alloy steel weld metal.Therefore,it is important to determine the size of austenite grain in the weld metal.In this paper,a model based upon the carbon diffusion rate is developed for computing austenite grain size in low-alloy steel weld metal during continuous cooling.The model takes into account the effects of the weld thermal cycles,inclusion particles and various alloy elements on the austenite grain growth.The calculating results agree reasonably with those reported experimental observations.The model demonstrates a significant promise to understand the weld microstructure and properties based on the welding science.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Field, Kevin G.; Busby, Jeremy T.

    2015-05-15

    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 (σ{sub 0.2}). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ{sub 0.2}, 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.

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

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

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

    Science.gov (United States)

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

    2016-07-01

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

  15. Irradiation induced surface segregation in concentrated alloys: a contribution

    International Nuclear Information System (INIS)

    A new computer modelization of irradiation induced surface segregation is presented together with some experimental determinations in binary and ternary alloys. The model we propose handles the alloy thermodynamics and kinetics at the same level of sophistication. Diffusion is described at the atomistic level and proceeds vis the jumps of point defects (vacancies, dumb-bell interstitials): the various jump frequencies depend on the local composition in a manner consistent with the thermodynamics of the alloy. For application to specific alloys, we have chosen the simplest statistical approximation: pair interactions in the Bragg Williams approximation. For a system which exhibits the thermodynamics and kinetics features of Ni-Cu alloys, the model generates the behaviour parameters (flux and temperature) and of alloy composition. Quantitative agreement with the published experimental results (two compositions, three temperatures) is obtained with a single set of parameters. Modelling austenitic steels used in nuclear industry requires taking into account the contribution of dumbbells to mass transport. The effects of this latter contribution are studied on a model of Ni-Fe. Interstitial trapping on dilute impurities is shown to delay or even suppress the irradiation induced segregation. Such an effect is indeed observed in the experiments we report on Fe50Ni50 and Fe49Ni50Hf1 alloys. (author)

  16. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-15

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

  17. The effects of fast-neutron irradiation on the mechanical properties of austenitic stainless steel

    International Nuclear Information System (INIS)

    The paper reviews the effects of fast-neutron irradiation on the tensile properties of austenitic stainless steels at irradiation temperatures of less than 400 degrees Celcius, using as an example, work carried out at Pelindaba on an AISI 316 type steel produced in South Africa. Damage produced in these steels at higher irradiation temperatures and fluences is also briefly discussed. The paper concludes with a discussion of some methods of overcoming or decreasing the effects of irradiation damage

  18. Microstructure and properties of laser surface alloyed PM austenitic stainless steel

    Directory of Open Access Journals (Sweden)

    Z. Brytan

    2010-05-01

    Full Text Available Purpose: The purpose of this paper is to analyse the effect of laser surface alloying with chromium on the microstructural changes and properties of vacuum sintered austenitic stainless steel type AISI 316L (EN 1.4404.Design/methodology/approach: Surface modification of AISI 316L sintered austenitic stainless steel was carried out by laser surface alloying with chromium powder using high power diode laser (HPDL. The influence of laser alloying conditions, both laser beam power (between 0.7 and 2.0 kW and powder feed rate (1.0-4.5 g/min at constant scanning rate of 0.5m/min on the width of alloyed surface layer, penetration depth, microstructure evaluated by LOM, SEM x-ray analysis, surface roughness and microhardness were presented.Findings: The microstructures of Cr laser alloyed surface consist of different zones, starting from the superficial zone rich in alloying powder particles embedded in the surface; these particles protrude from the surface and thus considerably increase the surface roughness. Next is alloyed zone enriched in alloying element where ferrite and austenite coexists. The following transient zone is located between properly alloyed material and the base metal and can be considered as a very narrow HAZ zone. The optimal microstructure homogeneity of Cr alloyed austenitic stainless steel was obtained for powder feed rate of 2.0 and 4.5 g/min and laser beam power of 1.4 kW and 2 kW.Practical implications: Laser surface alloying can be an efficient method of surface layer modification of sintered stainless steel and by this way the surface chromium enrichment can produce microstructural changes affecting mechanical properties.Originality/value: Application of high power diode laser can guarantee uniform heating of treated surface, thus uniform thermal cycle across treated area and uniform penetration depth of chromium alloyed surface layer.

  19. Measurement techniques of magnetic properties for evaluation of neutron irradiation damage on austenitic stainless steels

    International Nuclear Information System (INIS)

    The remote-controlled equipment for measurement of magnetic flux density has been developed in order to evaluate the irradiation damage of austenitic stainless steels. Magnetic flux densities by neutron irradiation in austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), have been measured by the equipment. The results show that irradiation damage affected to magnetic flux density, and indicate the measuring method of magnetic flux density using a small magnetizer with a permanent magnet of 2 mm in diameter is less affected by specimen shape. (author)

  20. low temperature irradiation effects in iron-alloys and ceramics

    International Nuclear Information System (INIS)

    Electron beam irradiation at 77K and neutron irradiation at 20K were carried out on Fe-Cr and Fe-Cr-Ni alloys and ZnO and graphite system ceramics, and by measuring positron annihilation lifetime, the micro-information about irradiation-introduced defects was obtained. The temperature of the movement of atomic vacancies in pure iron is about 200K, but it was clarified that by the addition of Cr, it was not much affected. However, in the case of high concentration Cr alloys, the number of atomic vacancies which take part in the formation of micro-voids decreased as compared with the case of pure iron. It is considered that among the irradiation defects of ZnO, O-vac. restored below 300degC. It is considered that in the samples without irradiation, the stage of restoration exists around 550degC, which copes with structural defects. By the measurement of graphite without irradiation, the positron annihilation lifetime corresponding with the interface of matrix and crystal grains, grain boundaries and internal surfaces was almost determined. The materials taken up most actively in the research and development of nuclear fusion reactor materials are austenitic and ferritic stainless steels, and their irradiation defects have been studied. (K.I.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    International Nuclear Information System (INIS)

    This report summarizes the research and development work performed during the second year of an M.I.T. project directed toward the development of improved structural alloys for the fusion reactor first wall application. Several new alloys have been produced by rapid solidification. Emphasis in alloy design and production has been placed on producing austenitic Type 316SS with fine dispersions of TiC and Al2O3 particles. Results of mechanical and microstructural tests are presented. A number of neutron irradiations have been initiated on samples fabricated from alloys produced in this project. A dual beam, heavy ion and helium ion, irradiation was completed using several alloys and a range of temperatures, damage rates and total doses. Modeling of irradiation phenomena has been continued with emphasis in the last year upon understanding the effect of recoil resolution on relatively stable second phase particles. Work continued to fully characterize the microstructure of several ZrB2 doped stainless steels

  3. Effects of milling process and alloying additions on oxide particle dispersion in austenitic stainless steel

    International Nuclear Information System (INIS)

    An oxide dispersion strengthened (ODS) austenitic stainless steel was developed by mechanical alloying (MA) of advanced SUS316 stainless steel. A nano-characterization was performed to understand details of the effect of minor alloying elements in the distribution of dispersoids. It is shown that Y2O3 particles dissolve into the austenitic matrix after the MA for 6 h. Annealing at 1073 K or higher temperatures result in a distribution of fine oxide particles in the recrystallized grains in the ODS austenitic stainless steel. Additions of Hafnium or Zirconium led to the distribution of finer oxide particles than in samples without these elements, resulting in an increase in the hardness of the samples. The most effective concentration of Hf and Zr to increase the hardness was 0.6 and 0.2–0.3 wt%, respectively

  4. Phase stability in an austenitic Fe-Cr-Mn (W,V) alloy

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    By means of deformation and long term aging, the stability and phase equilibrim characteristic of the C+N synthetically strengthening austenitic Fe-Cr-Mn (W,V) alloy were investigated. Experimental results indicate that the austenitic alloy remains stability and no →transformation occurs under 500℃. Synthetic addition of C and N causes the grains to refine and powerfully retards formation of martensite and precipitation of phase. Ms point is elevated with long term aging at elevated temperature (500-700℃) due to a large number ofstrain induced carbides precipitate. Along with accelerated decomposition of strain induced ' martensite and occurrence of recrystallization,γ →α transformation and phase precipitation are promoted so that austenite becomes unstable.

  5. Reducing heat tint effects on the corrosion resistance of austenitic stainless alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kearns, J.R. (Allegheny Ludlum Corp., Brackenridge, PA (United States)); Moller, G.E. (Allegheny Ludlum Corp., Evergreen, CO (United States))

    1994-05-01

    Arc welding can produce a heat tint on the surface of stainless and nickel-based alloys. In some services, a heat tint can decrease corrosion resistance. The conditions that cause heat tinting are discussed, and laboratory studies on post-weld cleaning procedures for removing this surface oxide scale from a 6% molybdenum super-austenitic alloy (UNS N08367) are reviewed. Cleaning can be done by either mechanical or chemical methods; a combination of both is recommended.

  6. Influence of substructure on mechanical properties of austenitic alloys deformed by warm rolling

    Energy Technology Data Exchange (ETDEWEB)

    Izotov, V.I.; Virakhovskij, Yu.G.; Marusenko, S.Ya. (Tsentral' nyj Nauchno-Issledovatel' skij Inst. Chernoj Metallurgii, Moscow (USSR). Inst. Metallovedeniya i Fiziki Metallov)

    1983-08-01

    A connection between a substructure and mechanical properties of some iron base austenitic alloys, differing in carbon, and carbide-forming element contents and in stacking fault energies after warm rolling, is studied. It is shown that the maximum value of yield strength after cold hardening is achieved in the alloy with low stacking fault energy due to the formation of high density of thin twins.

  7. Influence of substructure on mechanical properties of austenitic alloys deformed by warm rolling

    International Nuclear Information System (INIS)

    A connection between a substructure and mechanical properties of some iron base austenitic alloys, differing in carbon, and carbide-forming element contents and in stacking fault energies after warm rolling, is studied. It is shown that the maximum value of yield strength after cold hardening is achieved in the alloy with low stacking fault energy due to the formation of high density of thin twins

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

    Science.gov (United States)

    Dong, Dingqian; Chen, Fei; Cui, Zhenshan

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

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

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

    Science.gov (United States)

    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.

  12. Precipitation of K phase in austenitic alloys of Fe-Mn-Al system

    International Nuclear Information System (INIS)

    The kinetics of austenite decomposition in a fully austenitic Fe-Mn-Al-Si-C alloy aged for up to 400 hours at 500, 550, 600 and 6500C was investigated. Mettalographic studies using optical and scanning electron microscopy, microprobe analysis and X-ray diffraction showed the presence only of the K-phase in the aged samples. Ferrite and other phases such as β-Mn were not detected at the aging temperatures employed. The activation energy for the K phase precipitation was evaluated by means of the evaluation of hardness peaks associated to the early stages of precipitation. (author)

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

    Science.gov (United States)

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

    2016-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-22

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-31

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

  16. The importance of metallurgical variables in environment sensitive fracture of austenitic alloys

    International Nuclear Information System (INIS)

    The effects of metallurgical variables on environment sensitive cracking of austenitic Fe-Cr-Ni alloys, in particular austenitic stainless steels, have been examined. It is demonstrated by reviewing available literature data and by new, unpublished results that the nature and extent of susceptibility are sensitive such metallurgical variables as composition, grain size, microstructure, thermal treatment and radiation damage. Environment sensitive cracking has been classified as hydrogen-induced cracking or selective dissolution of an active path (Cr-depleted zone, segregations or deformation structures). The common factors between stress corrosion cracking and hydrogen embrittlement of these alloys are identified. Finally, possible aspects of the role and mechanism of hydrogen-induced cracking in environment sensitive cracking are discussed. (author)

  17. Laser surface melting of an austenitic Fe-26Mn-7Al-0.9C alloy

    International Nuclear Information System (INIS)

    A laser surface melting technique was used to modify and improve the surface properties of an austenitic Fe-26Mn-7Al-0.9C alloy. Scanning electron microscopy observations were made of the structural features of the laser melted zone and the substrate aged at 600 and 710 C respectively for different periods. Metallographic examination revealed that the laser melted region consisted of columnar and equiaxed dendrites. Aging treatment resulted in the development of ferrite and brittle β-Mn phases into large modules which grew into the initial austenitic grains of the substrate alloy. However, the laser melting resulted in an appreciable decrease in the fraction of β-Mn phase after aging treatment. (orig.)

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-25

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

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

    International Nuclear Information System (INIS)

    Highlights: • NiMnIn austenite and martensite have similar Ising-type critical exponents. • NiMnIn critical exponents rule out disordered states as spin-glass in martensite. • In NiMnIn alloys, magnetism arises mainly from moments localized at Mn atoms. • NiCoMnSn critical exponents are close to the ones from tricritical mean field model. • NiCoMnSn complex magnetic state results from three different magnetic atoms. - Abstract: Two different metamagnetic shape memory alloys of nominal composition Ni50Mn36In14 and Ni42Co8Mn39Sn11 have been studied by means of modified Arrott plots to give insight into the magnetic states of both the austenitic and martensitic phases. For Ni50Mn36In14 alloy, the same critical exponents (β = 0.32 and γ = 2.0) are obtained in austenite and martensite. They suggest that localized moments at Mn atoms are responsible for the magnetism of both phases according to the Ising model. The martensite, however, displays a rather complex behavior because β continuously changes with temperature. In Ni43Co6.5Mn39Sn11.5, critical exponents in the austenite are β = 0.27 and γ = 1.0. They are close to the tricritical mean field model, but no reliable fits were obtained in the martensite. The results are discussed in terms of microscopically different magnetic states in two alloys reflecting a complex interplay between the ferromagnetic and antiferromagnetic contributions

  1. The welding of austenitic-ferritic Mo-alloyed Cr-Ni-Steel

    International Nuclear Information System (INIS)

    This paper provides general information and guidance on the welding of austenitic-ferritic Mo-alloyed Cr-Ni stainless steel. Information is given on the various chemical compositions and on resistance to corrosion and on the mechanical and physical properties of commercially available steels. The effect of welding on the base metal and the selection of welding processes and welding consumables are described

  2. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    Science.gov (United States)

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

    2016-07-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

  5. Carburization of austenitic alloys by gaseous impurities in helium

    International Nuclear Information System (INIS)

    The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H2, CO, CH4, H2O and CO2 was studied. Corrosion tests were conducted in a temperature range from 649 to 10000C (1200 to 18320F) 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 μatm H2, 450 μatm CO, 50 μatm CH4 and 50 μatm H2O for Environment A; 200 μatm H2, 100 μatm CO, 20 μatm CH4, 50 μatm H2O and 5 μatm CO2 for Environment B; 500 μatm H2, 50 μatm CO, 50 μatm CH4 and 2O for Environment C; and 500 μatm H2, 50 μatm CO, 50 μatm CH4 and 1.5 μatm H2O 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

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

    Institute of Scientific and Technical Information of China (English)

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

    2015-01-01

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

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

    Science.gov (United States)

    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.

  8. Thermal property characterization of a titanium modified austenitic stainless steel (alloy D9)

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Aritra [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Raju, S. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)]. E-mail: sraju@igcar.ernet.in; Divakar, R. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Mohandas, E. [Physical Metallurgy Section, Materials Characterisation Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Panneerselvam, G. [Fuel Chemistry Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Antony, M.P. [Fuel Chemistry Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2005-12-01

    The temperature dependence of lattice parameter and enthalpy increment of alloy D9, a titanium modified nuclear grade austenitic stainless steel were studied using high temperature X-ray diffraction and inverse drop calorimetry techniques, respectively. A smooth variation of the lattice parameter of the austenite with temperature was found. The instantaneous and mean linear thermal expansion coefficients at 1350 K were estimated to be 2.12 x 10{sup -5} K{sup -1} and 1.72 x 10{sup -5} K{sup -1}, respectively. The measured enthalpy data were made use of in estimating heat capacity, entropy and Gibbs energy values. The estimated isobaric heat capacity C {sub p} at 298 K was found to be 406 J kg{sup -1} K{sup -1}. An integrated theoretical analysis of the thermal expansion and enthalpy data was performed to obtain approximate values of bulk modulus as a function of temperature.

  9. Microstructural Changes on Tensile Property of Austenitic Alloys Exposed to High Temperature Supercritical-CO{sub 2} Environment

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Several studies have been conducted on corrosion and mechanical properties of ferritic martensitic steels (FMSs) in liquid sodium coolant environments. As candidate materials for S-CO{sub 2} intermediate heat exchanger (IHX), corrosion study on tensile property for long-term integrity of austenitic alloys is in great demand. Therefore, in this study, corrosion behavior on tensile property of austenitic alloys after exposure to high temperature S-CO{sub 2} is presented. Microstructural changes are related to the changes in tensile property. The following conclusions can be drawn from this study of corrosion behavior on tensile property of austenitic alloys after exposure to high temperature S-CO{sub 2}: 1. Both Fe-base and Ni-base austenitic alloys showed a good corrosion resistance at 550 .deg. C, whereas at higher temperatures (over 600.deg.C) the corrosion characteristics of the Fe-base alloys were severely worsened compared to the Ni-base. 2. Changes in tensile property seemed to have no effects of base elements. Rather, SS 316H, Alloy 625 and 800HT - showed a reduced ductility at over 600 .deg.C regardless of their base elements. 3. SS 316H showed grain boundary precipitates while a large quantity of precipitates were found within/along the grain boundary for Alloy 625 and 800HT after ageing at higher temperatures.

  10. Irradiation creep of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Science.gov (United States)

    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.

  12. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

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

    Science.gov (United States)

    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 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.

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

    Science.gov (United States)

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

    2010-12-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    Science.gov (United States)

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

    2016-07-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  20. Study of the microstructure and of microhardness variation of a Ni-Fe-Cr austenitic alloy by niobium

    International Nuclear Information System (INIS)

    The mechanisms of hardening and corrosion resistance increase in Ni-Fe-Cr austenitic stainless steels by Nb additions are of interest to nuclear technology Niobium additions to a 321 type stainless steel were made in order to study the microhardness, electrical resistivity and metallography. Experimental measurements results are shown. The effect of Nb additions as a micro-alloying element and the thermal and mechanical processes (cold working in particular) in the microstructure and microhardness properties of the 11% Ni - 70%Fe - 17% Cr austenitic alloys were studied. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

    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

  3. Study of the sensitisation of a highly alloyed austenitic stainless steel, Alloy 926 (UNS N08926), by means of scanning electrochemical microscopy

    OpenAIRE

    Leiva García, Rafael; Akid, R.; Greenfield, D.; Gittens, J.; Muñoz Portero, María José; García Antón, José

    2012-01-01

    The feedback mode of a scanning electrochemical microscope (SECM) was applied to study differences in the reactivity of a highly alloyed austenitic stainless steel, Alloy 926 (UNS N08926), in its unsensitised and sensitised state. Alloy 926 was heated at 825 °C for 1 h in an inert atmosphere in order to produce a sensitised metallurgical condition. Sensitisation was due to chromium carbide formation at the grain boundaries. The oxygen reduction reaction was used as an indicator to monitor the...

  4. Influence of KCl deposit morphology on corrosion of austenitic alloys at 500 C

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, P.; Norell, M.; Gautheron, R. [Dep. of Materials Science and Engineering, Chalmers Univ. of Tech., Goeteborg (Sweden)

    2004-07-01

    In biofuel combustion corrosion of the superheater tubes induced by alkali chlorides in the deposits limits the efficiency in electricity production. The most severe corrosion generally occurs at the edge of the deposits. This location may be governed by the transport through the deposit. While most of the literature is focused on the effect of the deposit composition this study examined how the morphology of solid KCl deposits affects the attack. Coupons of two austenitic alloys (Alloy 310 and Sanicro 28) inside tablets of pressed KCl with different density and thickness were exposed to N{sub 2}5%O{sub 2}10%H{sub 2}O at 500 C for 168h. Prior to the exposure tablets were shaped to examine the effect of thickness gradients, edges and cracks. Potassium chromate and iron-chromium oxides formed for all deposit morphologies and chlorine was frequently observed at the interface to the metal. The thicknesses of the deposit clearly affected that of the reaction products, especially for Alloy 310. The thickest products formed at intermediate deposit thickness. This behaviour is similar to that observed for these alloys in a field test. Cracks in the deposits enhanced the attack. At least for Sanicro 28, the chromate formation was observed to break down the protective chromia and thus accelerate the attack. Both alloys were preferentially attacked at metal grain boundaries. (orig.)

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

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

    S. Buyukakkas; H. Aktas; S. Akturk

    2007-01-01

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

  8. A ferric-austenitic CrNiMoN steel alloy to be used as material to manufacture welded components

    International Nuclear Information System (INIS)

    A chromium-nickel-molybdenum-nitrogen steel alloy (ferritic-austenite) is used to manufacture welded articles which without thermal treatment are resistant to pitting corrosion, intergranular corrosion (Monypenny-Stauss test) or boiling in 65% nitric acid with subsequent cross-breaking test. (IHOE)

  9. Evaluation of neutron irradiation effect on SCC crack growth behaviour for austenitic stainless steels

    International Nuclear Information System (INIS)

    Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x1024n/m2 (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  12. Irradiation Behavior in High Entropy Alloys

    Institute of Scientific and Technical Information of China (English)

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

    2015-01-01

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

  13. Radiation damage simulation studies of selected austenitic and ferritic/martensitic alloys for fusion reactor structural applications

    International Nuclear Information System (INIS)

    Results are given of an investigation of the radiation damage stability of selected austenitic and ferritic alloys following ion bombardment in the Harwell VEC to simulate fusion-reactor exposures up to 110 dpa at temperatures from 425 deg to 625 deg C. Gas production rates appropriate to CTR conditions were simulated using a mixed beam of (4 MeV He + 2 MeV H2) in the ratio 1:4 He:H. A beam of 46 MeV Ni or 20 MeV Cr ions was used in sequence with the mixed gas beam to provide a gas/damage ratio of 13 appm He/dpa at a damage rate of approx. 1 dpa/hr. The materials were investigated using TEM and comprised three austenitic alloys: European reference 316L, 316-Ti, 316-Nb; four high-nickel alloys: Fe/25 Ni/8Cr, Inconel 625, Inconel 706 and Nimonic PE16, and four ferritic/martensitic alloys: FV 448, FV 607, CRM 12 and FI. Some data were obtained for a non-magnetic structural alloy Nonmagne-30. The swelling behaviour is reported. The overall results of the study indicate that on a comparative basis the ferritic alloys are the most swelling-resistant, whilst the high-nickel alloys have an acceptable low swelling response up to 110 dpa. The 316 alloys tested have shown an unfavourable swelling response. (author)

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

    Directory of Open Access Journals (Sweden)

    Serna, A.

    2003-12-01

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

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  16. The role of nitrogen in improving pitting corrosion resistance of high-alloy austenitic and duplex stainless steel welds

    International Nuclear Information System (INIS)

    The effects of nitrogen alloyed shielding gas on weld nitrogen content and pitting corrosion resistance of super austenitic (6%Mo) and super duplex stainless steels have been studied with special emphasis on microsegregation behaviour of Cr, Mo and N. The measurements performed with the 6%Mo steel indicate that all these elements segregate interdendritically in the fully austenitic weld metal. With nitrogen addition to the shielding gas the enrichment of nitrogen to the interdendritic regions is more pronounced than to the dendrite cores due to which the pitting corrosion resistance of the dendrite cores increases only marginally. In the super duplex steel welds nitrogen enriches in austenite increasing its pitting corrosion resistance more effectively. In these welds the pitting corrosion resistance of the ferrite phase remains lower. (orig.)

  17. The role of nitrogen in improving pitting corrosion resistance of high-alloy austenitic and duplex stainless steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Vilpas, M. [VTT Manuf. Technol. (Finland); Haenninen, H. [Helsinki Univ. of Technol., Espoo (Finland). Lab. of Eng. Mater.

    1999-07-01

    The effects of nitrogen alloyed shielding gas on weld nitrogen content and pitting corrosion resistance of super austenitic (6%Mo) and super duplex stainless steels have been studied with special emphasis on microsegregation behaviour of Cr, Mo and N. The measurements performed with the 6%Mo steel indicate that all these elements segregate interdendritically in the fully austenitic weld metal. With nitrogen addition to the shielding gas the enrichment of nitrogen to the interdendritic regions is more pronounced than to the dendrite cores due to which the pitting corrosion resistance of the dendrite cores increases only marginally. In the super duplex steel welds nitrogen enriches in austenite increasing its pitting corrosion resistance more effectively. In these welds the pitting corrosion resistance of the ferrite phase remains lower. (orig.)

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

    DEFF Research Database (Denmark)

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

    2011-01-01

    Martensitic transformation in a model Fe-Ni-0.6wt%C alloy was investigated at sub-zero Celsius temperature. The influence of the thermal path in determining the conditions leading to the formation of martensite was studied. In the investigation, samples were austenitized and quenched, whereafter...... isochronal (constant cooling rate) and isothermal sub-zero Celsius treatments were applied. Magnetometry was used for describing the overall kinetics of the transformation in terms of the Johnson-Mehl-Avrami-Kolmogorov kinetics. The evolution of the transformation was also investigated with in......-situ synchrotron X-ray diffraction by evaluating austenite and martensite Bragg reflections. Also, the state of internal strain in austenite was determined....

  19. Modelling irradiation creep of zirconium alloys

    International Nuclear Information System (INIS)

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

  20. Effects of austenite grain size and cooling rate on Widmanstaetten ferrite formation in low-alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Bodnar, R.L.; Hansen, S.S. (Bethlehem Steel Corp., PA (United States). Hot Rolled Products Div.)

    1994-04-01

    Deformation dilatometry is used to simulate the hot rolling of 0.20 pct C-1.10 pct Mn steels over a product thickness range of 6 to 170 mm. In addition to a base steel, steels with additions of 0.02 pct Ti, 0.06 pct V, or 0.02 pct Nb are included in the study. The transformation behavior of each steel is explored for three different austenite grain sizes, nominally 30, 55, and 100 [mu]m. In general, the volume fraction of Widmanstaetten ferrite increases in all four steels with increasing austenite grain size and cooling rate, with austenite grain size having the more significant effect. The Nb steel has the lowest transformation temperature range and the greatest propensity for Widmanstaetten ferrite formation, while the amount of Widmanstaetten ferrite is minimized in the Ti steel (as a result of intragranular nucleation of polygonal ferrite on coarse TiN particles). The data emphasize the importance of a refined austenite grain size in minimizing the formation of a coarse Widmanstaetten structure. With a sufficiently fine prior austenite grain size (e.g., [le]30 [mu]m), significant amounts of Widmanstaetten structure can be avoided, even in a Nb-alloyed steel.

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

    Science.gov (United States)

    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.

  2. Alloy development for irradiation performance: program strategy

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

  3. The kinetics of phase transformations of undercooled austenite of the Mn-Ni iron based model alloy

    Directory of Open Access Journals (Sweden)

    E. Rożniata

    2011-12-01

    Full Text Available Purpose: Present work corresponds to the research on the kinetics of phase transformations of undercooled austenite of Mn-Ni iron based model alloy. The kinetics of phase transformations of undercooled austenite of investigated alloy was presented on CCT diagram (continuous cooling transformation. Also the methodology of a dilatometric samples preparation and the method of the critical points determination were described.Design/methodology/approach: The austenitising temperature was defined in a standard way i.e. 30-50°C higher than Ac3 temperature for model alloy. A technique of full annealing was proposed for the model alloy. The CCT diagrams were made on the basis of dilatograms recorded for samples cooled at various rates. The microstructure of each dilatometric sample was photographed after its cooling to the room temperature and the hardness of the samples was measured.Findings: The test material was a Mn-Ni hypoeutectoid iron based alloy. The microstructure of test Mn-Ni alloy on CCT diagram changes depending on the cooling rate. At the cooling rates of 10°C/s and 5°C/s there is ferrite in Widmannstätten structure present in the structure of tested alloy.Research limitations/implications: The new Mn-Ni iron based model alloy and a new CCT diagram.Practical implications: The paper contains a description of one from a group of iron based model alloys with 0.35-0.40% carbon content. According to PN-EN 10027 standard this steel should have a symbol 38MnNi6-4.Originality/value: The new Mn-Ni iron based model alloy.

  4. Effect of austenitizing and tempering conditions on the structure and mechanical properties of the 9Cr-1Mo martensitic alloy

    International Nuclear Information System (INIS)

    The structure and mechanical properties of the 9Cr-1Mo martensitic alloy, planned to be used as structural materials of the fuel subassembly for fast breeder reactors, has been investigated. Phase transformation temperatures on heating and the continuous cooling transformation diagram were determined by dilatometric techniques. Results concerning the effect of solution-treatment and tempering conditions on austenitic grain size, hardness, tensile properties, creep strength and toughness impact curves are also given

  5. Self-irradiation study of plutonium alloys

    International Nuclear Information System (INIS)

    The plutonium is unstable and produces α or β decays depending on the isotope. These decays generate americium, uranium, helium and different kinds of structural defects. The effects of self-irradiation damage are observed at macroscopic scale, the mechanism occurs from atomic scale. In order to improve our understanding of the self-irradiation effects in PuGa alloys, a technique sensitive to the vacancies and vacancies clusters has been developed: the Positron Annihilation Spectroscopy (PAS). The swelling has been characterized by XRD at a microscopic scale and by dilatometry at a macroscopic scale. Swelling starts just after melting and reaches a saturation between 6 and 36 months depending on the degree of gallium homogeneity in the alloy. Swelling at saturation increases with the gallium content, but the absolute change in the cell parameters is constant during time. PAS showed that vacancies clusters develop immediately. Their concentration increase with time. A part of these clusters is stabilized by helium atoms and leads to the creation of bubbles, which contribution to swelling is negligible. The vacancies and vacancies clusters which are not stabilized by helium contribute to the swelling increase by mechanisms known for other materials. These mechanisms are based on a 'dislocation bias'. The presence of these dislocations can furthermore explain the low mean life time value of positrons at the saturation point. (author)

  6. Relations between the Lattice Parameter and the Stability of Austenite againstεMartensite for the Fe-Mn Based Alloys

    Institute of Scientific and Technical Information of China (English)

    Xing LU; Zuoxiang QIN; Xing TIAN; Yansheng ZHANG; Bingzhe DING; Zhuangqi HU

    2003-01-01

    The influences of lattice parameter of austenite, the electron concentration, the yield strength of parent phase on γ→εmartensite start temperature Ms in the Fe-Mn alloys containing C, Al, Ge and Si have been experimentally investigated. Theresults show that the lattice parameter of austenite is more important than the electron concentration and the yield strength ofparent phase in governing the γ→ε martensitic transformation in Fe-Mn based alloys. A relation between the Ms and latticeparameter of austenite in Fe-Mn based alloys is suggested. The elements Mn, C, Al, Ge, which increase the lattice parameterof austenite lower the Ms; while the element Si, which decreases the lattice parameter increases the Ms. The depressing effectof antiferromagnetic transition on the γ→ε martensitic transformation may be related to the increase of lattice parameterdue to the positive magnetostriction during the antiferromagnetic transition.

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

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

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

  8. Irradiation induced surface segregation in concentrated alloys: a contribution; Contribution a l`etude de la segregation de surface induite par irradiation dans les alliages concentres

    Energy Technology Data Exchange (ETDEWEB)

    Grandjean, Y.

    1996-12-31

    A new computer modelization of irradiation induced surface segregation is presented together with some experimental determinations in binary and ternary alloys. The model we propose handles the alloy thermodynamics and kinetics at the same level of sophistication. Diffusion is described at the atomistic level and proceeds vis the jumps of point defects (vacancies, dumb-bell interstitials): the various jump frequencies depend on the local composition in a manner consistent with the thermodynamics of the alloy. For application to specific alloys, we have chosen the simplest statistical approximation: pair interactions in the Bragg Williams approximation. For a system which exhibits the thermodynamics and kinetics features of Ni-Cu alloys, the model generates the behaviour parameters (flux and temperature) and of alloy composition. Quantitative agreement with the published experimental results (two compositions, three temperatures) is obtained with a single set of parameters. Modelling austenitic steels used in nuclear industry requires taking into account the contribution of dumbbells to mass transport. The effects of this latter contribution are studied on a model of Ni-Fe. Interstitial trapping on dilute impurities is shown to delay or even suppress the irradiation induced segregation. Such an effect is indeed observed in the experiments we report on Fe{sub 50}Ni{sub 50} and Fe{sub 49}Ni{sub 50}Hf{sub 1} alloys. (author). 190 refs.

  9. Fracture toughness of irradiated stainless steel alloys

    International Nuclear Information System (INIS)

    The postirradiation fracture toughness responses of Types 316 and 304 stainless steel (SS) wrought products, cast CF8 SS and Type 308 SS weld deposit were characterized at 4270C using J/sub R/-curve techniques. Fast-neutron irradiation of these alloys caused an order of magnitude reduction in J/sub c/ and two orders of magnitude reduction in tearing modulus at neutron exposures above 10 dpa, where radiation-induced losses in toughness appeared to saturate. Saturation J/sub c/ values for the wrought materials ranged from 28 to 31 kJ/m2; the weld exhibited a saturation level of 11 kJ/m2. Maximum allowable flaw sizes for highly irradiated stainless steel components stressed to 90% of the unirradiated yield strength are on the order of 3 cm for the wrought material and 1 cm for the weld. Electron fractographic examination revealed that irradiation displacement damage brought about a transition from ductile microvoid coalescence to channel fracture, associated with local separation along planar deformation bands. The lower saturation toughness value for the weld relative to that for the wrought products was attributed to local failure of ferrite particles ahead of the advancing crack which prematurely initiated channel fracture

  10. Stabilization of retained austenite by the two-step intercritical heat treatment and its effect on the toughness of a low alloyed steel

    International Nuclear Information System (INIS)

    Highlights: • Fine film-like stable retained austenite was obtained in a low alloyed steel. • Stabilization of retained austenite was studied. • Intercritical partition of C, Mn and Ni was revealed by TEM study. • Effect of retained austenite on toughness was investigated. • Fracture process of the steel was studied by instrument impact test. - Abstract: Fine film-like stable retained austenite was obtained in a Fe–0.08C–0.5Si–2.4Mn–0.5Ni in weight percent (wt.%) steel by the two-step intercritical heat treatment. The first step of intercritical annealing creates a mixed microstructure of preliminary alloy-enriched martensite and lean alloyed intercritical ferrite, which is called as “reverted structure” and “un-reverted structure”, respectively. The second step of intercritical tempering is beneficial for producing film-like stable reverted austenite along the reverted structure. The stabilization of retained austenite was studied by using scanning electron microscopy (SEM), transmission electron microscopy (TEM), dilatometry and X-ray diffraction (XRD) analysis. The two-step austenite reverted transformation associated with intercritical partition of C, Mn and Ni is believed to be the underlying basis for stabilization of retained austenite during the two-step intercritical heat treatment. Stable retained austenite is not only beneficial for high ductility, but also for low temperature toughness by restricting brittle fracture. With 10% (volume fraction) of retained austenite in the steel, high low temperature toughness with average Charpy impact energy of 65 J at −80 °C was obtained

  11. GRAIN-BOUNDARY PRECIPITATION UNDER IRRADIATION IN DILUTE BINARY ALLOYS

    Institute of Scientific and Technical Information of China (English)

    S.H. Song; Z.X. Yuan; J. Liu; R.G.Faulkner

    2003-01-01

    Irradiation-induced grain boundary segregation of solute atoms frequently bring about grain boundary precipitation of a second phase because of its making the solubility limit of the solute surpassed at grain boundaries. Until now the kinetic models for irradiation-induced grain boundary precipitation have been sparse. For this reason, we have theoretically treated grain boundary precipitation under irradiation in dilute binary alloys. Predictions ofγ'-Ni3Si precipitation at grain boundaries ave made for a dilute Ni-Si alloy subjected to irradiation. It is demonstrated that grain boundary silicon segregation under irradiation may lead to grain boundaryγ'-Ni3 Si precipitation over a certain temperature range.

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

    Science.gov (United States)

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

    2016-08-01

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

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

    CERN Document Server

    Subbotina, A V

    2016-01-01

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

  14. Effects of alloying elements and solution-annealing temperature on the mechanical properties of austenitic Fe-Mn-C alloy

    International Nuclear Information System (INIS)

    In order to investigate the effects of various alloying elements including S as a free-machining element on the mechanical properties of high manganese non-magnetic steel, tensile and Charpy impact tests were carried out in the annealed condition. The mechanism of the observed large strengthening effect of V especially on the 0.2% proof stress was investigated by examining Petch relation and its solution hardening effect. A linear regression equation which relates the 0.2% proof stress to the chemical composition is obtained. The strengthening effect of ferrite-forming substitutional element becomes greater in the order of Cr, Mo and V. Especially, the effect of V on the 0.2% proof stress is comparable with that of interstitial element C. While, austenite-forming substitutional elements Ni and Mn have little effect on the strength. The elongation and Charpy impact toughness show decreasing tendencies by the additions of ferrite-forming substitutional elements and S. However, interstitial elements C and N hardly decrease the elongation irrespective of their large strengthening effect. 0.2% proof stress and tensile strength decrease with increasing solution annealing temperature and a Petch relation is found. The large strengthening effect of V cannot be explained by its small solution hardening effect and is rather considered to be mainly attributable to grain refining by the V addition. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  16. Triple Ion-Beam Studies of Radiation Damage Effects in a 316LN Austenitic Alloy for a High Power Spallation Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Lee, EH

    2001-08-01

    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 2}, 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 {approx} 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.

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

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

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

  18. Ni segregation and thermal stability of reversed austenite in a Fe-Ni alloy processed by QLT heat treatment

    Institute of Scientific and Technical Information of China (English)

    Tao Pan; Jing Zhu; Hang Su; Cai-Fu Yang

    2015-01-01

    High-resolution transmission electron microscopy (HRTEM) and X-ray diffraction (XRD) were used to investigate Ni segregation and thermal stability of reversed austenite (RA) in a Fe-Ni alloy processed by quenchlamellarize-temper (QLT) heat treatment.The results show that the 77 K impact energy of the alloy increases with RA content increasing.As an austenite-stabilizing element,Ni is found to segregate in RA,though Ni is not evenly distributed within RA.The amount of segregations increases near the boundary (twice as high as the balanced content) and decreases to some extent in the center of the RA regions.Ni concentration in matrix near the boundary is lower than that in matrix far from the boundary because of Ni atom transportation from α to γ near the boundary.RA in this alloy has high heat and mechanical stability but is likely to lose its stability and transform to martensite when a mechanical load is applied at ultralow temperatures (77 K),which induces plasticity.

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

    Science.gov (United States)

    Rayner, Garrett

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  1. Development of hard intermetallic coatings on austenitic stainless steel by hot dipping in an Al-Si alloy

    OpenAIRE

    Frutos, E.; González-Carrasco, José Luis; Capdevila, Carlos; Jiménez, José Antonio

    2009-01-01

    The austenitic stainless steel was coated by dipping it into a molten Al–12.4%Si alloy at 765 °C. The effect of immersion times in the range of 60 to 900 s was investigated with respect to the crystalline structure, thickness, and microhardness of the coating. A uniform layer (~12 μm) of intermetallic Al12(Fe,Cr)3Si2 with hexagonal crystalline structure is formed, irrespective of the immersion time. Incorporation of Si to the coating changes the growth mode of the coating from inw...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-08-28

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

  3. A study on the constitutive model of irradiated austenitic stainless steel for the functionality analysis of nuclear internals

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Il Sup [Yeungnam University, Gyeongsan (Korea, Republic of)

    2016-04-15

    The internal components of nuclear reactor are exposed to the neutron irradiation environment. The constituent materials of the components are susceptible to remarkable changes in the mechanical properties such as elastic modulus, strength, ductility and toughness. The mechanical and thermal deformations are accompanied with the void swelling and the irradiation creep in the environment. The functionality analysis which evaluates the structural integrity of the aged internals needs to take the degradation characteristics of the material into account. In this paper, a constitutive model of austenitic stainless steel developed by EPRI is studied and implemented into numerical analysis vehicles. The mechanical properties of irradiated 304 stainless steel are presented and the deformation behaviors are simulated. The criteria and methodology for the functionality analysis are also discussed and illustrated.

  4. Constitutive description of flow behaviour of post-irradiated type 316 austenitic stainless steel at low dpa

    Energy Technology Data Exchange (ETDEWEB)

    Christopher, J.; Choudhary, B.K., E-mail: bkc@igcar.gov.in; Kumar, Ran Vijay; Karthik, V.

    2015-09-15

    Highlights: • Tensile flow behaviour of irradiated 316 SS has been examined. • Annihilation of network dislocations and dislocation loops at 623 K. • Insignificant influence of annihilation of dislocation loops at 300 K. - Abstract: Tensile flow behaviour of type 316 austenitic stainless steel irradiated at 623 K up to 2.57 dpa has been examined in the framework of internal-variable approach based on the evolution of network dislocation and irradiation induced defect (dislocation loops) densities with plastic strain at 300 and 623 K. Apart from network dislocation annihilation, the dominance of the annihilation of dislocation loops on strain softening at 623 K has been demonstrated. Insignificant influence of dislocation loops annihilation was observed during deformation at 300 K. Dominance of network dislocation annihilation on strain softening at 300 K was observed.

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

    Directory of Open Access Journals (Sweden)

    Białobrzeska B.

    2015-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  7. Microstructural characterization and modeling of the hardening of irradiated austenitic steels from the internal structures of PWRs

    International Nuclear Information System (INIS)

    The screws and bolts of the lower internal structures of PWRs made of 316L cold-drawn austenitic steels is submitted to a neutron flux at a temperature comprised between 280 deg. C and 380 deg. C, which modifies their operation properties. These modifications of the mechanical properties are the consequence of the modifications of the microstructure of this steel which depends on the flux, fluence, reactor spectrum and irradiation temperature. Samples of 316L cold-drawn steels irradiated in a mixed flux reactor (Osiris at 330 deg. C between 0.8 dpa and 3.4 dpa) and in fast breeder reactors (Bor-60 at 330 deg. C up to 40 dpa and EBR-II at 375 deg. C up to 10 dpa) have been observed in transmission electron microscopy. Irradiation defects are Frank dislocation loops and the presence of cavities has been evidenced in materials irradiated at 375 deg. C. The evolution of the irradiation loops population has been modeled using an 'accumulation dynamics'-type simulation. The adjustment of the parameters of the model has permitted to describe quantitatively the experimental results. This description of the irradiation microstructure has been coupled with a Frank loops hardening model which has permitted to describe the observed hardening. The range of explored doses goes up to 40 dpa and is representative of the irradiation dose corresponding to the half life of the reactors design. (J.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

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

    2016-02-01

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

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

    Science.gov (United States)

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

    2016-12-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    NARCIS (Netherlands)

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

    2012-01-01

    We present results of an extensive set of first-principles density functional theory calculations of point defect formation, binding, and clustering energies in austenitic Fe with dilute concentrations of Cr and Ni solutes. A large number of possible collinear magnetic structures were investigated a

  14. Irradiation growth of titanium alloy VT1-0 under proton irradiation

    International Nuclear Information System (INIS)

    A specially developed procedure was used to study the irradiation growth of the rods of titanium alloy VT1-0 under proton irradiation. There was determined the relation between the dimensional changes induced by irradiation growth and the texture. The effect of various types of heat-treatment on the texture, structure and irradiation growth of the VT1-0 rods was studied. It is demonstrated that destruction of the initial texture of VT1-0 rods by the mechanical and microwave heat-treatment results in almost complete suppression of irradiation growth

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

    Science.gov (United States)

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

    2016-05-01

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

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

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

    2016-11-01

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

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

    Science.gov (United States)

    Maksimkin, O. P.

    2016-04-01

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

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

    Science.gov (United States)

    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.

  20. Gas porosity in metals and alloys irradiated by helium ions

    International Nuclear Information System (INIS)

    Experimental studies of the development of gas porosity in metals and alloys during irradiation with helium ions up to high doses and during post-irradiation annealings, are reviewed. The main theoretical problems of the mechanisms of bubble formation and growth, the regularities and peculiarities of bubble development in a thin near-the surface layer during the introduction of helium with the energy of tens of kiloelectron volt, are considered

  1. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

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

    Science.gov (United States)

    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.

  4. 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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    The U.S. 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

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

    Institute of Scientific and Technical Information of China (English)

    Y. Sun

    2004-01-01

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

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

    Science.gov (United States)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

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

  9. Evolution of microstructure in zirconium alloys during irradiation

    CERN Document Server

    Griffiths, M; Winegar, J E

    1997-01-01

    X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 x 10 sup 2 sup 5 n.m sup - sup 2 (E > I MeV) for a range of temperatures between 330 and 580 K. Irradiation modifies the dislocation structure through nucleation and growth of dislocation loops and, for cold-worked materials in particular, climb of existing network dislocations. In general, the a-type dislocation structure tends to saturate at low fluences (10 x l0 sup 2 sup 5 n.m sup - sup 2 - in some cases). The phase structure is also modified by irradiation. The common alloying/impurity elements, Fe, Cr, and Ni, are relatively insoluble in the alpha-phase but are dispersed into the alpha-phase during irradiation irrespective of the state of the phase initial...

  10. Phases stability of shape memory alloys Cu based under irradiation

    International Nuclear Information System (INIS)

    The effects of irradiation on the relative phase stability of phases related by a martensitic transformation in copper based shape memory alloys were studied in this work.Different kind of particles and energies were employed in the irradiation experiments.The first kind of irradiation was performed with 2,6 MeV electrons, the second one with 170 keV and 300 keV Cu ions and the third one with swift heavy ions (Kr, Xe, Au) with energies between 200 and 600 MeV.Stabilization of the 18 R martensite in Cu-Zn-Al-Ni induced by electron irradiation was studied.The results were compared to those of the stabilization induced by quenching and ageing in the same alloy, and the ones obtained by irradiation in 18 R-Cu-Zn-Al alloys.The effects of Cu irradiation over b phase were analyzed with several electron microscopy techniques including: scanning electron microscopy (S E M), high resolution electron microscopy (H R E M), micro diffraction and X-ray energy dispersive spectroscopy (E D S). Structural changes in Cu-Zn-Al b phase into a closed packed structure were induced by Cu ion implantation.The closed packed structures depend on the irradiation fluence.Based on these results, the interface between these structures (closed packed and b) and the stability of disordered phases were analyzed. It was also compared the evolution of long range order in the Cu-Zn-Al and in the Cu-Zn-Al-Ni b phase as a function of fluence.The evolution of the g phase was also compared. Both results were discussed in terms of the mobility of irradiation induced point defects.Finally, the effects induced by swift heavy ions in b phase and 18 R martensite were studied. The results of the irradiation in b phase were qualitatively similar to those produced by irradiation with lower energies. On the contrary, nano metric defects were found in the irradiated 18 R martensite.These defects were characterized by H R E M.The characteristic contrast of the defects was associated to a local change in the

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-07-01

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

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

    Science.gov (United States)

    Khan, Mahmud; Brock, Jeffrey; Sugerman, Ian

    2016-02-01

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

  16. Optimized chemical composition, working and heat treatment condition for resistance to irradiation assisted stress corrosion cracking of cold worked 316 and high-chromium austenitic stainless steel

    International Nuclear Information System (INIS)

    The authors have reported that the primary water stress corrosion cracking (PWSCC) in baffle former bolts made of austenitic stainless steels for PWR after long-term operation is caused by irradiation-induced grain boundary segregation. The resistance to PWSCC of simulated austenitic stainless steels whose chemical compositions are simulated to the grain boundary chemical composition of 316 stainless steel after irradiation increased with decrease of the silicon content, increases of the chromium content, and precipitation of M23C6 carbides at the grain boundaries. In order to develop resistance to irradiation assisted stress corrosion cracking in austenitic stainless steels, optimized chemical compositions and heat treatment conditions for 316CW and high-chromium austenitic stainless steels for PWR baffle former bolts were investigated. For 316CW stainless steel, ultra-low-impurities and high-chromium content are beneficial. About 20% cold working before aging and after solution treatment has also been recommended to recover sensitization and make M23C6 carbides coherent with the matrix at the grain boundaries. Heating at 700 to 725degC for 20 to 50 h was selected as a suitable aging procedure. Cold working of 5 to 10% after aging produced the required mechanical properties. The optimized composition of the high-chromium austenitic stainless steel contents 30% chromium, 30% nickel, and ultra-low impurity levels. This composition also reduces the difference between its thermal expansion coefficient and that of 304 stainless steel for baffle plates. Aging at 700 to 725degC for longer than 40 h and cold working of 10 to 15% after aging were selected to meet mechanical property specifications. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

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

    Science.gov (United States)

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

    2004-08-01

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

  1. State diagram of copper-aluminium alloys after neutron irradiation

    International Nuclear Information System (INIS)

    It is ascertained that under reactor irradiation of copper-aluminium alloys (18.0-31.2 at% of Al) radiation-induced phase transformations occur, alpha-phase is decomposed into two ones with alpha'-phase precipitation, in gamma2-phase separate regions of its high-temperature disordered modification (gamma1-phase) are formed. Thermal stability of precipitations is investigated, regions of their existence are defined on the state diagram

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-23

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    International Nuclear Information System (INIS)

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

  6. Microstructural examination of irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  7. Influence of helium and other impurities on mechanical properties of irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Manichev, V.M. [Rossijskij Nauchnyj Tsentr, Moscow (Russian Federation). Kurchatovskij Inst.; Ryazonov, A.I. [Rossijskij Nauchnyj Tsentr, Moscow (Russian Federation). Kurchatovskij Inst.; Witzenburg, W. van [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1994-09-01

    Helium preimplantation and neutron irradiation influence on the tensile properties of the vanadium alloys in the irradiation/test temperature range of 500-800 C was investigated and principal physical mechanisms of mechanical behaviour of irradiated alloys were considered. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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

  9. Atomic scale effects of alloying, partitioning, solute drag and austempering on the mechanical properties of high-carbon bainitic–austenitic TRIP steels

    International Nuclear Information System (INIS)

    Understanding alloying and thermal processing at an atomic scale is essential for the optimal design of high-carbon (0.71 wt.%) bainitic–austenitic transformation-induced plasticity (TRIP) steels. We investigate the influence of the austempering temperature, chemical composition (especially the Si:Al ratio) and partitioning on the nanostructure and mechanical behavior of these steels by atom probe tomography. The effects of the austempering temperature and of Si and Al on the compositional gradients across the phase boundaries between retained austenite and bainitic ferrite are studied. We observe that controlling these parameters (i.e. Si, Al content and austempering temperature) can be used to tune the stability of the retained austenite and hence the mechanical behavior of these steels. We also study the atomic scale redistribution of Mn and Si at the bainitic ferrite/austenite interface. The observations suggest that either para-equilibrium or local equilibrium-negligible partitioning conditions prevail depending on the Si:Al ratio during bainite transformation.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-01

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

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

    Directory of Open Access Journals (Sweden)

    Vicente Braz Trindade

    2005-12-01

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

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

    Science.gov (United States)

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

    2000-01-01

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

  14. Electron irradiation-induced mechanical property changes in reactor pressure vessel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Alexander, D.E.; Rehn, L.E. [Argonne National Lab., IL (United States); Odette, G.R.; Lucas, G.E. [California Univ., Santa Barbara, CA (United States). Dept. of Mechanical Engineering

    1995-11-01

    High-energy electrons were used to study tensile property changes in simple Fe-Cu and Fe-Cu-Mn alloys irradiated at 288C. A comparison was made with neutron irradiation data on the same alloys. An apparent effect of alloy chemistry was observed in which the presence of Mn affected embrittlement differently for electron and neutron irradiation. Comparison of previous experimental studies with the present experimental results indicates that electrons may be more efficient than fast neutrons at producing embrittlement.

  15. IASCC crack growth rate of neutron irradiated low carbon austenitic stainless steels in simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Chatani, K. [Nippon Nuclear Fuel Development Co. Ltd (NFD), Oarai (Japan); Takakura, K.; Ando, M.; Nakata, K. [Japan Nuclear Energy Safety Organization (JNES), Tokyo (Japan); Tanaka, S. [Toshiba Corp., Yokohama (Japan); Ishiyama, Y. [Hitachi Ltd., Hitachi (Japan); Hishida, M. [Inst. of Research and Innovation (IRI), Tokyo (Japan); Kaji, Y. [Japan Atomic Energy Agency (JAEA), Tokai (Japan)

    2007-07-01

    Crack Growth Rate (CGR) tests have been conducted with neutron irradiated Compact Tension (CT) specimens. The specimens were irradiated at core region of Japan Material Testing Reactor (JMTR) in simulated BWR water environments at 288 {sup o}C. The CGR tests of 316L and 304L base metals irradiated from 0.516 to 1.07 x 10{sup 25} n/m{sup 2} (E>1MeV), 316L and 308L weld metals irradiated up to 0.523 to 0.541 x 10{sup 25} n/m{sup 2} (E>1MeV) were performed with reversing DC potential drop method under constant load in a few stress intensity factor (K) and corrosion potential (ECP) conditions at 288 {sup o}C in water. CGRs of base metals were increased with increasing neutron fluence, Clear reductions in CGRs of base metals and weld metals were measured with decreasing ECP levels. This paper will discuss the relationship between CGR and radiation hardening / RIS. (author)

  16. Study on irradiation assisted stress corrosion cracking of austenitic stainless steels in nuclear energy environments (Thesis)

    International Nuclear Information System (INIS)

    Irradiation assisted stress corrosion cracking (IASCC) is known as a degradation phenomenon that is caused by synergistic effects of neutron /gamma irradiation, stress/strain and high temperature water on structural materials. It is important to investigate stress corrosion cracking (SCC) and IASCC mechanisms from the viewpoint of the safety and reliability improvement in the nuclear energy system. To evaluate the influence of minor additional elements, heat treatment, cold working and neutron fluence on IASCC behavior, a slow strain rate technique (SSRT) facility for irradiated specimens has been developed and post irradiation examinations have been conducted. Based on results obtained from the IASCC studies, discussion regarding IASCC susceptibility, crack initiation and growth behaviors are described comprehensively in this paper. The followings are summarized typical findings. (1) 1 or 2 cracks of IASCC are introduced at 98-99 % of maximum stress. (2) The increase and decrease in crack growth rate are repeated alternately in the process of crack growth. (3) Although suppression of radiation hardening can be introduced with Si or Mo addition, the suppression disappear with increasing in neutron fluence. (4) Fracture mode changes from intergranular (IG) SCC to transgranular (TG) SCC with increasing in hardness which is introduced with neutron radiation and /or cold working. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  19. Magnetocaloric and critical behavior in the austenitic phase of Gd-doped Ni{sub 50}Mn{sub 37}Sn{sub 13} Heusler alloys

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, P.; Phan, T.L.; Dan, N.H. [Department of Physics, Chungbuk National University, Cheongju 361-763 (Korea, Republic of); Thanh, T.D. [Department of Physics, Chungbuk National University, Cheongju 361-763 (Korea, Republic of); Institute of Materials Science, Vietnam Academy of Science and Technology, Hanoi (Viet Nam); Yu, S.C., E-mail: scyu@chungbuk.ac.kr [Department of Physics, Chungbuk National University, Cheongju 361-763 (Korea, Republic of)

    2014-12-05

    Highlights: • The martensitic phase of Ni{sub 50}Mn{sub 37}Sn{sub 13} Heusler alloy was suppressed by Gd doping. • The ferromagnetism in the austenitic phase basically belongs to the mean-field. • Ferromagnetic order can be easily influenced by the magnetically inhomogeneity. - Abstract: The magnetic phase transition behavior were investigated in detail in Ni{sub 50−y}Gd{sub y}Mn{sub 37}Sn{sub 13} (y = 1 and 3) alloys prepared by arc-melting method. The martensite phase was found to be strongly suppressed by a small amount of Gd doping. Based on isothermal magnetization curves around Curie temperature of the austenite (T{sub C}{sup A}) phase, critical behavior in the austenite phases of both alloys were determined carefully by the Kouvel–Fisher method. The critical exponents were found to be β = 0.473 ± 0.020 and γ = 1.141 ± 0.017 with T{sub C}{sup A} = 299.0 ± 0.2 K for y = 1, and β = 0.469 ± 0.068 and γ = 1.214 ± 0.042 with T{sub C}{sup A} = 302.9 ± 0.7 K for y = 3, respectively. The values of the critical exponents for the ferromagnetic phase transition in the A phase of two alloys can be basically ascribed in the mean-field model (with β = 0.5, γ = 1) with slightly deviation, revealing a long-range order of ferromagnetic interactions. Such critical behavior can be attributed to the magnetic inhomogeneities originated from the atomic disorder introduced by Gd doping.

  20. Radiation effects in the aluminium alloys irradiated with neutrons

    International Nuclear Information System (INIS)

    Full text: Materials of fuel elements for water cooled nuclear reactors are exposed to simultaneous action of an ionizing radiation, temperature and yields of water radiolysis. In particular, irradiation by fast neutrons (En> 0.1 MeV) in research reactors influences mainly the mechanical properties of aluminium alloys, increasing their strength and reducing the plasticity. Radiation can essentially affect the stability of the heat-generating assembly material, changing its structure state. The structure change may also be the result of post-radiation ageing. This paper presents the results of studying the influence of reactor neutrons (research reactor of INP AS RU) on microstructure, electrical characteristics and length changes of SAV-1 and AMG-2 aluminium alloys used in nuclear industry. These alloys are low-alloyed solid solutions and intermetallic phases of CuAl2, Mg2Si, CuMgAl2, CuMg4Al6, Al2Mg2 in an equilibrium state. Samples were cut with orientation in 111 crystallographic axis in the shape of disks with the diameter d= 15 mm and thickness h= 3 mm for the metallographic analysis, and rods with the length of 40 mm and width d = 5 mm for measuring specific electrical resistance and linear dimension changes prior and after irradiations. For precise measurements the sample surfaces were mechanically handled and polished in a chemical solution, and then washed out in the distilled water and ethanol. Further samples, were put into the aluminum container and irradiated in a vertical channel of the reactor to fluencies 1018, 1019, 1020 n/cm2. The relative elongation (extension) δ was calculated as the measured length ratio of the non-irradiated and irradiated sample: δ=L0/L1x100%. Determination of element composition and the metallographic analysis of studied samples were done at the X-ray microanalyzer 'Jeol' JSM 5910 IV. Specific resistance (ρ) values were measures with four probe technique by compensation method at the direct voltage. The sample lengths

  1. Hardness modification of Al–Mg–Si alloy by using energetic ion beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ueyama, D. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Saitoh, Y. [Takasaki Advanced Radiation Research Institute, Japan Atomic Energy Agency, Takasaki, Gunma 370-1292 (Japan); Ishikawa, N. [Tokai Research and Development Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Ohmura, T. [Structural Metals Center, National Institute for Materials Science, Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Semboshi, S. [Kansai-Center, Institute of Materials Research, Tohoku University, Sakai, Osaka 599-8531 (Japan); Hori, F. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Iwase, A., E-mail: iwase@mtr.osakafu-u.ac.jp [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan)

    2015-05-15

    So far, we have irradiated Al–Mg–Si alloy with 5.4–16 MeV several ions at room temperature, and have found that ion irradiation is a useful tool for controlling the surface hardness for Al–Mg–Si alloys. In the present study, we tried several experiments as some applications of ion beam irradiation for hardness modifications of Al–Mg–Si alloy. Main results are as follows; (1) the combination of ion beam irradiation and the subsequent thermal aging can be used as an effective tool for the hardness modification of Al–Mg–Si alloy, and (2) designated regions and areas of the specimen can be hardened by changing the energy of ion beam and producing the irradiated area and unirradiated area of the surface. Then, we can expand the possibility of the ion beam irradiation as a new process for the three-dimensional hardness modification of Al–Mg–Si alloy.

  2. A modeling of radiation induced microstructural evolution under applied stress in austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hiroyasu; Kohyama, Akira [Kyoto Univ., Uji (Japan). Inst. of Advanced Energy; Katoh, Yutai; Kohno, Yutaka

    1996-10-01

    Effects of applied stress on interstitial type Frank loop evolution at early stages of irradiation were investigated by both numerical calculation and irradiation experiments. In the experimental part of this work, microstructural inspection has been made by transmission electron microscopy with a special emphasis on Frank loops and perfect loops on every {l_brace}111{r_brace} plane. The results of the TEM observation revealed that Frank loop concentration on a {l_brace}111{r_brace} plane increased as the resolved normal stress to a {l_brace}111{r_brace} plane increased and that small perfect loops were more likely produced on {l_brace}111{r_brace} planes where larger resolved shear stress was applied. The model of a stress effect on Frank loop unfaulting was provided, which is triggered by nucleation of a Shockley partial dislocation loop in a Frank loop, was proposed. The results of the numerical calculation was successful to predict the strong dependence of Frank loop concentration on the resolved normal stress to {l_brace}111{r_brace} plane, which was the characteristic feature seen in the irradiation experiments. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W.

    1996-06-01

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

  4. The Self-Irradiation of Plutonium and its Delta Alloys

    International Nuclear Information System (INIS)

    Electrical resistivity and heat capacity measurements have shown that seit-irradiation trom the alpha activity in plutonium produces damage at low temperatures. Measurements were made on specimens of nominally pure, a-phase plutonium and on specimens of plutonium-rich, α-phase, solid-solution alloys containing aluminium. The resistivity was found to increase linearly with time at 20.1°K and 4.5°K in both phases, but the rates of increase in α-phase are significantly higher than in δ-phase. Annealing of stored damage was studied by means of electrical resistivity and heat capacity measurements, and the results from both types of measurements are in concordance. Calculations by Vineyard on self-irradiation damage in plutonium are presented, and differences in the radiation damage behaviours of α-phase and δ-phase are discussed and correlated with differences in other physical properties of these phases. (author)

  5. The effect of Alloying elements on pitting resistance of ferritic and austenitic stainless steels in terms of pitting resistance equivalents (PRE)

    International Nuclear Information System (INIS)

    The alloying elements, such as Cr, Mo, and N of stainless steels play important roles in their resistances to pitting corrosion. The pitting resistances of stainless steels ha e long been characterized in terms of electrochemical parameters such as pitting potentials. however, in order to better understand the resistances to pitting of stainless steels, Pit Propagation Rate (PPR) and Critical Pitting Temperature (CPT) tests were carried out in deaerated 0.1N H2SO4 + 0.1N NaCl solution. The effect of Cr, Mo, and N alloying elements on the pitting corrosion resistances of both ferritic Fe-Cr, Fe-Cr-Mo stainless steels and austenitic stainless steels was examined by performing polarization, PPR, and CPT tests. The comparison between test results was made in terms of the Pitting Resistance Equivalent (PRE). Results showed that PRE values are the good parameters representing the extents of pitting corrosion resistance on a single scale regardless of both kinds of alloying elements and types of ferritic or austenitic stainless steels

  6. Fracture mechanics behaviour of neutron irradiated Alloy A-286

    International Nuclear Information System (INIS)

    The effect of fast-neutron irradiation on the fatigue-crack propagation and fracture toughness behaviour of Alloy A-286 was characterized using fracture mechanics techniques. The fracture toughness was found to decrease continuously with increasing irradiation damage at both 24 deg. C and 427 deg. C. In the unirradiated and low fluence conditions, specimens displayed appreciable plasticity prior to fracture, and equivalent Ksub(Ic) values were determined from Jsub(Ic) fracture toughness results. At high irradiation exposure levels, specimens exhibited a brittle Ksub(Ic) fracture mode. The 427 deg. C fracture toughness fell from 129 MPa√m in the unirradiated condition to 35 MPa√m at an exposure of 16.2 dpa (total fluence of 5.2x1022n/cm2). Room temperature fracture toughness values were consistently 40 to 60 percent higher than the 427 deg. C values. Electron fractography revealed that the reduction in fracture resistance was attributed to a fracture mechanism transition from ductile microvoid coalescence to channel fracture. Fatigue-crack propagation tests were conducted at 427 deg. C on specimens irradiated at 2.4 dpa and 16.2 dpa. Crack growth rates at the lower exposure level were comparable to those in unirradiated material, while those at the higher exposure were slightly higher than in unirradiated material. (author)

  7. Void formation in neutron-irradiated Cu and Cu alloys

    International Nuclear Information System (INIS)

    Pure copper and copper-aluminium alloys were neutron-irradiated at high temperatures in the as-received condition, and after being melted under high vacuum or in argon. Melting under high vacuum was done to reduce the residual gas amount in the specimens. The number density of voids in the vacuum-melted Cu was one tenth of that in as-received Cu after JMTR irradiation to 5.2 x 1024 n/m2 at 603 K. Similarly, voids were also formed in an argon-melted Cu-1at%Al specimen but were not formed in a vacuum-melted one. Following higher dose irradiation in the JOYO reactor, nearly the same number density and size of voids were formed in both argon and vacuum-melted Cu. In Cu-5at%Al, many voids were formed in argon-melted specimens, whereas in vacuum-melted specimens voids were not formed. These results show that voids nucleate at vacancy clusters which trap gas atoms. In the JOYO irradiation, diffused-in gas atoms play an important role in the formation of voids in Cu. In Cu-5at%Al, diffused-in gas atoms were trapped by Al atoms, resulting in a difference of void formation between the two types of specimens. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, B.N.

    1998-01-01

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

  10. Heavy ion irradiation with 200 keV Ni-ions to study the swelling of iron and nickel and the technical alloys

    International Nuclear Information System (INIS)

    Radiation induced swelling of iron, nickel and of the steels 1.4914 and 1.4970 was studied. Selection criterion for the materials was the equal lattice structure of iron and 1.4914 as well as of nickel and 1.4970. The relative biasfactors of edge dislocations in iron and nickel were determined from swelling measurements. The experimentally determined biasfactor in iron is lower than in nickel, in good agreement with calculated biasfactors. The technical alloys 1.4970 and 1.4914 show the same maximum swelling. The opposite is known from swelling, induced by neutron- or MeV- ion irradiation. This may be explained by the modification in the chemical composition of the alloys, caused by ion implantation and by segregation. Dislocation- and loop structure are similar in iron and the martensitic steel 1.4914 with a predominant loop structure, whereas in nickel and the austenitic steel 1.4970 line dislocations are dominant. (orig.)

  11. Exploration of the influence of welding variables on notch ductility of irradiated austenitic stainless steel welds

    International Nuclear Information System (INIS)

    Postirradiation notch ductility and fracture toughness (K/sub J/) trends of AISI Type 308 weld deposits were explored for radiation exposures in the range of 260 to 6490C. The welds were produced by the shielded metal arc (SMA) process and represented compositional variations (CRE vs non-CRE) and controlled delta ferrite content variations. Fracture toughness determinations were made with fatigue precracked Charpy-V specimens and J-Integral assessment procedures. Specimen irradiations were conducted in EBR-II and UCRR reactors. Large postirradiation decreases in Charpy-V (C/sub V/) energy absorption and fracture toughness were observed. The possibility for K/sub J/ values to be reduced below 88 MPa√m (80 ksi√in.) by moderate fluence exposures was demonstrated. The SMA weld with CRE appeared more sensitive to radiation than a non-CRE Type 308 submerged arc (S/A) weld. An influence of delta ferrite content on 260 or 6490C radiation resistance was not found. Fracture toughness assessments revealed high weld sensitivity to testing rate: a 2 : 1 difference in K/sub J/ level was observed between static vs dynamic test conditions

  12. Relationship between irradiation swelling behaviour of alloys and their valence electron structure

    International Nuclear Information System (INIS)

    The relationship between valence electron structure of alloys and their irradiation swelling behaviour has been investigated on basis of results of valence electron structure calculated by means of the empirical electron theory. The difference of the irradiation swelling behaviour among three prior candidate alloys has been explained by their different valence electron structure, and the intrinsic relation between nickel content of iron-nickel-based alloys and their irradiation swelling behaviour has been clarified. From the viewpoint of valence electron structure, intermetallic compounds are potential structural materials with excellent resistance to irradiation swelling. (4 tabs.)

  13. Microstructural and microchemical evolution in vanadium alloys by heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Sekimura, Naoto; Kakiuchi, Hironori; Shirao, Yasuyuki; Iwai, Takeo [Tokyo Univ. (Japan)

    1996-10-01

    Microstructural and microchemical evolution in vanadium alloys were investigated using heavy ion irradiation. No cavities were observed in V-5Cr-5Ti alloys irradiated to 30 dpa at 520 and 600degC. Energy dispersive X-ray spectroscopy analyses showed that Ti peaks around grain boundaries. Segregation of Cr atoms was not clearly detected. Co-implanted helium was also found to enhance dislocation evolution in V-5Cr-5Ti. High density of matrix cavities were observed in V-5Fe alloys irradiated with dual ions, whereas cavities were formed only around grain boundaries in single ion irradiated V-5Fe. (author)

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

    Institute of Scientific and Technical Information of China (English)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  16. Compatibility of graphite with a martensitic-ferritic steel, an austenitic stainless steel and a Ni-base alloy up to 1250 C

    International Nuclear Information System (INIS)

    To study the chemical interactions between graphite and a martensitic-ferritic steel (1.4914), an austenitic stainless steel (1.4919; AISI 316), and a Ni-base alloy (Hastelloy X) isothermal reaction experiments were performed in the temperature range between 900 and 1250 C. At higher temperatures a rapid and complete liquefaction of the components occurred as a result of eutectic interactions. The chemical interactions are diffusion-controlled processes and can be described by parabolic rate laws. The reaction behavior of the two steels is very similar. The chemical interactions of the steels with graphite are much faster above 1100 C than those for the Ni-base alloy. Below 1000 C the effect is opposite. (orig.)

  17. Environmentally assisted cracking and irradiation embrittlement of CF-8 and CF-8M cast austenitic stainless steels in high-purity water

    International Nuclear Information System (INIS)

    Cast austenitic stainless steels (CASS) are used for components with complex geometries in the cooling system of light water reactors (LWRs). Due to both thermal ageing and irradiation embrittlement, the long-term performance of CASS materials is of concern. To assess the impact of thermal ageing and irradiation embrittlement on the cracking behaviour of CASS materials, crack growth rate and fracture toughness JR curve tests were carried out on CF-8 and CF-8M compact-tension specimens in high-purity water with low dissolved oxygen. The as-received and thermally aged CASS specimens were irradiated to 0.08 dpa to investigate the combined effect of thermal ageing and neutron irradiation. The crack growth rates of irradiated CASS materials were compared with previous results on unirradiated specimens. While no elevated cracking susceptibility was observed for the irradiated specimens at this dose level, a slightly better corrosion fatigue performance was found in the CF-8 than in CF-8M materials. Thermal ageing history had little effect on the crack growth behaviour in the test environment. Trans-granular cleavage-like cracking was the main fracture mode in the crack growth rate tests, and delta ferrite morphology could be seen in some areas on the fracture surfaces. Compared to thermal ageing, neutron irradiation had a dominant role in the fracture toughness JR curve tests. The loss of toughness due to neutron irradiation was much more significant in the as-received than in the thermally aged CASS specimens. The fracture toughness of CASS specimens was reduced to a similar level after neutron irradiation regardless of their thermal ageing history. This suggests a more rapid development of embrittlement in the as-received than in the thermally aged CASS specimens under neutron irradiation. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-01-01

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

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

    Science.gov (United States)

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

    2015-04-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    Ohnuki, Soumei; Takahashi, Heishichiro; Takeyama, Taro

    1984-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  3. The effects of. gamma. -irradiation on Ti-Ni shape-memory alloy

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Guilin; Xu Feng; Liu Wenhong; Hu Wenxiang; Yu Fanghua; Zhang Yiping (Academia Sinica, Shanghai, SH (China). Shanghai Inst. of Nuclear Research); Wang Jingcheng; Shao Zichang (Shanghai Iron and Steel Research Inst, SH (China))

    1992-04-01

    Because gamma irradiation provides a means of introducing lattice defects into crystalline solids in a controlled fashion, it can be used to study the influence of lattice defects on the physical properties of solids such as shape-memory alloys (SMAs). The study described here shows that gamma irradiation can be used to ameliorate the performance of SMAs and to understand the mechanism of the shape memory further in these alloys. In particular it shows the effect of gamma irradiation on the martensitic transformation temperatures of Ti-Ni alloys. (UK).

  4. The effects of γ-irradiation on Ti-Ni shape-memory alloy

    International Nuclear Information System (INIS)

    Because gamma irradiation provides a means of introducing lattice defects into crystalline solids in a controlled fashion, it can be used to study the influence of lattice defects on the physical properties of solids such as shape-memory alloys (SMAs). The study described here shows that gamma irradiation can be used to ameliorate the performance of SMAs and to understand the mechanism of the shape memory further in these alloys. In particular it shows the effect of gamma irradiation on the martensitic transformation temperatures of Ti-Ni alloys. (UK)

  5. Mechanical properties of neutron irradiated vanadium alloys under liquid sodium environment

    International Nuclear Information System (INIS)

    Full text of publication follows: Vanadium alloys are candidate materials for fusion reactor blanket structural materials, but its knowledge about the mechanical properties at high temperatures during neutron irradiation is limited and there are uncertainties that may have influenced the results such as the interstitial impurity content of specimens. The objective of this study is to investigate the mechanical properties and microstructural changes of the high-purified V-4Cr-4Ti alloys, NIFS-HEAT2 during neutron irradiation. In this study, tensile test, Charpy impact test and microstructural observation were done for V-4Cr-4Ti alloys and vanadium binary alloys. Small sized tensile specimens, 1.5 Charpy V-notched specimens and TEM specimens of highly purified V-4Cr-4Ti alloys, NIFS-Heat and vanadium binary alloys were irradiated in Joyo in the temperature range from 450 deg. C to 650 deg. C with a damage level from 1 to 5 dpa. In the irradiation experiment, we have developed Na-enclosed irradiation rig in Joyo in order to equalize the irradiation temperature of large scale specimens and prevent the invasion of interstitial impurities from the circumstance in irradiation rig during irradiation for irradiation specimens. After dismantling the Na-enclosed capsule and cleaning the surface of specimens, tensile tests at room temperature, Charpy impact tests and TEM observation were performed. Irradiation hardening and reduction of ductility for NIFS-Heat alloys could be seen at 450 deg. C irradiation in tensile tests, but the destructive loss of plasticity could not be in any vanadium specimens even at 450 deg. C irradiation. Results of Charpy impact test showed that the amounts of upper shelf energy of NIFS-heat specimens irradiated at 450 deg. C and 600 deg. C were about 0.1-0.2 J at room temperature and brittle behavior could not be seen from load displacement relationship and SEM observation of fracture surface. From the TEM observation of NIFS-Heat alloys

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

    Science.gov (United States)

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

    2010-12-01

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

  7. Irradiation-induced creep and microstructural development in precipitation-hardened nickel-aluminium alloys

    International Nuclear Information System (INIS)

    Irradiation-induced creep in solid-solution Ni-8.5 at% AL and precipitation-hardened Ni-13.1 at% Al alloys was studied by bombarding miniaturized specimens with 6.2 MeV protons at 3000C under different tensile stresses. After irradiation transmission electron microscopic (TEM) investigations were made to observe the precipitate structure under irradiation for different experimental parameters. Moreover, the irradiation-induced changes in precipitate structure and changes of Al-concentrations in the matrix in Ni-13.1 at% Al alloys were studied by electrical resistivity measurements during irradiation. For comparison, the electrical resistivity of unirradiated specimens was also measured after thermal aging for different times. For correlation, TEM analysis was performed on irradiated and unirradiated aged specimens. Tensile tests on annealed and aged Ni-Al alloys were also done at various temperatures. (orig./RK)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.

    1996-04-01

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

  9. Neutron irradiation test of copper alloy/stainless steel joint materials

    OpenAIRE

    山田 弘一; 河村 弘

    2006-01-01

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al2O3-dispersed strengthened copper or CuCrZr was joined to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The avera...

  10. Microstructural characterization and modeling of the hardening of irradiated austenitic steels from the internal structures of PWRs; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C.; Dubuisson, P. [CEA Saclay, DMN/SRMA, 91 - Gif-sur-Yvette (France); Massoud, J.P. [Electricite de France (EDF/MMC), 78 - Saint Moret sur Loing (France); Brechet, Y. [Institut Polytechnique de Grenoble, Lab. de Thermodynamique et de Physico Chimie Metallurgiques, CNRS, 38 (France); Barbu, A. [Ecole Polytechnique, Lab. des Solides Irradies, CEA / CNRS, 91 - Palaiseau (France)

    2002-07-01

    The screws and bolts of the lower internal structures of PWRs made of 316L cold-drawn austenitic steels is submitted to a neutron flux at a temperature comprised between 280 deg. C and 380 deg. C, which modifies their operation properties. These modifications of the mechanical properties are the consequence of the modifications of the microstructure of this steel which depends on the flux, fluence, reactor spectrum and irradiation temperature. Samples of 316L cold-drawn steels irradiated in a mixed flux reactor (Osiris at 330 deg. C between 0.8 dpa and 3.4 dpa) and in fast breeder reactors (Bor-60 at 330 deg. C up to 40 dpa and EBR-II at 375 deg. C up to 10 dpa) have been observed in transmission electron microscopy. Irradiation defects are Frank dislocation loops and the presence of cavities has been evidenced in materials irradiated at 375 deg. C. The evolution of the irradiation loops population has been modeled using an 'accumulation dynamics'-type simulation. The adjustment of the parameters of the model has permitted to describe quantitatively the experimental results. This description of the irradiation microstructure has been coupled with a Frank loops hardening model which has permitted to describe the observed hardening. The range of explored doses goes up to 40 dpa and is representative of the irradiation dose corresponding to the half life of the reactors design. (J.S.)

  11. Mechanistic understanding of irradiation corrosion of zirconium alloys in nuclear power plants: stimuli, status and outlook

    International Nuclear Information System (INIS)

    Extensive information about the corrosion behaviour of zirconium alloys under irradiation is presented. Review of the existing models of radiation corrosion is given. An accent is made on a necessity in conducting basic investigations to overcome contradictions in interpreting the experimental data available. Importance of solving the problem of zirconium alloy corrosion for safe NPP operation is underlined. 34 refs.; 6 figs.; 4 tabs

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

    Science.gov (United States)

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

    2015-10-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

  14. Evaluation of High-Temperature Tensile Property of Diffusion Bond of Austenitic Alloys for S-CO2 Cycle Heat Exchangers

    International Nuclear Information System (INIS)

    To improve the inherent safety of the sodium-cooled fast reactor (SFR), the supercritical CO2 (S-CO2) Brayton cycle is being considered as an alternative power conversion system to steam the Rankine cycle. In the S-CO2 system, a PCHE (printed circuit heat exchanger) is being considered. In this type of heat exchangers, diffusion bonding is used for joining the thin plates. In this study, the diffusion bonding characteristics of various austenitic alloys were evaluated. The tensile properties were measured at temperatures starting from the room temperature up to 650℃. For the 316H and 347H types of stainless steel, the tensile ductility was well maintained up to 550℃. However, the Incoloy 800HT showed lower strength and ductility at all temperatures. The microstructure near the bond line was examined to understand the reason for the loss of ductility at high temperatures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-14

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

  16. Effect of water depth on the underwater wet welding of ferritic steels using austenitic Ni-based alloy electrodes.

    OpenAIRE

    Sheakley, Brian J.

    2000-01-01

    Underwater welding using shielded metal arc welding (SMAW) on US naval Vessels is very attractive because of the ability to effect repairs without costly dry dock expenses. In the past the primary problems with underwater wet weldments on steels utilizing SMAW with ferritic electrodes, were underbead cracking in the heat affected zone (HAZ), slag inclusions, oxide inclusions, and porosity. To avoid underbead cracking three weld samples were made using an austenitic nickel weld metal with an O...

  17. Structural defects in Fe–Pd-based ferromagnetic shape memory alloys: tuning transformation properties by ion irradiation and severe plastic deformation

    International Nuclear Information System (INIS)

    Fe–Pd-based ferromagnetic shape memory alloys constitute an exciting class of magnetically switchable smart materials that reveal excellent mechanical properties and biocompatibility. However, their application is severely hampered by a lack of understanding of the physics at the atomic scale. A many-body potential is presented that matched ab inito calculations and can account for the energetics of martensite ↔ austenite transition along the Bain path and relative phase stabilities in the ordered and disordered phases of Fe–Pd. Employed in massively parallel classical molecular dynamics simulations, the impact of order/disorder, point defects and severe plastic deformation in the presence of single- and polycrystalline microstructures are explored as a function of temperature. The model predictions are in agreement with experiments on phase changes induced by ion irradiation, cold rolling and hammering, which are also presented. (paper)

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

    Science.gov (United States)

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

    2013-07-01

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

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

    Science.gov (United States)

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

    2015-10-01

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

  20. Microstructural changes in a neutron-irradiated Fe–6 at.%Cr alloy

    Energy Technology Data Exchange (ETDEWEB)

    Bachhav, Mukesh; Yao, Lan [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Robert Odette, G. [Materials Department, University of California, Santa Barbara, CA 93106 (United States); Marquis, Emmanuelle A., E-mail: emarq@umich.edu [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-10-15

    The microstructural and chemical changes in a Fe–6 at.%Cr binary model alloy neutron irradiated to 1.82 dpa at 290 °C were investigated using atom probe tomography. After irradiation, Si and Cr are found segregated to dislocation loops, which were analyzed in terms of number density, size, and habit plane. Grain boundary chemistry was quantitatively compared between the as-received and the neutron irradiated alloys. The results are discussed in the context of equilibrium segregation, radiation-enhanced diffusion, and/or radiation induced segregation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Shuoxue [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); He, Xinfu [China Institute of Atomic Energy, Beijing 102413 (China); Li, Tiecheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ma, Shuli; Tang, Rui [Nuclear Power Institute of China, Chengdu 610041 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China)

    2012-10-15

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

  4. Irradiation-induced precipitation and solute segregation in alloys. Fourth annual progress report, February 1, 1981-March 31, 1982

    International Nuclear Information System (INIS)

    The studies of irradiation-induced solute segregation (IISS) and irradiation-induced precipitation (IIP) in Ni-Si and Pd-Fe alloys have been completed. Progress is reported for several other projects: irradiation damage in binary Pd-Cr, -Mn and -V alloys (15 at. %); IIP in Pd-Mo and Pd-W alloys; IIP in Pd-25 at. % Cr alloy; and irradiation damage effects in proton-bombarded metallic glasses (Ni-65 Zr, 40 Fe 40 Ni 14 P6B). 27 figures

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

    Energy Technology Data Exchange (ETDEWEB)

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-27

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

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

    Science.gov (United States)

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

    2011-07-01

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

  8. Elimination of casting heterogeneities by high temperature heat treatment on a titanium stabilized austenitic alloy. Effect on the microstructure

    International Nuclear Information System (INIS)

    Microstructural observation on a longitudinal section of stainless steels often reveals the presence of a ''veined'' structure showing a segregation remainder due to the setting of the ingot. This casting heterogeneity can be eliminated by high temperature treatments. This study shows the change in the structure and the state of solubilization produced by these high temperature treatments and the effect of a stabilizing element such as titanium on Z6CNDT17.13 and Z10CNDT15.15B alloys compared with the Z6CND17.13 alloy. It is also shown that a high temperature treatment applied to these stabilized alloys deeply modifies the recrystallization kinetics

  9. The electrochemical corrosion behavior of austenitic alloys, cobalt or nickel based super alloys, structurally hardened martensitic, Inconel, zircaloy, super austenitic, duplex and of Ni-Cr or NTi deposits in tritiated water. 3 volumes

    International Nuclear Information System (INIS)

    The redox potential of 3 H2O, as well as the corrosion potentials in this medium are found, abnormally, in the trans-passive region. This is completely different from the behavior in the chemical industry or in the water in nuclear powers. With such behavior, there will be breakdowns of the protective oxide layers, and in the presence of chloride there will be immediate pitting. The steels that are most resistant to this behavior are the super austenitic and super Duplex. To avoid corrosion, another solution is to decompose the radiolytic products by imposing a slight reducing potential. Corrosion inhibitors, which are stable in tritiated water, can be used. (author). 69 refs., 421 figs., tabs

  10. Role of alloyed molybdenum on corrosion resistance of austenitic Ni–Cr–Mo–Fe alloys in H2S–Cl– environments

    International Nuclear Information System (INIS)

    Highlights: • The alloyed molybdenum improves corrosion resistance in the H2S–Cl– environment. • The formed surface film comprises sulfide including molybdenum and chromium oxide. • The Ni–Mo–Fe alloy shows good corrosion resistance in the H2S–Cl– environment. • It is revealed that molybdenum sulfide is stable and cation selective. • A possible role of alloyed molybdenum is proposed. - Abstract: Corrosion test and surface analysis were conducted in the H2S–Cl– environments to elucidate the role of alloyed molybdenum on the corrosion resistance of Ni–Cr–Mo–Fe alloys. The alloyed molybdenum improves the localized corrosion resistance. The surface film is of double layers which comprise sulfide including molybdenum and chromium oxide. However, the Ni–Mo–Fe alloy also shows good corrosion resistance in the H2S–Cl– environment. This good corrosion resistance is considered to be due to the cation selectivity of molybdenum sulfide, which can form in such environments. The role of alloyed molybdenum on the corrosion resistance of Ni–Cr–Mo–Fe alloys in H2S–Cl– environments is proposed

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Science.gov (United States)

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

    2016-08-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  14. Statistical thermodynamics and mean-field theory for the alloy under irradiation model

    International Nuclear Information System (INIS)

    A generalization of statistical thermodynamics to the open systems case, is discussed, using as an example the alloy-under-irradiation model. The statistical properties of stationary states are described with the use of generalized thermodynamic potentials and 'quasi-interactions' determined from the master equation for micro-configuration probabilities. Methods for resolving this equation are illustrated by the mean-field type calculations of correlators, thermodynamic potentials and phase diagrams for disordered alloys

  15. Heavy ion irradiation effects in Zr excel alloy pressure tube material

    International Nuclear Information System (INIS)

    Zirconium Excel alloy (Zr-3.5wt.%Sn-0.8%Nb-0.8%Mo) is the candidate material for pressure tubes in the Generation-IV CANDU® Super Critical Water-cooled Reactor (SCWR) design. Changes in microstructure induced by neutron irradiation are known to have important consequences on the in-reactor deformation behavior. The in-situ ion irradiation technique has been employed to elucidate the irradiation damage in dual phase Zr-excel alloy (~60% hcp alpha and ~40% bcc beta). 1 MeV Kr ion irradiation experiments were conducted at different temperatures ranging from 100oC-400oC. Damage microstructures have been characterized by Transmission Electron Microscopy in both the alpha and beta phases at different temperatures after a maximum dose of 10 dpa. Several new observations including irradiation induced omega (ω) phase precipitation have been reported. The ω/β orientation relationship was determined by the detailed analysis of selected area diffraction patterns. In-situ irradiation provided an opportunity to observe the nucleation and growth of basal plane c-component loops. It has been shown that under Kr ion irradiation the c-loops start to nucleate and grow above a threshold dose, as has been observed for neutron irradiation. Furthermore, the role of temperature, material composition and pre-irradiation microstructure has been discussed in detail. (author)

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

    Science.gov (United States)

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

    2013-08-01

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

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

    Science.gov (United States)

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

    2013-08-01

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

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

    Science.gov (United States)

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

    1996-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Christine Eckstein; Heinz- Joachim Spies; Jochen Albrecht

    2004-01-01

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

  20. Effects of irradiation on properties of refractory alloys with emphasis on space power reactor applications

    International Nuclear Information System (INIS)

    The probable effects of irradiation on niobium and tungsten alloys in use as components of thermionic convertors in a space reactor were reviewed by the author in 1971. While considerably more data on refractory metals have been generated since that time, the data have not been reviewed with respect to space reactor applications. This paper attempts such a review. The approach used is to work from the most recently available review of irradiation effects for each alloy system (where such a review is available) and to discuss that review and more recent data judged to be the most useful in establishing likely behavior in high-temperature reactor service. 28 figures, 6 tables

  1. Characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

  2. Microstructure and mechanical properties of medium energy (600-800 MeV) proton irradiated commercial aluminium alloys

    International Nuclear Information System (INIS)

    Commercial AlMg- and AlMgSi-alloys were irradiated with medium energy (600-800 MeV) protons to a nominal fluence of 3.2 x 1024 p/m2 which yields by calculation a displacement damage of 0.2 dpa and helium and hydrogen generation of 67 and 275 appm, respectively. Post-irradiation tensile testing revealed a very marked degree of irradiation-induced softening in the cold-worked AlMg-alloy as well as in the precipitation-hardened AlMgSi-alloy. The TEM examination of the irradiated specimens showed that neither the cold-work microstructure in the AlMg-alloy nor the G.P. zone type precipitates in the AlMgSi-alloy survive under the irradiation conditions used in the present experiment. Results of complimentary investigations (i.e., hardness measurements, optical microscopy and SEM-fractography) are also presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  6. Processing of Refractory Metal Alloys for JOYO Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    RF Luther; ME Petrichek

    2006-02-21

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

  7. Processing of Refractory Metal Alloys for JOYO Irradiations

    International Nuclear Information System (INIS)

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

  8. Effect of rare earth alloying on creep rupture of economical 21Cr-11Ni-N heat-resistant austenitic steel at 650 °C

    Institute of Scientific and Technical Information of China (English)

    陈雷; 龙红军; 刘鑫刚; 金淼; 马筱聪

    2016-01-01

    The effect of rare earth (RE) on creep rupture of economical 21Cr-11Ni-N heat-resistant austenitic steel was investigated at 650 °C under different stress levels. It was found that RE could increase the time to creep rupture, especially at long-term creep dura-tion. The logarithm of the time to creep rupture (lgtr) was a linear function of the applied stress (σ). RE addition was favorable to gen-erating a high fraction of low-coincidence site lattice (CSL) boundaries which was a possible cause for improving the creep rupture resistance. The fracture surface of RE-added steel exhibited less intergranular cracks suggesting the alteration on the nature of grain boundaries due to the presence of RE. RE addition changed the morphology of the intergranular chromium carbides from continuous network shape to fragmentary distribution which was another cause for longer creep duration. These results strongly suggested that the effect of RE alloying played a crucial role in improving the creep rupture resistance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  10. Irradiation effects in Fe-30%Ni alloy during Ar ion implantation

    International Nuclear Information System (INIS)

    The use of metallic thin films for studying the processes which take place during ion irradiation has recently increased. For example, ion implantation is widely used to study the structural defects in transition metallic thin films such as (Fe, Ni, Co), because it can simulate the effects occurring in nuclear reactors during neutron irradiation especially the swelling of reactor materials. The swelling of metals and alloys is strongly related to the material structure and to the irradiation conditions. The general feature of formation of structural defects as a function of irradiation dosage and annealing temperature is well known. However, the detailed mechanisms are still not well understood. For example, the swelling of iron alloy with 30-35% nickel is very small in comparison with other Ni concentrations, and there is no clear information on the possibility of phase transitions in fe-Ni alloys during irradiation. The aim of this work is to study the phase-structural changes in Fe-30% Ni implanted by high dose of argon ions. The effect of irradiation with low energy argon ions (40 KeV, and fluences of 10.E15 to 10.E17 ions/cm) on the deposited thin films of Fe-30% Ni alloy was investigated using RBS and TEM techniques. The thicknesses of these films were about 65+-10 nm deposited on ceramic, KBr, and Be fiols substrates. Gas bubble formation and profile distribution of the implanted argon ions were investigated. Formation of an ordered phase Fe3 Ni during irradiation appears to inhibit gas bubble formations in the film structure. (author). 17 refs., 15 figs., 7 tabs

  11. Irradiation creep in path A alloys irradiated to 5 dpa in the ORR-MFE-4B spectral tailoring experiment at 500 and 6000C

    International Nuclear Information System (INIS)

    The experiment will determine irradiation creep in an environment that produces helium with the He:dpa ratio characteristic of a fusion reactor. Pressurized tubes of 20%-cold-worked type 316 stainless steel and 25%-cold-worked Prime Candidate Alloy (PCA) were irradiated at 500 and 6000C in the Oak Ridge Research Reactor (ORR) spectral tailoring experiment to 5.1 dpa. Diametral measurements were made to determine irradiation creep rates. Both alloys behaved rather similarly but exhibited lower creep rates than did the Fast Flux Test Facility (FFTF) first core type 316 stainless steel irradiated in EBR-II

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  13. Ion irradiation testing and characterization of FeCrAl candidate alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-29

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. Effects of neutron irradiation in magnetic properties of metals and alloys

    International Nuclear Information System (INIS)

    The effects of neutron irradiation on the magnetic properties of metals and alloys, namely magnetic anisotropy, hysteresis loop, initial magnetic permeability, which are sensitives to structural changes, are studied. First a short review is made, followed by experimentals results and the plot of the vacancies supersaturation, which are obtained in the reactor of the Instituto de Pesquisas Energeticas e Nucleares. (Author)

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

    Science.gov (United States)

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

    2002-08-01

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

  17. The effect of neutron irradiation on the electrical resistivity of high-strength copper alloys

    International Nuclear Information System (INIS)

    The effects of neutron irradiation on the electrical resistivity of precipitation hardened (PH) and dispersion strengthened (DS) copper alloys are discussed. The analysis is based on the experimental study of radiation damage of PH and DS copper alloys, irradiated in the fast neutron reactor BOR-60 with doses of 8-16 x 1025 n/m2 and in the mixed spectrum neutron reactor SM-2 with doses of 3.7-5.5 x 1025 n/m2. The experimental data on the change Δρ in electrical resistivity of DS-type copper alloys irradiated in the BOR-60 reactor show that irradiation to 7-10 dpa at T=340-450 C causes a drop in electrical conductivity by not more than 20%. The obtained results show that in mixed-spectrum reactors the rate of Δρ normalized to the dpa is about 20 times as high as in fast neutron reactors. The conclusion is made that the calculations performed for ITER must take into account the presence of appreciable fluxes of thermal neutrons in certain components of the reactor. The latter will play a decisive role in the drop in thermal conductivity of copper alloys in these components. (orig.)

  18. Dislocation Climb Sources Activated by 1 MeV Electron Irradiation of Copper-Nickel Alloys

    DEFF Research Database (Denmark)

    Barlow, P.; Leffers, Torben

    1977-01-01

    irradiation temperatures corresponding to the highest source densities is approximately 350°–500°C. The climb sources are not related to any pre-existing dislocations resolved in the microscope. The sources emit three types of loop: ‘rectangular’ loops with a100 Burgers vector and {100} habit plane, normal...... prismatic loops with Burgers vector a/2110, and Frank loops. There is no significant difference between the apparent activation energy for growth of the three types of loops. The source points are suggested to be submicroscopic nickel precipitates-with reference to the existing evidence that......Climb sources emitting dislocation loops are observed in Cu-Ni alloys during irradiation with 1 MeV electrons in a high voltage electron microscope. High source densities are found in alloys containing 5, 10 and 20% Ni, but sources are also observed in alloys containing 1 and 2% Ni. The range of...

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

    CERN Document Server

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

    2003-01-01

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

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

    Science.gov (United States)

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

    1979-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  2. Tensile properties of vanadium alloys irradiated at 390 degrees C in EBR-II

    International Nuclear Information System (INIS)

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

  3. Response of unirradiated and neutron-irradiated vanadium alloys to Charpy-impact loading

    International Nuclear Information System (INIS)

    The ductile-brittle transition temperature (DBTT) was determined by Charpy-impact impact tests for dehydrogenated (<30 appm H) and hydrogenated (400--1200 appm H) V-7.2Cr-14.5Ti, V-9.9Cr-9.2Ti, V-13.5Cr-5.2Ti, V-17.7Ti, V-9.2Cr-4.9Ti, V-9.0Cr-3.2Fe-1.2Zr, V-3.1Ti-0.5Si, V-4.1Cr-4.3Ti, V-4.6Ti, and V-2.5Ti-1.0Si alloys. The DBTT was also determined for the V-13.5Cr-5.2Ti, V-9.2Cr-4.9Ti, V-7.2Cr-14.5Ti, and V-17.7Ti alloys after neutron irradiation at 420 and 600 degrees C to 41--44 atom displacements per atom. The DBTTs determined for these vanadium alloys show that a vanadium alloy containing Cr and/or Ti and Si alloying additions to be used as a structural material in a fusion reactor should contain 3--11 wt % total alloying addition for maximum resistance to hydrogen- and/or irradiation-induced embrittlement. 4 refs., 3 figs., 2 tabs

  4. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  5. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

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

    OpenAIRE

    Białobrzeska B.; Dudziński W.

    2015-01-01

    Today low-alloy steels with boron achieve high resistance to abrasive wear and high strength. These features are obtained by using advanced technology of manufacturing. This makes boron steels increasingly popular and their application more diverse. Application of these steels can extend the lifetime of very expensive machine construction in many industries such as mining, the automotive, and agriculture industries. An interesting subgroup of these materials is steel with boron intended for h...

  7. Clustering of point defects under electron irradiation in dilute iron alloys and an iron manganese nickel alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hardouin-Duparc, A.; Barbu, A. [CEA-CEREM-DECM, Palaiseau (France)

    1997-11-01

    In low copper steels for nuclear reactor pressure vessel, point defect clustering seems to play an important role in hardening. In order to study the hardening component which results from the clustering of freely migrating point defects, the authors irradiated, in a high voltage electron microscope, Fe, the alloys Fe0.13% Cu and Fe0.014%P, alloys and the alloy Fe1.5%Mn0.8%Ni0.1%Cu0.01%P, the composition of which is close to the matrix of pressure vessel steels. They studied the nucleation of dislocation loops and their growth velocity. They find out that copper and phosphorus have no effect on the vacancy migration energy but that this parameter decreased significantly in the complex alloy. The main point is certainly that loops are nucleated in the complex model alloy up to 500 C while no loop appears above 300 C in Fe and in FeCu. FeP shows an intermediate behavior.

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

    Directory of Open Access Journals (Sweden)

    Buxiang Zheng

    2014-02-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

  10. Effect of high temperature annealing on ferromagnetism induced by energetic ion irradiation in FeRh alloy

    Energy Technology Data Exchange (ETDEWEB)

    Kosugi, S.; Fujita, Nao; Matsui, T.; Hori, F. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Saitoh, Y. [Japan Atomic Energy Agency (JAEA-Takasaki), Takasaki, Gunma 370-1292 (Japan); Ishikawa, N.; Okamoto, Y. [Japan Atomic Energy Agency (JAEA-Tokai), Tokai, Ibaraki 319-1195 (Japan); Iwase, A., E-mail: iwase@mtr.osakafu-u.ac.j [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan)

    2011-05-01

    Effects of thermal annealing on ion-irradiation induced ferromagnetism of Fe-50at.%Rh bulk alloy and the related structural change were investigated by means of superconducting quantum interference device (SQUID) and extended X-ray absorption fine structure (EXAFS), respectively. Depending on the annealing temperature from 100 to 500 {sup o}C, the magnetization induced by 10 MeV iodine ion irradiation and the lattice structure of the alloy were remarkably changed. After 500 {sup o}C annealing, the magnetization and the lattice ordering of the alloy become similar to the states before the irradiation. The experimental result indicates that the thermal relaxation of irradiation-induced atomic disordering dominates the magnetic state of ion-irradiated Fe-50at.% Rh alloy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    Science.gov (United States)

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

    2010-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2007-01-01

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

  18. Order-disorder transformations, vacancy behaviors and damage recovery in electron-irradiated Ni- and Fe-base alloys

    International Nuclear Information System (INIS)

    Full text : In this study the effect of irradiation by 2 MeV electrons on order-disorder phase transformation characteristics of Ni3Fe alloys and the behavior of vacancies and damage recovery after irradiation in a 316 stainless steel and Ni3Fe ordered alloys have been investigated by using positron annihilation and differential scanning calorimetry techniques. It is well-known that irradiation of metallic alloys by high-energy electrons, neutrons and heavy ions can have profound effects on the formation or dissolution of phases by alteration of the stability of these phases

  19. Evaluation of radiation hardening in ion-irradiated Fe based alloys by nanoindentation

    International Nuclear Information System (INIS)

    Nanoindentation in combination with ion irradiation offers the possibility to quantify irradiation hardening due to radiation damage. Irradiation experiments for Fe–1.0wt.%Cu alloys, China A508-3 steels, and 16MND5 steels were carried out at about 100 °C by proton and Fe-ions with the energy of 240 keV, 3 MeV respectively. The constant stiffness measurement (CSM) with a diamond Berkovich indenter was used to obtain the depth profile of hardness. The results showed that under 240 keV proton irradiation (peak damage up to 0.5 dpa), Fe–1.0wt.%Cu alloys exhibited the largest hardening (∼55%), 16MND5 steels resided in medium hardening (∼46%), and China A508-3(2) steels had the least hardening (∼10%). Under 3 MeV Fe ions irradiation (peak damage up to 1.37 dpa), both China A508-3(1) and 16MND5 steels showed the same hardening (∼26%). The sequence of irradiation tolerance for these materials is China A508-3(2) > 16MND5 ≈ China A508-3(1) > Fe–1.0wt.%Cu. Based on the determination of the transition depth, the nominal hardness H0irr was also calculated by Kasada method

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  1. Effects of neutron irradiation on structural stability and mechanical properties of copper alloys for ITER divertor

    International Nuclear Information System (INIS)

    The effects of neutron irradiation on fine structure and mechanical properties of the precipitation-hardened alloy Cu-Cr-Zr-Mg and dispersion-strengthened alloys Cu-Mo and MAGT-0.2 have been investigated. Irradiation were performed in the BOR-60 (Tirr = 420deg C, 1.5 x 1022 n/cm2), SM-2 (Tirr = 300deg C, 1021 n/cm2) and WWR (Tirr = 80deg C, 9.3 x 1019 n/cm2) reactors. It was found that the precipitation-hardened alloy Cu-Cr-Zr-Mg exhibits high resistance to swelling. At Tirr ≅ 420deg C, however, polygonization and precipitate coarsening processes go on at a high rate, which results in a drastic decrease in strength. Dispersion-strengthened alloy Cu-Mo and MAGT-0.2 exhibit high resistance to radiation swelling and have high stability of strengthening structure during neutron irradiation over the temperature and dose range corresponding to ITER physics and technological stages. (orig.)

  2. Helium effects on irradiation dmage in V alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel. The materials included vitange type ASTM A302 Grade B (A302B) plates and welds containing different nickel (Ni) and copper (Cu) concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with 'superclean' composition (extremely low phosphorus, sulfur, manganese and silicon). To determine irradiation damage behavior, all materials were irradiated at several different irradiation damage levels ranging from 0.0003 dpa to 0.06 dpa at an irradiation damage levels ranging from 0.003 dpa to 0.06 dpa at an irradiation temperature of about 232 degrees C (450 degrees F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine the transition temperature at 41J (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. The irradiation damage behavior was measured by the shift in the Charpy 41J or 47J transition temperature (ΔTT41J or ΔTT47J) and lowering of the upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior. The highest irradiation damage (greatest ΔTT) was found in an A302B type weld contaiNing 1.28% Ni and 0.20% CU while the least irradiation damage was found in the 3.5% Ni, 0.05% Cu, superclean wrought materials

  4. Microstructure analysis of magnesium alloy melted by laser irradiation

    Science.gov (United States)

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

    2005-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Science.gov (United States)

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

    2014-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  9. Irradiation Embritlement in Alloy HT-­9

    Energy Technology Data Exchange (ETDEWEB)

    Serrano De Caro, Magdalena [Los Alamos National Laboratory

    2012-08-27

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

  10. Austenitic structure formation in an Fe-32% Ni alloy during slow heating in the critical temperature range

    Science.gov (United States)

    Zemtsova, N. D.

    2014-08-01

    Electron diffraction is used to show (for the first time) that the reverse α → γ transformation in an Fe-32% Ni during slow heating develops via the formation of an intermediate paramagnetic 9 R phase. Coarse extended lamellae form according to a shear mechanism in the central part of the temperature range of the reverse transformation, which is called the critical range (here, the physical properties of the alloy change anomalously). The extended lamellae consist of 9 R-phase lamellae with γ-phase interlayers. A high density of periodic stacking faults in the structure of the 9 R phase and a high density of chaotic stacking faults in the complex 9 R + γ phase determine the nature of phase transformation-induced hardening.

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

  12. Alloy development for irradiation performance. Quarterly progress report for period ending March 31, 1981

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

    International Nuclear Information System (INIS)

    Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

  15. On effect of second-plasma precipitates on void formation and growth in irradiated alloys

    International Nuclear Information System (INIS)

    Effects of coherent and incoherent precipitates on point defect concentrations, recombination rate, and growth of vacancy voids in irradiated materials were considered. A new mechanism of defect loss by enhanced recombination inside coherent precipitates was described. Vacancy swelling suppression, based on the recombination mechanism of point defect loss to coherent precipitates, was shown to be efficient in a wide range of irradiation conditions - from heavy ion irradiation to in-reactor irradiation. Point defect fluxes arising in the vicinity of the coherent precipitate due to difference between recombination rates in the precipitate and matrix, results in segregation fluxes of alloying elements. This effect, being an analog of the inverse Kirkendall effect, influences stability of the coherent precipitates. Modification of chemical composition of coherent precipitates - phenomenon observed experimentally - is likely to be caused by the recombination-driven segregation, followed by infiltration of segregating elements into the precipitates. 39 refs.; 7 figs.; 1 table. (author)

  16. Study of the formation of solute clusters under irradiation in model ferritic alloys

    International Nuclear Information System (INIS)

    Neutron irradiation results in the formation of a high number density (1023 to 1024 m-3) of ultrafine (2 nm in diameter) solute clusters in reactor vessel steels. These clusters contain a supersaturated element (copper), and some others solutes (Mn, Ni, Si and P) soluble at the temperature of irradiation (300 C). The aim of the work described in this report is to understand what are the basic processes at the origin of the formation of these clusters, and to obtain information about the effect of the different solutes. The microstructure of model alloys, after different irradiation experiments is characterised by atom probe. The comparison between experimental results and results obtained by mean field modelling (evolution of point defects under irradiation) suggests that the precipitation of the solute clusters is heterogeneous, on point defects clusters. Precipitation kinetic is slowed down by solutes other than copper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-22

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

  1. Ultrasonic irradiation and its application for improving the corrosion resistance of phosphate coatings on aluminum alloys.

    Science.gov (United States)

    Sheng, Minqi; Wang, Chao; Zhong, Qingdong; Wei, Yinyin; Wang, Yi

    2010-01-01

    In this paper, ultrasonic irradiation was utilized for improving the corrosion resistance of phosphate coatings on aluminum alloys. The chemical composition and morphology of the coatings were analyzed by X-ray diffraction analysis (XRD) and scanning electron microscopy (SEM). The effect of ultrasonic irradiation on the corrosion resistance of phosphate coatings was investigated by polarization curves and electrochemical impedance spectroscopy (EIS). Various effects of the addition of Nd(2)O(3) in phosphating bath on the performance of the coatings were also investigated. Results show that the composition of phosphate coating were Zn(3)(PO(4))(2).4H(2)O(hopeite) and Zn crystals. The phosphate coatings became denser with fewer microscopic holes by utilizing ultrasonic irradiation treatment. The addition of Nd(2)O(3) reduced the crystallinity of the coatings, with the additional result that the crystallites were increasingly nubby and spherical. The corrosion resistance of the coatings was also significantly improved by ultrasonic irradiation treatment; both the anodic and cathodic processes of corrosion taking place on the aluminum alloy substrate were suppressed consequently. In addition, the electrochemical impedance of the coatings was also increased by utilizing ultrasonic irradiation treatment compared with traditional treatment. PMID:19692286

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    Science.gov (United States)

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

    2011-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kuksenko, V. [Groupe de Physique des Materiaux, Universite et INSA de Rouen, UMR 6634 CNRS Avenue de l' Universite, BP 12, 76801 Saint Etienne du Rouvray (France); Pareige, C., E-mail: cristelle.pareige@univ-rouen.fr [Groupe de Physique des Materiaux, Universite et INSA de Rouen, UMR 6634 CNRS Avenue de l' Universite, BP 12, 76801 Saint Etienne du Rouvray (France); Genevois, C.; Cuvilly, F.; Roussel, M.; Pareige, P. [Groupe de Physique des Materiaux, Universite et INSA de Rouen, UMR 6634 CNRS Avenue de l' Universite, BP 12, 76801 Saint Etienne du Rouvray (France)

    2011-08-01

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

  5. Dry sliding tribological behavior of AZ31 magnesium alloy irradiated by high-intensity pulsed ion beam

    International Nuclear Information System (INIS)

    The dry sliding tribological behavior of AZ31 magnesium alloy irradiated by high-intensity pulsed ion beam (HIPIB) at energy density of 3.4 J/cm2 with 10 shots is investigated by dry sliding wear tests in order to explore the effect of HIPIB irradiation on tribological property of magnesium alloy. Surface morphologies, composition and structure of the irradiated AZ31 magnesium alloys are examined by electron probe microanalysis (EPMA) and X-ray diffraction (XRD). The results indicated that HIPIB irradiation led to the increase in surface microhardness and the reduction in friction coefficient and wear rate. Wear rate for both the original and the irradiated samples increased with increasing sliding load from 0.1 to 0.5 N. The transition from severe metallic wear to mild oxidative wear induced by HIPIB irradiation was observed by a combined analysis in surface morphology and chemical composition of wear tracks, mechanically mixed materials and wear debris, which is mainly attributed to the significant increase in microhardness resulting from grain refinement on the irradiated surface. In addition, defects induced by HIPIB irradiation promoted the diffusion of oxygen during sliding wear and therefore led to the formation of compact mixed materials and protective films on the wear tracks surface, which also contributes to the transition in wear mechanism of AZ31 magnesium alloy induced by HIPIB irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  7. Previsions of the microstructural evolution of ferritic alloys under irradiation by numerical atomic scale simulations

    International Nuclear Information System (INIS)

    In this work, we have improved a diffusion model for point defects (vacancies and self-interstitials) by introducing hetero-interstitials. The model has been used to simulate by Kinetic Monte Carlo (KMC) the formation of solute rich clusters that are observed experimentally in irradiated ferritic model alloys of type Fe - CuMnNiSiP - C.Electronic structure calculations have been used to characterize the interactions between self-interstitials and all solute atoms, and also carbon. P interacts with vacancies and strongly with self-interstitials. Mn also interacts with self-interstitials to form mixed dumbbells. C, with occupies octahedral sites, interacts strongly with vacancies and less with self-interstitials. Binding and migration energies, as well as others atomic scale properties, obtained by ab initio calculations, have been used as parameters for the KMC code. Firstly, these parameters have been optimized over isochronal annealing experiments, in the literature, of binary alloys that have been electron-irradiated. Isochronal annealing simulations, by reproducing experimental results, have allowed us to link each mechanism to a single evolution of the resistivity during annealing. Moreover, solubility limits of all the elements have been determined by Metropolis Monte Carlo. Secondly, we have simulated the evolution at 300 C of the microstructure under irradiation of different alloys of increasing complexity: pure Fe, binary alloys, ternaries, quaternaries, and finally complex alloys which compositions are close to those of pressure vessel steels. The results show that the model globally reproduces all the experimental tendencies, what has led us to propose mechanisms to explain the behaviours observed. (author)

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

    Institute of Scientific and Technical Information of China (English)

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

    2013-01-01

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

  9. Alloy development for irradiation performance. Semiannual progress report for period ending September 30, 1985

    International Nuclear Information System (INIS)

    This report is the twenty-second in a series of Technical Progress Reports on ''Alloy Development for Irradiation Performance'' (ADIP), which is one element of the Fusion Reactor Materials Program, conducted in support of the magnetic Fusion Energy Program of the US Department of energy. This report is organized along topical lines with Chapters 3 through 8 devoted to the various alloy classes that are currently under investigation. Thus the work of a given laboratory may appear at several different places in the report. The materials compatibility and environmental effects work on all alloy classes is collected together in Chapter 9. The Table of Contents is annotated for the convenience of the reader

  10. Microstructure modelling of ferritic alloys under high flux 1 MeV electron irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hardouin Duparc, A.; Moingeon, C.; Smetniansky-de-Grande, N.; Barbu, A. E-mail: alain.barbu@poly.polytechnique.fr

    2002-04-01

    The point-defect clustering is an important component of the hardening of low copper content pressure vessel steels. This study reports the first steps of a project devoted to the modelling of the nucleation and growth of point-defect clusters in ferritic alloys under irradiation at large fluence. A cluster-dynamics modelling based on rate equations giving the evolution of the population of interstitial loops up to some 0.1 {mu}m and of vacancy clusters is developed. It is applied to two alloys FeCu (0.13 wt%) and FeMn(1.5 wt%)Ni(0.8 wt%)Cu(0.13 wt%)P(0.01 wt%) the composition of which is close to the one of pressure vessel steels and to non-alloyed Fe for comparison. The model was calibrated by carrying out 1 MeV irradiations in a high voltage microscope on these three materials and by using the results of experiments and atomic simulations reported in the literature. It is shown that the presence of copper in iron stabilises the interstitial clusters and more important that the parameters relative to the interstitials in the complex alloys are totally different from those for iron: the migration energy must be increased from 0.3 to 1 eV and the binding energy of di-interstitials must be decrease from 0.9 to 0.2 eV.

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

    Science.gov (United States)

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

    2016-04-01

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

  12. Structure and phase transformations in WC-Co hard alloys irradiated with a low-flux electron beam

    International Nuclear Information System (INIS)

    The structure and phase composition in electron irradiated WC-Co hard alloys have been studied by X-ray diffraction analysis and electron microscopy methods. It is shown that the dose dependences of WC and Co lattice parameters are significantly different for the initial alloys and the electrolytically etched alloys, from the surface of which either cobalt or tungsten carbide was removed. Microstress level, size and volume of primary grains of WC were decreased under irradiation. It is assumed, the radiation-stimulated ordering-disordering transformation processes in tungsten carbide take place, and WC particles redistribution in Co matrix occurs

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

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

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

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

    Science.gov (United States)

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

    1996-11-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  16. Microstructural changes in a neutron-irradiated Fe–15 at.%Cr alloy

    Energy Technology Data Exchange (ETDEWEB)

    Bachhav, Mukesh [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Robert Odette, G. [Materials Department, University of California, Santa Barbara, CA 93106 (United States); Marquis, Emmanuelle A., E-mail: emarq@umich.edu [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-11-15

    Microstructural changes in a Fe–15 at.%Cr model alloy neutron irradiated to 1.82 dpa at 290 °C were characterized by atom probe tomography. Homogenously distributed α′ precipitates as well as fewer clusters containing Si, P, Ni, and Cr, were observed in the matrix. Grain boundary analyses before and after irradiation revealed segregation of Cr, with W-shape concentration profiles developing in the vicinity of grain boundary carbide and nitride particles. After irradiation, impurities such as C, Si and P were segregated to the grain boundaries. Zones depleted of α′ clusters, and Si were found at the interfaces of carbide and nitride precipitates and along grain boundaries in the vicinity of these precipitates.

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

    Science.gov (United States)

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

    2014-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  19. Experimental approach and micro-mechanical modeling of the creep behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    The fuel rod cladding, strongly affected by microstructural changes due to irradiation such as high density of dislocation loops, is strained by the end-of-life fuel rod internal pressure and the potential release of fission gases and helium during dry storage. Within the temperature range that is expected during dry interim storage, cladding undergoes long term creep under over-pressure. So, in order to have a predictive approach of the behavior of zirconium alloys cladding in dry storage conditions it is essential to take into account: initial dislocation loops, thermal annealing of loops and creep straining due to over pressure. Specific experiments and modelling for irradiated samples have been developed to improve our knowledge in that field. A Zr-1%Nb-O alloy was studied using fine microstructural investigations and mechanical testing. The observations conducted by transmission electron microscopy show that the high density of loops disappears during a heat treatment. The loop size becomes higher and higher while their density falls. The microhardness tests reveal that the fall of loop density leads to the softening of the irradiated material. During a creep test, both temperature and applied stress are responsible of the disappearance of loops. The loops could be swept by the activation of the basal slip system while the prism slip system is inhibited. Once deprived of loops, the creep properties of the irradiated materials are closed to the non irradiated state, a result whose consequence is a sudden acceleration of the creep rate. Finally, a micro-mechanical modeling based on microscopic deformation mechanisms taking into account experimental dislocation loop analyses and creep test, was used for a predictive approach by constructing a deformation mechanism map of the creep behavior of the irradiated material. (author)

  20. Effect of microstructure on radiation induced segregation and depletion in ion irradiated SS316 steel

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung Ha; Kwon, Sang Chul; Kwon, Jun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Irradiation assisted stress corrosion cracking (IASCC), void swelling and irradiation induced hardening are caused by change of characteristics of material by neutron irradiation, stress state of material and environmental situation. It has been known that chemical compositions varies at grain boundary (GB) significantly with fluence level and the depletion of Cr element at GB has been considered as one of important factors causing material degradation, especially, IASCC in austenitic stainless steel. However, experimental results of IASCC under PWR condition were directly not connected with Cr depletion phenomenon by neutron irradiation. Because the mechanism of IASCC under PWR has not yet been clearly understood in spite of many energetic researches, fundamental researches about radiation induced segregation and depletion in irradiated austenitic stainless steels have been attracted again. In this work, an effect of residual microstructure on radiation induced segregation and depletion of alloy elements at GB was investigated in ion irradiated SS316 steel using transmission electron microscope (TEM) with energy dispersive spectrometer (EDS)

  1. Investigation of vacancy-type defects in helium irradiated FeCrNi alloy by slow positron beam

    International Nuclear Information System (INIS)

    Graphical abstract: The variation of S parameter-incident positron energy profile for He ion irradiated Fe16.7Cr14.5Ni model alloy with different helium irradiation fluences at room temperature (RT) and 573 K, respectively. - Highlights: • He ions implanted into FeCrNi alloys with different fluence and temperature. • Large amount of vacancy-type defects formed after He ion irradiation. • He-vacancy complexes formed as helium atoms deposited in the material. • The diffusion mechanism of helium atoms might be changed at 573 K irradiation. - Abstract: The evolution of microstruture for Fe16.7Cr14.5Ni model alloy and 316 stainless steel irradiated with 140 keV He ions were studied by Positron annihilation spectroscopy. The fluences were 1 × 1016 and 5 × 1016 He ions/cm2. The irradiation temperature was room temperature and 573 K, respectively. The variation of S parameter-incident positron energy profile indicated that large amount of vacancy-type defects formed after He ion irradiation. Meanwhile, helium atoms deposited in bulk and certain amount of He-vacancy complexes were formed. The vacancy-type defects could be the major defects in track region and He-vacancy complexes would be the main defects in cascade region. The vacancy-type defects could migrate and aggregate to form vacancy clusters and even microvoids at elevated temperature irradiation. The diffusion mechanism of helium atoms might be changed at different irradiation temperature

  2. Experimental study and numerical modeling of the plastic behavior of zirconium alloys under and after irradiation

    International Nuclear Information System (INIS)

    Recrystallized zirconium alloys are widely used as constitutive material of cladding tubes in Pressurized Water Reactors. During their lifetime in reactor, these materials are submitted to irradiation, creating a large amount of defects and changing their mechanical behavior. Despite the broad knowledge of macroscopic modifications due to irradiation, microscopic mechanisms involved remain partially known and understood. This study aims at understanding this issue using two different means, experimental and numerical, to investigate interactions between moving dislocations and dislocation loops created by irradiation. The experimental approach is based on irradiating with Zr ions Zircaloy-4 samples. Then, these samples are strained in a transmission electron microscope (TEM). Mobile dislocations interacting with irradiation induced loops are observed, following different mechanisms. Loops can act as strong obstacles to moving dislocations, pinning their further glide and hardening the material. Therefore, this type of mechanism participates in irradiation hardening. Dislocations absorbing loops have also been observed, showing the ability of dislocations to clear up defects. This mechanism explains the formation of clear bands observed in the material after irradiation and mechanical testings. The numerical approach is based on Dislocation Dynamics (DD) simulations of mobile dislocations gliding in prismatic or basal planes of the hexagonal close packed lattice and loops, using NUMODIS. The results of this study are consistent with a recent study of interactions of dislocations in a prismatic plane and loops studied by molecular dynamics. The counterpart of this study with gliding dislocations in the basal plane, performed only using DD simulations, show interesting explanations of the observed clear band formation in basal and prismatic planes, with broader channels in basal planes. A situation observed during in situ TEM experiments has been simulated using DD

  3. Experimental evidence and thermodynamics analysis of high magnetic field effects on the austenite to ferrite transformation temperature in Fe-C-Mn alloys

    International Nuclear Information System (INIS)

    The non-isothermal decomposition of austenite into ferrite and pearlite in Fe-xC-1.5 wt.% Mn steels with x = 0.1, 0.2 and 0.3 wt.% C is investigated by in situ dilatometry and microstructure characterization in magnetic fields up to 16 T. The global shift towards higher temperatures of the respective austenite, ferrite + austenite and ferrite + pearlite stability regions is experimentally quantified. A systematic increase in the ferrite area fraction and proportional reduction of the Vickers hardness values with the magnetic field intensity are also reported. Moreover, the steels' magnetizations, measured up to 3.5 T and 1100 K, are used to calculate the magnetic contribution to the free energy of the transformation and to account thermodynamically for the field dependence of the transformation temperature. The impact of magnetic field is found to be greater with increasing carbon content in the steels.

  4. Experimental evidence and thermodynamics analysis of high magnetic field effects on the austenite to ferrite transformation temperature in Fe-C-Mn alloys

    Energy Technology Data Exchange (ETDEWEB)

    Garcin, T., E-mail: thomas.garcin@grenoble.cnrs.fr [CNRS/CRETA, 25 rue des martyrs BP166, 38042 Grenoble CEDEX 9 (France); Rivoirard, S. [CNRS/CRETA, 25 rue des martyrs BP166, 38042 Grenoble CEDEX 9 (France); Elgoyhen, C. [CRM Gent, Technologiepark 903c, B-9052 Zwijnaarde (Belgium); Beaugnon, E. [CNRS/CRETA, 25 rue des martyrs BP166, 38042 Grenoble CEDEX 9 (France)

    2010-04-15

    The non-isothermal decomposition of austenite into ferrite and pearlite in Fe-xC-1.5 wt.% Mn steels with x = 0.1, 0.2 and 0.3 wt.% C is investigated by in situ dilatometry and microstructure characterization in magnetic fields up to 16 T. The global shift towards higher temperatures of the respective austenite, ferrite + austenite and ferrite + pearlite stability regions is experimentally quantified. A systematic increase in the ferrite area fraction and proportional reduction of the Vickers hardness values with the magnetic field intensity are also reported. Moreover, the steels' magnetizations, measured up to 3.5 T and 1100 K, are used to calculate the magnetic contribution to the free energy of the transformation and to account thermodynamically for the field dependence of the transformation temperature. The impact of magnetic field is found to be greater with increasing carbon content in the steels.

  5. Austenitic stainless steels with cryogenic resistance

    International Nuclear Information System (INIS)

    The most used austenitic stainless steels are alloyed with chromium and nickel and have a reduced carbon content, usually lower than 0.1 % what ensures corresponding properties for processing by plastic deformation at welding, corrosion resistance in aggressive environment and toughness at low temperatures. Steels of this kind alloyed with manganese are also used to reduce the nickel content. By alloying with manganese which is a gammageneous element one ensures the stability of austenites. Being cheaper these steels may be used extensively for components and equipment used in cryogenics field. The best results were obtained with steels of second group, AMnNi, in which the designed chemical composition was achieved, i.e. the partial replacement of nickel by manganese ensured the toughness at cryogenic temperatures. If these steels are supplementary alloyed, their strength properties may increase to the detriment of plasticity and toughness, although the cryogenic character is preserved

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

    Samples of pure aluminium (99.9999%) and commercial Al-2.7%Mg (AlMg3) and Al-1.1%Mg-0.5%Si (Al6061) alloys were irradiated with 800 MeV protons at the Los Alamos Meson Physics Facility (LAMPF) at a temperature between 40-1000C to a maximum dose of 0.2 dpa. Transmission electron microscopy (TEM) showed a complete absence of voids or bubbles in the grain interiors of the aluminium and the aluminium-alloys. Bubbles were clearly visible by TEM at grain boundaries in pure Al and the AlMg3 alloy; but bubbles were not visible in the Al6061 alloy. The bubble density in the AlMg3 alloy was considerably higher than in pure Al. The amount of gas accumulation at grain boundaries was found to depend on gas generation rate, alloying and cold-work microstructure. (orig.)

  8. Mechanical Properties and Microstructure of Neutron Irradiated Cold-worked Al-1050 and Al-6063 Alloys

    International Nuclear Information System (INIS)

    The impact of neutron irradiation on the internal microstructure, mechanical properties and fracture morphology of cold-worked Al-1050 and Al-6063 alloys was studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens consisting of 50 mm long and 6 mm wide gauge sections, were punched out from Al-1050 and Al-6063 23% cold-worked tubes. They were exposed to prolonged neutron irradiation of up to 4.5x1025 and 8x1025 thermal neutrons/m2 (E -3 s-1. In general, the uniform and total elongation, the yield stress, and the ultimate tensile strength increase as functions of fluence. However, for Al-1050 a decrease in the ultimate tensile strength and yield stress was observed up to a fluence of 1x1025 thermal neutrons/m2 which then increase with thermal neutrons fluence. Metallographic examination and fractography for Al-6063 revealed a decrease in the local area reduction of the final fracture necking. This reduction is accompanied by a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. In contrast, for Al-1050, fracture morphology remains ductile transgranular shear rupture and the final local area reduction remains almost constant No voids could be observed in either alloy up to the maximum fluence. The dislocation density of cold-worked Al was found to decrease with the thermal neutron fluence. Prolonged annealing of unirradiated cold-worked Al-6063 at 52 degree led to similar results. Thus, it appears that, under our irradiation conditions, whereby the temperature encompassing the samples increases the exposure to this thermal field is the major factor influencing the mechanical properties and microstructure of aluminum alloys

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Directory of Open Access Journals (Sweden)

    Jie Gao

    2016-10-01

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

  12. Tunneling and migration of the dumbbell defect in electron-irradiated aluminum-zinc alloys

    International Nuclear Information System (INIS)

    Ultrasonic attenuation and velocity measurements on irradiated Al-Zn alloys (0.01, 0.1, and 0.5 at %) indicate a tunneling relaxation of the predominant mixed dumbbell defect at low temperatures, and mixed dumbbell migration at the Stage II anneal temperature. The effect of an internal strain varying with the zinc concentration on the measured decrement and modulus change is striking. Evaluated in the framework of a six-level system, this reveals the simultaneous action of resonance and nonclassical relaxation processes. Using Fe as a probe atom, it is shown that mixed dumbbell dissociation is in an insignificant component of the annealing of this defect. The decrease of the annealing temperature at higher zinc concentrations provides evidence that the mixed dumbbell migrates as a unit during annealing. The energies associated with dumbbell migration and interstitial escape are derived. Further evidence for the migration mechanism is obtained from successive irradiation and annealing

  13. Ion irradiation effects on high purity bcc Fe and model FeCr alloys

    International Nuclear Information System (INIS)

    FeCr binary alloys are a simple representative of the reduced activation ferritic/martensitic (F-M) steels, which are currently the most promising candidates as structural materials for the sodium cooled fast reactors (SFR) and future fusion systems. However, the impact of Cr on the evolution of the irradiated microstructure in these materials is not well understood in these materials. Moreover, particularly for fusion applications, the radiation damage scenario is expected to be complicated further by the presence of large quantities of He produced by the nuclear transmutation (∼ 10 appm He/dpa). Within this context, an elaborate ion irradiation study was performed at 500 C on a wide variety of high purity FeCr alloys (with Cr content ranging from ∼ 3 wt.% to 14 wt.%) and a bcc Fe, to probe in detail the influence of Cr and He on the evolution of microstructure. The irradiations were performed using Fe self-ions, in single beam mode and in dual beam mode (damage by Fe ions and co-implantation of He), to separate ballistic damage effect from the impact of simultaneous He injection. Three different dose ranges were studied: high dose (157 dpa, 17 appm He/dpa for the dual beam case), intermediate dose (45 dpa, 57 appm He/dpa for dual beam case) and in-situ low dose (0.33 dpa, 3030 appm He/dpa for the dual beam case). The experiments were performed at the JANNuS triple beam facility and dual beam in situ irradiation facility at CEA-Saclay and CSNSM, Orsay respectively. The microstructure was principally characterized by conventional TEM, APT and EDS in STEM mode. The main results are as follows: 1) A comparison of the cavity microstructure in high dose irradiated Fe revealed strong swelling reduction by the addition of He. It was achieved by a drastic reduction in cavity sizes and an increased number density. This behaviour was observed all along the damage depth, up to the damage peak. 2) Cavity microstructure was also studied in the dual beam high dose

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Onimus, F

    2003-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [Scientific Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States)] [and others

    1996-04-01

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

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

    Science.gov (United States)

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

    2016-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-01

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

  20. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    DEFF Research Database (Denmark)

    Singh, B.N.; Tähtinen, S.

    2002-01-01

    The present report summarizes and highlights the main results of the work carried out during the last 5-6 years on effects of neutron irradiation on physical and mechanical properties of copper and copper alloys. The work was an European contribution toITER Research and Development programme and...... regarding the suitability of a copper alloy for its use in the first wall and divertor components of ITER. It is pointed out that the present work has managed onlyto identify some of the critical problems and limitations of the copper alloys for their employment in the hostile environment of 14 MeV neutrons...

  3. Copper and phosphorus effect on residual embrittlement of irradiated model alloys and RPV steels after annealing

    International Nuclear Information System (INIS)

    The dependence of the recovery of the transition temperature shift after annealing (475 deg. C, 100 h) on copper and phosphorus contents has been studied on irradiated reactor pressure vessel (RPV) materials. A set of model alloys with low nickel content, lower than 0.2 mass%, was used for the study. Copper and phosphorus contents were varied in a wide range: 0.005-0.99 and 0.002-0.039 mass%, respectively. Recovery efficiency has been estimated by the value of residual embrittlement after annealing, measured in terms of a shift in transition temperature (ΔTKres). A comparison of the results obtained on model alloys with data for VVER-440 RPV materials has also been carried out. Comparative analysis has confirmed the conclusion that ΔTKres is independent of phosphorus content while the effect of copper on ΔTKres is not significant for typical VVER-440 RPV materials with a typical range of Cu contents between 0.10 and 0.24 mass%. However, for model alloys with a wider range of copper content, copper mainly controls the value of ΔTKres

  4. Possibilities of electric resistance method in study of metals and alloys irradiated to a high doses

    International Nuclear Information System (INIS)

    On the base of metals and alloys and reactor core materials fulfilled its task after high neutron dose and charged particles irradiation the electric resistance method possibilities are shown. It is determined that in the pure metal with BCC-structure (α-Fe and Mo) and TiAl alloy the point defects and its fine clusters and defect areas are mostly generating on atomic cascade shifts places partly relaxing into the dislocation loops. The loops contribution in electric resistance increase does not excess 3-4 and 1.6 % relatively for BCC-metals and TiAl alloy. The addition of impurity leads to impunity atom - radiation defect complex formation. The martensite decay in the U-7 hardened steels is carried out during the process of high power proton damage accompanying with the carbon isolation from the solids solution and dispersion carbide phase particles generation. In the materials of the WWR-K reactor control rod (12Cr18Ni9Ti, SAV-1), EhP-172 steel the process of radiation induced impurities redistributions has been carried out with its isolation from solid solution in the form of the fine phase on the grain boundaries or other preferable discharge

  5. Clustering Effects Under Irradiation in Fe-0.1%Cu Alloy : An Atomic Scale Investigation with the Tomographic Atom Probe

    OpenAIRE

    Pareige, P.; Welzel, S; Auger, P.

    1996-01-01

    In order to understand the effect of displacement cascades on the evolution of the microstructure of ferritic low copper supersaturated materials, analyses by 3D atomic tomography of neutron, electron and self ion irradiated Fe-0.1%Cu, were performed. This alloy was chosen because of its low copper concentration, close to that of french pressure vessel steels. The comparison of the microstructure evolutions in these irradiated specimens reveals the appearance of tiny copper "clusters" or "agg...

  6. Microstructure and local strain fields in a high-alloyed austenitic cast steel and a steel-matrix composite material after in situ tensile and cyclic deformation

    Energy Technology Data Exchange (ETDEWEB)

    Weidner, A.; Biermann, H. [Institute of Materials Engineering, Technische Universitaet Bergakademie Freiberg (Germany); Yanina, A.; Guk, S.; Kawalla, R. [Institute of Metal Forming, Technische Universitaet Bergakademie Freiberg (Germany)

    2011-09-15

    The tensile and cyclic deformation behaviour of a new metastable austenitic stainless cast TRIP (TRansformation Induced Plasticity) steel and a composite material consisting of austenitic steel matrix (AISI 304) with 5% MgO partially stabilized ZrO{sub 2} (MgO-PSZ) was studied in-situ in a scanning electron microscope (SEM). In-situ tests in the SEM show the evolution of the microstructure with the strain for uniaxial deformation and the number of cycles during fatigue, respectively. Initially, deformation bands develop. In these bands, the face-centred cubic austenite transforms into the hexagonal {epsilon}-martensite and subsequently to the body-centred cubic {alpha}'- martensite. This evolution was studied by different SEM techniques. Electron backscatter diffraction (EBSD) was applied for phase and orientation identification. The dislocation arrangement was investigated applying the electron channelling contrast imaging (ECCI) technique to different deformation stages. The studies are completed with measurements of local displacement fields using digital image correlation (DIC). (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. The influence of Boron on creep-rupture behaviour of austenitic unstabilized and Nb-stabilized stainless steel X8CrNi 1613 in unirradiated and irradiated condition

    International Nuclear Information System (INIS)

    The present study deals with influence of boron on creep-rupture behaviour in unirradiated condition at 6500C along with precipitation behaviour, heat-treatment and recrystallization of unstabilized and stabilized steel. The results of creep-rupture tests on unirradiated specimens show that boron exerts a beneficial effect on the rupture life and ductility. Boron losses its beneficial effect on creep properties in unstabilized steel by prolong creeping. The magnitude of beneficial effect of Boron on creep properties depends upon the initial boron distribution which influences the number, size and distribution of the precipitates. Boron promotes the precipitation of type M23C6 Carbides in the grain as well as at the grain boundary. Boron segregates in atomic form during slow cooling from austenitizing temperature. The recrystallization will be delayed by the presence of boron. The results of creep tests at 6500C shows that boron exerts a beneficial effect on creep life of irradiated steels. (orig./GSC)

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

    Directory of Open Access Journals (Sweden)

    Camila de Brito Ferreira

    2015-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  10. Dual-beam irradiation of friction stir spot welding of nanostructured ferritic oxide dispersion strengthened alloy

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Chun-Liang, E-mail: chunliang@mail.ndhu.edu.tw [Department of Materials Science and Engineering, I-Shou University, Kaohsiung 840, Taiwan (China); Richter, Asta [Department of Engineering, Technical University of Applied Sciences, Wildau, Bahnhofstrasse 1, 15745 Wildau (Germany); Koegler, Reinhard [Institute of Ion Beam Physics and Materials Research, Helmholtz Center Dresden-Rossendorf (HZDR), Bautzner Landstrasse 400, 01328 Dresden (Germany); Wu, Lung-Tien [Metal Industries Research and Development Centre, Kaohsiung 811, Taiwan (China)

    2012-09-25

    Highlights: Black-Right-Pointing-Pointer FSSW has been successfully applied to join two sheets of ODS material. Black-Right-Pointing-Pointer Dynamic recrystallization occurs in the thermo mechanically affected zone, resulting in a decrease in hardness. Black-Right-Pointing-Pointer After irradiation, a clear maximum is visible in all depth dependent hardness charts, which is caused by defects generated during ion bombardment. Black-Right-Pointing-Pointer The material changes during FSSW are larger than the radiation induced hardness increase. - Abstract: Nanostructured ferritic oxide dispersion strengthened (ODS) alloys usually contain a high density of Y-Al-O and Y-Ti-O nanoparticles, high dislocation densities and fine grains. Friction stir spot welding (FSSW) is a very promising technique for the joining of ODS materials without oxide particle agglomeration and loss in mechanical properties in the weld zone. Heating and severe plastic deformation can significantly alter the originally as-received material. The local microstructure determines the weld mechanical properties, which are analyzed by nanoindentation. The FSSW region consists of three different zones: the base material, the thermo-mechanically affected zone and the heat affected zone. Irradiation of the FSSW area was performed with a Fe{sup +}/He{sup +} dual ion beam. Hardness changes within the welding zones and variation with irradiation damage are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  12. Prediction model for oxide thickness on aluminum alloy cladding during irradiation

    International Nuclear Information System (INIS)

    An empirical model predicting the oxide film thickness on aluminum alloy cladding during irradiation has been developed as a function of irradiation time, temperature, heat flux, pH, and coolant flow rate. The existing models in the literature are neither consistent among themselves nor fit the measured data very well. They also lack versatility for various reactor situations such as a pH other than 5, high coolant flow rates, and fuel life longer than ∼1200 hrs. Particularly, they were not intended for use in irradiation situations. The newly developed model is applicable to these in-reactor situations as well as ex-reactor tests, and has a more accurate prediction capability. The new model demonstrated with consistent predictions to the measured data of UMUS and SIMONE fuel tests performed in the HFR, Petten, tests results from the ORR, and IRIS tests from the OSIRIS and to the data from the out-of-pile tests available in the literature as well. (author)

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

    Science.gov (United States)

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

    2016-09-01

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

  14. Film formation on the surface of magnesium-beryllium PMB-2 alloy in a diphenyl mixture under reactor irradiation

    International Nuclear Information System (INIS)

    A film growth on the surfaces of PMB-2 magnesium-beryllium alloy specimens in a diphenyl mixture under reactor irradiation was studies. It is shown that film thickness increases linearly with absorbed dose up to 3500 Mrad. The possibility of film washing off the specimen surfaces by boiling in the diphenyl mixture is investigated

  15. High Mn austenitic stainless steel

    Science.gov (United States)

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

    2010-07-13

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

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

    Science.gov (United States)

    Huang, Chen-Hsi; Marian, Jaime

    2016-10-01

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

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

    Science.gov (United States)

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

    2016-12-01

    As is well documented, platinum nanoparticles, promising for catalysts for fuel cells, exhibit better catalytic activities, when alloyed with Zn. Pre-existing syntheses of Pt-Zn alloy catalysts are composed of a number of complex steps. In this study, we have demonstrated that nanoparticles of Pt-Zn alloys are simply generated by electron-beam irradiation in a transmission electron microscope of a wurtzite ZnO single-crystal specimen. The initial ZnO specimen is considered to have been contaminated by Pt during specimen preparation by focused ion beam milling. The formation of the nanoparticle is explained within the framework of ionization damage (radiolysis) by electron-beam irradiation and accompanying electrostatic charging. PMID:27440080

  18. Change in austenite transformation kinetics under hot rolling action

    International Nuclear Information System (INIS)

    The effect of hot plastic deformation on kinetics of austenite transformation both during continuous cooling and under isothermal conditions, is studied. Experiments are performed using the 40 Kh, 60 KhC2, 40KhNM and 30KhGSN2 steels. It is shown that hot working speeds up isothermal transformation of austenite of low- and medium alloyed steels in pearlite range. In medium-alloyed 30KhGSN2 40KhNM steels hot working does not speed up atherma.l austenite transformation in the pearlite range and somewhat hinders it in the bainite range, due to which hardenability must not reduce at high temperatUre thermomechanical treatment. The difference in the effect of hot working on isothermal and athermal austenite transformations is conditioned by the effect of after-deformation pauses, which are practically inevitable in cases of continuous cooling of products

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

    Science.gov (United States)

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

    2015-06-01

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

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

    International Nuclear Information System (INIS)

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

  1. EFFECT OF CHEMICAL COMPOSITION ON RETAINED AUSTENITE IN TRIP STEEL

    Institute of Scientific and Technical Information of China (English)

    Y. Chen; X. Chen; Q.F. Wang; G.L. Yuan; C.Y. Li; X.Y. Li; Y.X. Wang

    2002-01-01

    The systematic chemical compositions including common C, Si, Mn, Al, and micro- alloying elements of Ti and Nb were designed for high volume fraction of retained austenite as much as possible. The thermo-cycle experiments were conducted by using Gleeble 2000 thermo-dynamic test machine for finding the appropriate composition. The experimental results showed that chemical composition had a significant effect on retained austenite, and the appropriate compositions were determined for commercial production of TRIP steels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  3. Stable atomic structure of NiTi austenite

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-01

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

  4. Ballistic impact properties of mixed multi-layered amorphous surface alloyed materials fabricated by high-energy electron-beam irradiation

    International Nuclear Information System (INIS)

    The objective of this study is to investigate ballistic impact properties of multi-layered amorphous surface alloyed materials fabricated by high-energy electron-beam irradiation. The mixture of Zr-based amorphous alloy powders and LiF+MgF2 flux powders was deposited on a Ti alloy substrate, and then electron beam was irradiated on this powder mixture to fabricate an one-layered surface alloyed material. On top of this layer, the powder mixture was deposited again and then irradiated with electron beam whose beam current was decreased to fabricate the multi-layered surface alloyed material. In the mixed multi-layered surface alloyed materials fabricated with LM1 alloy powders and LM2 or LM10 alloy powders, the surface region consisted of amorphous phases, together with a small amount of crystalline particles, whereas the center region was complicatedly composed of amorphous phases, crystallized phases, and dendritic β phases. Since the surface region mostly composed of amorphous matrix was quite hard, the alloyed materials sufficiently blocked the travel of a projectile. When cracks formed at the surface region propagated into the center region, the formation of many cracks or debris was accelerated, which could beneficially work for absorbing the ballistic impact energy, thereby leading to the higher ballistic impact properties than the surface alloyed materials fabricated with LM1 or LM2 alloy powders

  5. Effect of irradiation temperature on microstructure, radiation hardening and embrittlement of pure copper and copper-based alloy

    International Nuclear Information System (INIS)

    Low-temperature radiation embrittlement is one of the main negative consequences of neutron irradiation for pure copper and copper-based alloys. But currently available data on copper radiation hardening and embrittlement have been obtained in the temperature range T irr = 60-90 oC. Systematic data on the effect of irradiation temperature in the range of radiation hardening and embrittlement (50-200 oC) are lacking. This paper presents the results of the analysis of two experiments on irradiation of pure copper and GlidCop Al25IG alloy in the RBT-6 reactor at irradiation temperatures of 80 oC and 150 oC. The irradiation dose range was 10-3-10-1 dpa. The comparison between the dose dependencies of materials hardening and embrittlement revealed that a rise in temperature causes hardening to drop and embrittlement to decrease. The microstructure data on the density and size of complexes in irradiated materials served as the basis for calculations of the level of radiation hardening, with the Orowan-Seeger model used

  6. Mechanical properties testing of several 800 MeV proton irradiated BCC metals and alloys

    International Nuclear Information System (INIS)

    A spallation neutron source for the 600-MeV proton accelerator facility at the Swiss Institute for Nuclear Research (SIN) consists of a vertical cylinder filled with molten Pb-Bi. The proton beam enters the cylinder, passing upward through a window in contact with the Pb-Bi eutectic liquid. Investigations are underway at the 800-MeV proton accelerator at LAMPF to test the performance of candidate SIN window materials. Based on considerations of chemical compatibility with molten Pb-Bi, as well as radiation damage mechanisms, Fe, Ta, Fe-2.25Cr-1Mo, and Fe-12Cr-1Mo (Ht-9) were chosen as candidate materials. Sheet tensile samples were sealed inside capsules containing Pb-Bi and were proton-irradiated at LAMPF to two fluences, 4.8 and 54 x 1023 p/m2. The beam current was approximately equal to the 1 mA anticipated for the upgraded SIN accelerator. Yield and ultimate strengths increased upon irradiation in all materials, while the ductility decreased. The pure metals, Ta and Fe, exhibited the greatest radiation hardening and embrittlement. The HT-9 alloy showed the smallest changes in strength and ductility

  7. Separation and purification of fission 99Mo from neutron irradiated UAl3 alloy

    International Nuclear Information System (INIS)

    A method has been developed for the separation of fission product 99Mo from irradiated uranium aluminum alloy. The method consists of dissolution of the irradiated target in 6 M NaOH, whereby only aluminium along with 99Mo, 131I and 103Ru get into the solution with traces of 95Zr, 95Nb and 132Te, while all other fission products, activation products (239Np) and uranium remain as solid residue. Al(OH)3 precipitation at lower pH (8-9) removed some of the impurities, e.g. 95Zr, 95Nb, 132Te while AgI/AgIO3 precipitation removed almost all the 131I. 103Ru was removed by addition of NaBiO3 and evaporation to dryness. Subsequently 99Mo was purified by precipitation as Mo-α-benzoin oxime which was dissolved in dilute NaOH. This was subjected to organic impurity and trace iodine separation by passing through silver coated activated charcoal. Final purification was carried out by anion exchange separation. 99Mo was obtained with an overall recovery of 80%. Purity of the 99Mo product was found to be in agreement with the US and European pharmacopoeia. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  9. The radiation hardening of austenitic stainless steel after heavy ion and neutron irradiation; Radiacyjne umocnienie stali austenitycznej 0H18N10T napromienionej ciezkimi jonami i neutronami

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, A.; Kochanski, T.; Chrzanowski, J. [Institute of Atomic Energy, Otwock-Swierk (Poland); Shchegolev, V. [Joint Inst. for Nuclear Research, Dubna (Russian Federation)

    1992-07-01

    The results of radiation hardening of steel 0H18N10T after Ne ion (E=230 MeV) and neutron (E>0.1 MeV) irradiation are described. The dose dependence of the change of the mechanical properties was investigated. It is shown that both after ion as neutron irradiation the yield strength increases and the deformation decreases. However a certain effect of hardening after neutron irradiation is achieved at a dose smaller than in case of Ne irradiation. Experimental results are compared with existing models of radiation hardening. The present experiment has been shown the possibility of practical utilization of high-energy ions modelling the changes of the mechanical properties in the materials after reactor irradiation. (author). 18 refs, 11 figs, 3 tabs.

  10. Environmentally Assisted cracking behaviors of austenitic steels representing GB compositions of irradiated 304 steel and implications to mitigation of IASCC in PWR

    International Nuclear Information System (INIS)

    Environment - assisted cracking of austenitic steels is studied in simulated PWR primary water at 325 deg C using solution treated specimens with 12 % Cr, 28 % Ni, 0 - 2.74 % Si doped with P or S and subjected to sensitization treatment at 650 deg C for 20 h. It is revealed that the solution treated steel with 2.74 % Si is highly susceptible to intergranular cracking under given conditions. Silicon in steel is shown to increase a crack growth rate, especially, in near - surface zones. Neither P nor S exhibit noticeable effects on intergranular cracking in the water. Heat treatment (650 deg C, 20 h) decreases a crack growth rate in high Si steels but does not prevent intergranular corrosion. Based on the results obtained for impurity, water chemistry and heat treatment effects, possible ways to decrease environment - assisted stress corrosion cracking are discussed

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

    Science.gov (United States)

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

    2014-12-01

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

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

    KAUST Repository

    Guo, Zaibing

    2014-02-01

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

  13. Influence of oxygen ion irradiation on the corrosion aspects of Ti-5%Ta-2%Nb alloy and oxide coated titanium

    International Nuclear Information System (INIS)

    The corrosion resistance of Ti-5%Ta-2%Nb alloy and DOCTOR (double oxide coating on titanium for reconditioning) coated titanium by O5+ ion irradiation were compared and investigated for their corrosion behaviour. O5+ ion irradiations were carried out at a dose rate of 1 x 1017, 1 x 1018 and 1 x 1019 ions/m2 at 116 MeV. The surface properties and corrosion resistance were evaluated by using scanning electron microscopy (SEM), atomic force microscopy (AFM), energy dispersive X-ray (EDX), glancing-angle X-ray diffraction (GXRD) and electrochemical testing methods. The results of electrochemical investigations in 11.5 N HNO3 indicated that the open circuit potential (OCP) of DOCTOR coated titanium is nobler than Ti-5%Ta-2%Nb alloy. The potentiodynamic polarization study of Ti-5%Ta-2%Nb alloy and DOCTOR coated specimen indicated decrease in passive current density with increase in ion doses (1 x 1017 to 1 x 1019 ions/m2) indicating decrease in anodic dissolution. Nyquist arc behaviour in the electrochemical impedance study substantiated the enhancement in oxide layer stability by O5+ ion irradiation. AFM results revealed that the DOCTOR coated Ti surface was dense without gross voids, and the surface roughness decreased by O5+ ion irradiation, but increased after corrosion test. EDX and GXRD patterns of DOCTOR coated Ti sample indicated that the coating was mainly composed of rutile TiO2. Based on the above results, the O5+ ion irradiation effect on corrosion behavior of Ti-5%Ta-2%Nb alloy and DOCTOR coated titanium are discussed in this paper

  14. Pitting corrosion resistant austenite stainless steel

    Science.gov (United States)

    van Rooyen, D.; Bandy, R.

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

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

    DEFF Research Database (Denmark)

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

    1986-01-01

    Samples of pure aluminium (99.9999%) and commercial Al-2.7%Mg(AlMg3) and Al-1.1%Mg-0.5%Si(Al6061) alloys were irradiated with 800 MeV protons at the Los Alamos Meson Physics Facility (LAMPF) at a temperature between 40-100°C to a maximum dose of 0.2 dpa. Transmission electron microscopy (TEM) sho...... higher than in the pure Al. The amount of gas accumulation at grain boundaries was found to depend on gas generation rate, alloying and cold-work microstructure...

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

    International Nuclear Information System (INIS)

    A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5 MeV helium ions to a fluence of 5 × 1021 ion m−2 (5000 appm) with a dose of about 1 dpa (displacements per atom). A uniform helium ion stopping damage region about 17 μm deep from the target surface was achieved by applying an energy degrading wheel. Radiation damage defects including helium-vacancy clusters and small helium bubbles were found in the microstructure of the samples irradiated at room temperature. With increasing irradiation temperature to 300 °C and 500 °C helium bubbles were clearly observed in both the α2 and γ phases of the irradiated microstructure. By means of nanoindentation significant irradiation hardening was measured. For the samples irradiated at room temperature the hardness increased from 5.6 GPa to 8.5 GPa and the irradiation-hardening effect reduced to approximately 8.0 GPa for the samples irradiated at 300 °C and 500 °C

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-15

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

  18. Neutron metrology in the HFR. Irradiation of vanadium alloys. Experiment R204-7/8/9 (VABONA)

    International Nuclear Information System (INIS)

    The irradiation experiment R204-7/8/9 is part of a material research program VABONA in which the radiation damage is investigated of different vanadium alloys. The experiment deals with the irradiation of vanadium specimens α-implanted at the cyclotron in Ispra. The irradiation has been performed in a SIENA capsule in the HFR Petten at three different temperatures (873, 973 and 1073 K). The irradiation took place in position C5 during seven HFR operating cycles. The target damage level was 5 dpa (displacements per atom). This report presents the final neutron metrology results obtained from activation monitors in the three specimen holders, coded as R204-7/8/9. Data about the number of displacements per atom are also included. The main results of the thermal and fast neutron fluence measurements are presented. (orig.)

  19. Effect of Chemistry on the Transformation of Austenite to Martensite for Intercritically Austempered Ductile Iron

    OpenAIRE

    Banerjee, Sayanti

    2013-01-01

    Intercritically austempered ductile iron (IADI) with a matrix microstructure of ferrite plus metastable austenite has an excellent combination of strength and toughness. The high strength and good ductility of this material is due to the transformation of metastable austenite to martensite during deformation. In the present study, the transformation of austenite to martensite for intercritically austempered ductile irons of varying alloy chemistry (varying amounts of nickel and/or manganese) ...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-01

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

  2. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation.

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-14

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic < a > dislocations has been dynamically observed before and after irradiation at room temperature and 300 degrees C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by < a > dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, {01 (1) over bar1}< 0 (1) over bar 12 > twinning was identified in the irradiated sample deformed at 300 degrees C.

  3. Nonequilibrium segregation and phase instability in alloy films during elevated-temperature irradiation in a high-voltage electron microscope

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-05-01

    The effects of defect-production rate gradients, caused by the radial nonuniformity in the electron flux distribution, on solute segregation and phase stability in alloy films undergoing high-voltage electron-microscope (HVEM) irradiation at high temperatures are assessed. Two-dimensional (axially symmetric) compositional redistributions were calculated, taking into account both axial and transverse radial defect fluxes. It was found that when highly focused beams were employed radiation-induced segregation consisted of two stages: dominant axial segregation at the film surfaces at short irradiation times and competitive radial segregation at longer times. The average alloy composition within the irradiated region could differ greatly from that irradiated with a uniform beam, because of the additional atom transport from or to the region surrounding the irradiated zone under the influence of radial fluxes. As a result, damage-rate gradient effects must be taken into account when interpreting in-situ HVEM observations of segregation-induced phase instabilities. The theoretical predictions are compared with experimental observations of the temporal and spatial dependence of segregation-induced precipitation in thin films of Ni-Al, Ni-Ge and Ni-Si solid solutions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  5. Computer simulations of the Ni2MnGa alloys

    Science.gov (United States)

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

    2008-07-01

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

  6. Effect of laser irradiation conditions on the laser welding strength of cobalt-chromium and gold alloys.

    Science.gov (United States)

    Kikuchi, Hisaji; Kurotani, Tomoko; Kaketani, Masahiro; Hiraguchi, Hisako; Hirose, Hideharu; Yoneyama, Takayuki

    2011-09-01

    Using tensile tests, this study investigated differences in the welding strength of casts of cobalt-chromium and gold alloys resulting from changes in the voltage and pulse duration in order to clarify the optimum conditions of laser irradiation for achieving favorable welding strength. Laser irradiation was performed at voltages of 150 V and 170 V with pulse durations of 4, 8, and 12 ms. For cobalt-chromium and gold alloys, it was found that a good welding strength could be achieved using a voltage of 170 V, a pulse duration of 8 ms, and a spot diameter of 0.5 mm. However, when the power density was set higher than this, defects tended to occur, suggesting the need for care when establishing welding conditions. PMID:21959656

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    Science.gov (United States)

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

    1988-07-01

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

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

    International Nuclear Information System (INIS)

    Graphical abstract: The effect of annealing on VmHn complexes in irradiated Fe–0.3%Cu alloy was investigated using positron annihilation spectroscopy based on a slow positron beam. The formation of vacancy-type defects due to the 0.1 dpa irradiation would lead to the increment of the S parameter compared to the unirradiated specimen. A larger number of hydrogen deposited at the damage area and hydrogen atoms (Hn) occupied vacancy sites (Vm), which could lead to the formation of numerous VmHn complexes. The irradiated sample was annealed isochronally for 30 min at 150 °C, 200 °C, 300 °C, 400 °C and 500 °C, respectively. An evident defect peak formed at 150 °C, which meant the VmHn complexes were broken and a larger of hydrogen atoms was escaping. The remaining vacancy-type defect induced the increment of the S parameter. The peaks migrated from the damage area towards to surface region (peaks 1–5). The S parameters in the damage area gradually decreased with the increasing annealing temperature. The shrinkage of vacancy-type defects and the density of defects induced the decrease of S parameters during elevated annealing treatment. - Highlights: • VmHn complexes formed in H+ irradiated Fe and Fe–0.3%Cu alloys. • S–W results confirmed the formation of Cu precipitates in irradiated Fe–0.3%Cu. • VmHn complexes were broken and H atoms were escaping after annealing treatment. • S peaks would migrate towards to surface with the annealing temperature. - Abstract: The effect of annealing on VmHn complexes and Cu precipitate behaviours in hydrogen ion irradiated Fe and Fe–0.3%Cu alloys was investigated by positron annihilation spectroscopy using a slow positron beam. The results of S parameters indicated that the room temperature irradiation was benefit for the formation of the VmHn complex compared to the elevated temperature irradiation. The S–W results confirmed the formation of Cu precipitates in Fe–0.3%Cu even at the irradiation dose of 0

  11. Influence of the austenite-martensite transformation in the dimensional stability of a new tool steel alloyed with niobium (0.08% wt.) and vanadium (0.12% wt.)

    International Nuclear Information System (INIS)

    Austenite-martensite transformation influence on the dimensional stability of a new experimental tool steel alloyed with niobium (0.08% wt.) and vanadium (0.12% wt.) has been studied. The dimensional stability of this new steel was compared with the dimensional stability of commercial steel, after and before two thermal treatments, T1 (860 degree centigrade) and T2 (900 degree centigrade). The thermal treatments consisted on heating and cooling, at 1 atmosphere of pressure, in N2 atmosphere furnace, following by heating in a conventional furnace at 180 degree centigrade during 1 hour. Initially, the experimental steel composition and Ac1 and Ac3 transformation temperatures were determined by glow-discharge luminescence (GDL) and dilatometric tests, respectively, in order to select the austenization temperatures of T1 and T2 treatments. After hardness measurement, the microstructure of both steels was characterized by X-Ray Diffraction (XRD) and optical metallography, before and after of T1 and T2 thermal treatments. Finally, longitudinal and angular dimensional stability analyses were realized for both commercial and experimental steels. After a contrastive hypothesis analysis, the results showed that the longitudinal relative variation of the experimental steel calculated was around 0.2% and the angular relative variation was not significant. (Author)

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

    Science.gov (United States)

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

    2016-07-01

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

  13. Irradiation creep in path A alloys irradiated to 5 dpa in the ORR-MFE-4A spectral tailoring experiment

    International Nuclear Information System (INIS)

    Pressurized tubes have been irradiated to 4.8 dpa at 330 and 4000C in the ORR-MFE-4A spectral tailoring experiment. At 3300C, 20%-cold-worked type 316 stainless steel (CW 316) and path A PCA demonstrated irradiation creep similar to predictions of the irradiation creep equation developed in the Fast Breeder Reactor Program. The creep rate for path A PCA was approximately 25% higher than that of type 316 stainless steel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  15. Anisotropic radiation-induced segregation in 316L austenitic stainless steel with grain boundary character

    International Nuclear Information System (INIS)

    Radiation-induced segregation (RIS) and subsequent depletion of chromium along grain boundaries has been shown to be an important factor in irradiation-assisted stress corrosion cracking in austenitic face-centered cubic (fcc)-based alloys used for nuclear energy systems. A full understanding of RIS requires examination of the effect of the grain boundary character on the segregation process. Understanding how specific grain boundary structures respond under irradiation would assist in developing or designing alloys that are more efficient at removing point defects, or reducing the overall rate of deleterious Cr segregation. This study shows that solute segregation is dependent not only on grain boundary misorientation, but also on the grain boundary plane, as highlighted by markedly different segregation behavior for the Σ3 incoherent and coherent grain boundaries. The link between RIS and atomistic modeling is also explored through molecular dynamic simulations of the interaction of vacancies at different grain boundary structures through defect energetics in a simple model system. A key insight from the coupled experimental RIS measurements and corresponding defect–grain boundary modeling is that grain boundary–vacancy formation energy may have a critical threshold value related to the major alloying elements’ solute segregation

  16. Alloy development for irradiation performance in fusion reactors. Annual report, September 1978-September 1979

    International Nuclear Information System (INIS)

    This report is the first annual report of research activities directed toward the development of improved performance alloys for such severe environments as the fusion reactor fist wall. Major project efforts are directed toward definition of alloy performance requirements, alloy design, alloy production and alloy performance evaluation. Rapid solidification from the melt is being used to manipulate alloy microstructure and to produce the desired design properties. Integrated testing and modeling procedures have been developed to minimize testing requirements. Progress during the first project year and future plans are summarized in this annual report

  17. A problem to determine short term mechanical properties changes of ferrite-martensite and austenitic steels as materials of fuel assembly of fast reactors under high dose neutron irradiation

    International Nuclear Information System (INIS)

    The results of mechanical tests of flat and ring-shaped samples of two ferrite-martensite steels C0.1-Cr13-Mo2-Nb-V-W and C0.1-Cr12-Mo-Nb-V-W irradiated to different damage doses (up to 100 dpa) have been performed in this work. It have been shown that values of plasticity and strength characteristics determined on this sample types are different. Specific elongation takes the values 8-12% for the flat samples, at the same time, t takes the values 1-3% for the ring-shaped samples at room temperature. A character of fluence dependence of mechanical properties is identical. The steels show viscous damage in all tests. Samples of fuel pin cladding fabricated from the austenitic steel C0.1-Cr16-Ni15-Mo3 were also investigated after there working out in BN-600 reactor up to 76 dpa. Ring-shaped samples were tested at standard single-axle tension. Tube samples were tested by internal pressure of solid filler. All samples were fabricated from one and the same section, mechanical properties obtained are different. Specific elongation of the brittlest section of fuel pin was 0-0.9% for the ring-shaped samples and 2-7% for the tube samples at room temperature. Fractographic investigations were carried out on the samples after mechanical tests. Possible reasons of such difference have been discussed in the work. (author)

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

    Science.gov (United States)

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

    2015-07-01

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

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

    Science.gov (United States)

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

    2011-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Voertler, K; Juslin, N; Nordlund, K [Association EURATOM-Tekes, Department of Physics, University of Helsinki, PO Box 43, FIN-00014 (Finland); Bonny, G; Malerba, L, E-mail: katharina.vortler@helsinki.fi [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium)

    2011-09-07

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-15

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

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

    Science.gov (United States)

    Kohnert, Aaron Anthony

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, L.; Kempe, H.

    1994-06-01

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

  4. Post-irradiation annealing behavior of neutron-irradiated FeCu, FeMnNi and FeMnNiCu model alloys investigated by means of small-angle neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F. [Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, 01314 Dresden (Germany); Ulbricht, A., E-mail: a.ulbricht@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, 01314 Dresden (Germany); Lindner, P. [Institut Laue-Langevin Grenoble, BP 156, 38042 Grenoble Cedex 9 (France); Keiderling, U. [Helmholtz-Zentrum Berlin, Hahn-Meitner-Platz 1, 14109 Berlin (Germany); Malerba, L. [SCK-CEN Mol, Boeretang 200, 2400 Mol (Belgium)

    2014-11-15

    Neutron irradiation of reactor pressure vessel steels gives rise to the formation of thermodynamically stable and unstable nano-features. The present work is focused on the stability of Cu-, Mn- and Ni-containing solute clusters in model alloys exposed to post-irradiation annealing. Fe0.1Cu, Fe1.2Mn0.7Ni and Fe1.2Mn0.7Ni0.1Cu (wt%) model alloys irradiated up to neutron exposures of 0.1 and 0.19 dpa (displacements per atom) were annealed at stepwise increasing temperatures in the range from 300 °C (i.e. near irradiation temperature) to 500 °C and characterized by means of small-angle neutron scattering (SANS). We have found characteristic differences in the annealing behavior of the alloys. In particular, there is a non-trivial (synergistic–antagonistic) interplay of Mn/Ni and Cu.

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

    Energy Technology Data Exchange (ETDEWEB)

    Frely, E

    1997-12-31

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

  6. Nanostructure evolution under irradiation in FeMnNi alloys: A “grey alloy” object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France)

    2015-07-15

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

  7. Nanostructure evolution under irradiation in FeMnNi alloys: A “grey alloy” object kinetic Monte Carlo model

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2000-01-01

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

  9. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Singh, B.N; Edwards, D.J.; Bilde-Sørensen, Jørgen

    2004-01-01

    has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons.Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron...... opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimensactivate dislocation sources at the sites of stress concentration...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  11. A study on the influence of trace elements (C, S, B, Al, N) on the hot ductility of the high purity austenitic alloy Fe-Ni 36% (INVAR)

    Energy Technology Data Exchange (ETDEWEB)

    Simonetta-Perrot, M.T.

    1994-11-01

    In order to study the damage mechanisms leading to the ductility decrease of the Invar alloy at 600 C, a high-purity Fe-Ni 36% sample has been doped with trace elements with the purpose of identifying the role of sulfur, sulfur with Al N or B N precipitates and sulfur with boron, on the ductility, the failure modes, the intergranular damage and the plastic deformation mechanisms prior to failure. A new AES segregation quantification method has been used to study the kinetics and thermodynamics of intergranular and surface segregations and determine the relation between sulfur segregation and grain joint fragility. refs., figs., tabs.

  12. Effects ofγ-irradiation and Deformation Temperature on Tensile Properties of Pb-2 mass% Sb Alloy

    Institute of Scientific and Technical Information of China (English)

    Gh MOHAMMED; SEI-GAMAL

    2016-01-01

    Effects ofγ-irradiation and deformation temperature (T)on the tensile properties of Pb-2 mass% Sb alloys were studied.The samples were annealed at 458 K for 2 h in air,then water quenched after they wereγ-irradiated (the different doses were 0·5,1·0,1·5,and 2·0 MGy).The tensile properties were performed using stress-strain measurements at a constant strain rate (1·2×10-3 s-1 )and at different T (303-393 K).It was found that at con-stant dose,the fracture stress (σF )decreases while the fracture strain (εF )increases as T increases.At particular T,σF increases whileεF decreases with increasing dose.The strain-hardening exponent (n),which is the slope of the relation between ln(σ)and ln(ε)of the parabolic part of the stress-strain curve,was determined and its values in-crease as T increases and decrease as the dose increases.The value of the activation energy increases as the dose in-creases from 0·07 eV for un-irradiated sample to 0·1 eV for the 2 MGy-irradiated sample.These values are in ac-cordance with that needed for dislocation movement and ordering process.An interpretation of the results was given, based on the creation of point and line defects due toγ-irradiation,and that results in a distribution of beta phase (Sb-phase),leading to a difficulty in the movement of dislocations,so there is an increase in alloy hardness.

  13. Impact of Ion Irradiation upon Structure and Magnetic Properties of NANOPERM-Type Amorphous and Nanocrystalline Alloys

    Directory of Open Access Journals (Sweden)

    Marcel Miglierini

    2015-01-01

    Full Text Available Structural modifications and their impact upon magnetic properties are studied in amorphous and nanocrystalline NANOPERM-type 57Fe75Mo8Cu1B16 alloy. They are introduced by irradiation with 130 keV N+ ions to the total fluencies of up to 2.5 × 1017 ions/cm2 under different cooling conditions. Increased temperature during the irradiation triggers formation of nanocrystallites of bcc-Fe in those subsurface regions that are affected by bombarding ions. No crystallization occurs when good thermal contact between the irradiated sample and a sample holder is assured. Instead, structural rearrangement which favours development of magnetically active regions was determined by the local probe methods of Mössbauer spectrometry. Dipole magnetic interactions dominate in subsurface regions on that side of the ribbons which was exposed to ion irradiation. Nevertheless, structural modifications demonstrate themselves also via macroscopic magnetic parameters such as temperature dependence of magnetization, Curie temperature, and hysteresis loops. Impact of only the temperature itself to the observed effects is assessed by the help of samples that were subjected just to heat treatment, that is, without ion irradiation.

  14. Defects in hyperpure Fe-based alloys created by 3 MeV e{sup -}-irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, X.H.; Moser, P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M.; Van Duysen, C. [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    Information about vacancy defects created in RPV (Reactor Pressure Vessels) steels after neutron irradiations are obtained via a simulation: the RPV steels are simulated by a series of high purity Fe-based alloys; the neutron irradiation is simulated by a 3 MeV electron irradiation; vacancy defects characteristics are obtained by positron lifetime techniques. Irradiations are made at 150 or 288 deg C, with a dose of 4*10{sup 19} e-/cm{sup 2}, and followed by isochronal annealing in the range 20-500 deg C. The observed vacancy defects are single trapped vacancies and small vacancy clusters, the size of which being lower than 10 empty atomic volumes (vacancy clusters containing more than 50 empty atomic volumes were never found). A large recovery step is observed between 200 and 400 deg C, after 150 deg C irradiation and attributed to vacancy-impurity detrapping, and also, vacancy cluster evaporation. The influence of C, Cu and Mo are presented. These results are in agreement with a model supposing, in pure Fe, single vacancy migration at -50 deg C and vacancy-impurity detrapping at 200 deg C. (authors). 4 figs., 15 refs.

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

    Science.gov (United States)

    Lu, Chenyang; Jin, Ke; Béland, Laurent K.; Zhang, Feifei; Yang, Taini; Qiao, Liang; Zhang, Yanwen; Bei, Hongbin; Christen, Hans M.; Stoller, Roger E.; Wang, Lumin

    2016-02-01

    Energetic ions have been widely used to evaluate the irradiation tolerance of structural materials for nuclear power applications and to modify material properties. It is important to understand the defect production, annihilation and migration mechanisms during and after collision cascades. In this study, single crystalline pure nickel metal and single-phase concentrated solid solution alloys of 50%Ni50%Co (NiCo) and 50%Ni50%Fe (NiFe) without apparent preexisting defect sinks were employed to study defect dynamics under ion irradiation. Both cross-sectional transmission electron microscopy characterization (TEM) and Rutherford backscattering spectrometry channeling (RBS-C) spectra show that the range of radiation-induced defect clusters far exceed the theoretically predicted depth in all materials after high-dose irradiation. Defects in nickel migrate faster than in NiCo and NiFe. Both vacancy-type stacking fault tetrahedra (SFT) and interstitial loops coexist in the same region, which is consistent with molecular dynamics simulations. Kinetic activation relaxation technique (k-ART) simulations for nickel showed that small vacancy clusters, such as di-vacancies and tri-vacancies, created by collision cascades are highly mobile, even at room temperature. The slower migration of defects in the alloy along with more localized energy dissipation of the displacement cascade may lead to enhanced radiation tolerance.

  16. Measurement of in-vivo dosage increase due to dental alloys during therapeutic irradiation of the mouth cavity

    International Nuclear Information System (INIS)

    The degree of dosage increase in the immediate surrounding of metallic dental materials was measured in an in-vivo study during therapeutic irradiation with 60 Co gamma rays in the area of mouth cavity of 11 patients. Measurements were carried out by thermoluminescent dosimetry at permamently fixed golden teeth and alloy specimens containing gold and palladium and amalgam. The following relative dodage values according to a simultanelusly measured reference value were measured at the surface of the different dental materials: 161% near fixed golden caps, 168% near the specimen containing gold in a high percentage, 133% near the specimen of palladium and 161% near the specimen of amalgam. The in vivo measured dosage increases due to metallic dental prosthesis are less than values obtained using back scatter arramgements for irradiating phantoms. Despite this, they could be of clinical relevance. Thus the usage of a mucous membrane protection during irradiation with 60 Co, as a means of preventing local lesions of the oral mucosa, due to dental alloys within the treatment volume remains inevitable. (orig.)

  17. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe–Cr model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); Pareige, C. [Groupe de Physique des Matériaux, Université et INSA de Rouen, UMR 6634 CNRS, Avenue de l’Université, BP 12, 76801 Saint Etienne du Rouvray (France); Hernández-Mayoral, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Malerba, L. [SCK-CEN, Nuclear Material Science Institute, Boeretang 200, B-2400 Mol (Belgium); Heintze, C. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2014-05-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe–Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α′-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

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

    Science.gov (United States)

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

    1990-07-01

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

  19. Zirconium alloy oxidation and hydriding under irradiation: Review of Pacific Northwest Laboratories' test program results: Final report

    International Nuclear Information System (INIS)

    Radiation effects on zirconium alloy oxidation and hydriding were investigated in the Advanced Test Reactor (ATR) and Engineering Test Reactor (ETR). The investigations represent one of the largest data bases on oxidation and hydriding of zirconium alloys. Much of the data base has been published, but some results were unpublished when the federal programs terminated. Due to the renewed interest in zirconium alloy cladding behavior, the Electric Power Research Institute sponsored documentation of the unpublished results and a summary of principal results from the prior publications. The data base involves nine zirconium alloys; multiple metallurgical conditions; neutron flux levels from ∼1012 to 1.8 x 1014 n/cm2 .sec, > 1 MeV; fluence levels to 1.5 x 1022 n/cm2, > 1 MeV; oxygenated and low-oxygen coolants; in flux, out-of-flux, and out-of-reactor comparisons on identical specimens; transfer of specimens exposed in one loop water chemistry to another loop chemistry; dissimilar metal combinations; investigation of surface pretreatment effects. The loop results parallel in several respects oxidation and hydriding characteristics of water reactor fuel cladding and pressure tubes. The report summarizes results on the following areas; oxidation and hydriding trends under irradiation; localized phenomena; unusual oxidation effects; dissimilar metal effects; effects of fluoride contamination; metal density changes; deposition phenomena

  20. Filtration–UV irradiation as an option for mitigating the risk of microbiologically influenced corrosion of subsea construction alloys in seawater

    International Nuclear Information System (INIS)

    Highlights: •Biofilms ennobled Ecorr of offshore construction alloys in natural seawater. •Filtration–UV irradiation delayed biofilm growth and activity on alloys. •Localized corrosion in seawater was lowered by the use of filtration–UV irradiation. •Biofilm community composition was affected by both substratum and seawater treatment. •Filtration–UV irradiation can be an ecofriendly practice for protection against MIC. -- Abstract: The effect of filtration–UV irradiation of seawater on the biofilm activity on several offshore structural alloys was evaluated in a continuous flow system over 90 days. Biofilms ennobled the electrode potential by +400 to 500 mV within a few days of exposure to raw untreated seawater. Filtration–UV irradiation of the seawater delayed the ennoblement of the steels for up to 40 days and lowered localized corrosion rates in susceptible alloys. Ennobling biofilms were composed of microbial cells, diatoms and extracellular polymeric substances and the bacterial community in biofilms was affected by both the alloy composition and seawater treatment

  1. Development of Cast Alumina-Forming Austenitic Stainless Steels

    Science.gov (United States)

    Muralidharan, G.; Yamamoto, Y.; Brady, M. P.; Walker, L. R.; Meyer, H. M., III; Leonard, D. N.

    2016-09-01

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

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

    Science.gov (United States)

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

    2016-08-01

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

  3. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    Science.gov (United States)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  4. Revised ANL-reported tensile data for unirradiated and irradiated (FFTF, HFIR) V-Ti and V-Cr-Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States)

    1998-03-01

    The tensile data for all unirradiated and irradiated vanadium alloys samples tested at Argonne National Laboratory (ANL) have been critically reviewed and, when necessary, revised. The review and revision are based on reanalyzing the original load-displacement strip chart recordings by a methodology consistent with current ASTM standards. For unirradiated alloys (162 samples), the revised values differ from the previous values as follows: {minus}11{+-}19 MPa ({minus}4{+-}6%) for yield strength (YS), {minus}3{+-}15 MPa ({minus}1{+-}3%) for ultimate tensile strength (UTS), {minus}5{+-}2% strain for uniform elongation (UE), and {minus}4{+-}2% strain for total elongation (TE). Of these changes, the decrease in {minus}1{+-}6 MPa (0{+-}1%) for UTS, {minus}5{+-}2% for UE, and {minus}4{+-}2% for TE. Of these changes, the decrease in UE values for alloys irradiated and tested at 400--435 C is the most significant. This decrease results from the proper subtraction of nongauge-length deformation from measured crosshead deformation. In previous analysis of the tensile curves, the nongauge-length deformation was not correctly determined and subtracted from the crosshead displacement. The previously reported and revised tensile values for unirradiated alloys (20--700 C) are tabulated in Appendix A. The revised tensile values for the FFTF-irradiated (400--600 C) and HFIR-irradiated (400 C) alloys are tabulated in Appendix B, along with the neutron damage and helium levels. Appendix C compares the revised values to the previously reported values for irradiated alloys. Appendix D contains previous and revised values for the tensile properties of unirradiated V-5Cr-5Ti (BL-63) alloy exposed to oxygen.

  5. Study of clustering point defects under irradiation in dilute iron alloys; Etude de la formation sous irradiation des amas de defauts ponctuels dans les alliages ferritiques faiblement allies

    Energy Technology Data Exchange (ETDEWEB)

    Duong-Hardouin Duparc, T.H.A

    1998-12-31

    In low copper steels for nuclear reactor pressure vessel, point defect clustering plays an important role in hardening. These clusters are very small and invisible by transmission electron microscopy. In order to study the hardening component which results from the clustering of freely migrating point defects, we irradiated in a high voltage electron microscope Fe, the FeCu{sub 0.13%}, FeP{sub 0.015%} and FeN{sub 33ppm} alloys and the complex FeMn{sub 1.5%}Ni{sub 0.8%}Cu{sub 0.13%}P{sub 0.01%} alloy the composition of which is close to the matrix of pressure vessel steel. We studied the nucleation of dislocation loops and their growth velocity. The observations and the analyses have shown that in the complex model alloy, the interstitial dislocation loops are smaller and their density is more important than for the others alloys. The diffusion coefficients of interstitials and vacancies are obtained with the help of a simplified model. The densities of dislocation loops and their growth velocities obtained experimentally are reproduced by means of a cluster dynamics model we have developed. This is achieved self-consistently by using as a first trial the approximated coefficients obtained with the simplified model. The results of calculations have shown that the binding energy of di-interstitials must be very important in the binary iron alloys and only 0.95 eV in iron. Copper, nitrogen and phosphorus stabilize di-interstitials in iron. Finally the distribution of interstitial loops at 290 deg C and at 2.10{sup -9} dpa/s is calculated with the diffusion coefficient of point defects adjusted in FeCu. A distribution of small loops appears which gives an increase of hardening estimated to 10 Hv instead of 33 Hv experimentally observed. This low value can be improved by assuming in agreement with molecular dynamics simulations that a little fraction of di-interstitials is created at 2.5 MeV. (author) 111 refs.

  6. Void swelling and defect processes in ti-modified steels using accelerator irradiation

    International Nuclear Information System (INIS)

    The void swelling behaviour of (15Ni-14Cr)-O.25Ti and (15Ni-14Cr)-O.15Ti steels are studied using heavy ion irradiation for understanding the influence of titanium in the void swelling resistance of these D9 alloys. The cold worked samples have been pre-implanted with a uniform helium concentration of 30 appm spanning a width of about 640 nm. This was followed by a 5-MeV nickel ion irradiation to create a peak damage of ∼ 100 dpa at a damage rate of 7 x 10-3 dpa/s at various irradiation temperatures between 700 and 970 K. The gross swelling in the implanted range is measured by step height measurements. It is found that the peak swelling temperatures and the magnitude of swelling for the alloys are different. The difference in void swelling behaviour with variation in titanium concentration in these two alloys is discussed on the basis of the role of titanium on the vacancy migration and TiC precipitate formation. Isochronal annealing study of the positron lifetime in the un-irradiated alloys reveals different TiC precipitates formation behaviour in the two alloys. Ab initio calculations of positron lifetime, using large super-cells, show that C vacancies at the TiC/austenite interface are the predominant positron trapping centres in these alloys. (authors)

  7. Oxidation resistant high creep strength austenitic stainless steel

    Science.gov (United States)

    Brady, Michael P.; Pint, Bruce A.; Liu, Chain-Tsuan; Maziasz, Philip J.; Yamamoto, Yukinori; Lu, Zhao P.

    2010-06-29

    An austenitic stainless steel displaying high temperature oxidation and creep resistance has a composition that includes in weight percent 15 to 21 Ni, 10 to 15 Cr, 2 to 3.5 Al, 0.1 to 1 Nb, and 0.05 to 0.15 C, and that is free of or has very low levels of N, Ti and V. The alloy forms an external continuous alumina protective scale to provide a high oxidation resistance at temperatures of 700 to 800.degree. C. and forms NbC nanocarbides and a stable essentially single phase fcc austenitic matrix microstructure to give high strength and high creep resistance at these temperatures.

  8. Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

    Energy Technology Data Exchange (ETDEWEB)

    Gary S. Was

    2009-03-31

    The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.

  9. Quantitative Microanalysis with high Spatial Resolution: Application of FEG-DTEM XEDS Microanalysis to the Characterization of Complex Microstructures in Irradiated Low Alloy Steet

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.B., Watanabe, M. and Burke, M.G.

    2001-11-14

    To assist in the characterization of microstructural changes associated with irradiation damage in low alloy steels, the technique of quantitative x-ray mapping using a field emission gun scanning transmission electron microscope (FEG-STEM) equipped with an x-ray energy Dispersive spectrometer (XEDS) has been employed. Quantitative XEDS microanalyses of the matrix and grain boundaries of irradiated specimens have been compared with previous quantitative analyses obtained using 3D-Atom Probe Field-Ion Microscopy (3D-APFIM). In addition, the FEG-STEM XEDS maps obtained from the irradiated steel have revealed the presence of 2 to 3 nm Ni-enriched 'precipitates' in the matrix, which had previously been detected using 3D-APFIM. These quantitative FEG-STEM XEDS results represent the first direct and independent microchemical corroboration of the 3D-APFIM results showing ultra-fine irradiation-induced hardening features in low alloy steel.

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

    Science.gov (United States)

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

    2015-06-01

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

  11. Effects of 600 MeV proton irradiation on nucleation and growth of precipitates and helium bubbles in a high-purity Al-Mg-Si alloy

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Leffers, Torben; Victoria, M.;

    1986-01-01

    Solution treated specimens of a high-purity Al-0.75%Mg-0.42%Si alloy were irradiated with 600 MeV protons at 150 and 240°C to a dose level of 0.47 and 0.55 dpa, respectively. Mg2Si-type precipitates formed during irradiation at 150 and 240°C; at 240°C, however, a large number of precipitates seem...

  12. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C. (ITER R and D Task no. T213)[International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N.; Stubbins, J.F. [Illinois Univ., Dept. of Nuclear Engineering, Illinois (United States); Toft, P

    2000-03-15

    The fatigue behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of {approx}2.5 x 10{sup 17} n/m{sup 2}s (E> 1 MeV) to influence levels of 1.0 - 1.5 x 10{sup 24} n/m{sup 2} (E> 1 MeV) at 250 and 350 deg. C. These irradiations were carried out in temperature controlled rigs where the irradiation temperature was monitored and controlled continuously throughout the whole irradiation experiment. Both unirradiated and irradiated specimens were fatigue tested in vacuum at the irradiation temperatures of 250 and 350 deg. C in a strain controlled mode with a loading frequency of 0.5Hz. Post-fatigue microstructures were examined using transmission electron microscopy and the fracture surfaces were investigated using scanning electron microscope. The present investigations demonstrated that the fatigue life decreases with increasing temperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures and the fracture surface morphology. Finally, the main conclusions and their implications are summarised. (au)

  13. Phase-field Model for Interstitial Loop Growth Kinetics and Thermodynamic and Kinetic Models of Irradiated Fe-Cr Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Sun, Xin; Khaleel, Mohammad A.

    2011-06-15

    strength of interstitial loop for interstitials. In part II, we present a generic phase field model and discuss the thermodynamic and kinetic properties in phase-field models including the reaction kinetics of radiation defects and local free energy of irradiated materials. In particular, a two-sublattice thermodynamic model is suggested to describe the local free energy of alloys with irradiated defects. Fe-Cr alloy is taken as an example to explain the required thermodynamic and kinetic properties for quantitative phase-field modeling. Finally the great challenges in phase-field modeling will be discussed.

  14. Radiation effects on BCC metals and alloys. Progress report, January 1, 1975--November 30, 1975. [V ion-irradiated Mo; neutron-irradiated Nb, Nb-1Zr, W, and W--25 Re

    Energy Technology Data Exchange (ETDEWEB)

    Moteff, J.

    1975-12-04

    Experimental results of molybdenum specimens which were ion irradiated at several temperatures in the range of 700 to about 1500/sup 0/C at dpa levels of about 2 to 60 are discussed. TEM studies on Nb and Nb--1Zr irradiated at 1 x 10/sup 22/ n cm/sup -2/, E greater than 1 MeV at six temperatures in the range 430 to 1050/sup 0/C are presented. In addition, resistivity studies on W and W--25Re specimens irradiated under similar conditions as the Nb and Nb--1Zr specimens show significant changes coupled with void swelling in W but not in the W--25Re alloy. However, the W--25Re alloy showed that large precipitates have been formed thereby precluding the formation of voids. Finally, it has been shown that a hot-hardness tester may be effectively used as a strength microprobe in advanced alloy development and of irradiated metal and alloys where only small quantities of material are normally available. (auth)

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

    Science.gov (United States)

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

    2016-06-01

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

  16. Annihilation behaviour under electron irradiation of athermal ω-phase crystals formed by cooling at 131K in a β-Ti-Mo alloy

    International Nuclear Information System (INIS)

    Formation of athermal ω-phase crystals due to cooling to 131 K has been directly observed in a β-type Ti-15mass%Mo alloy. The athermal ω-phase crystals easily disappear by electron irradiation during the in-situ observation at 131 K. Incubation phenomenon of the annihilation is also recognized. The annihilation behaviour was investigated based on the dependence on electron irradiation conditions and incubation phenomena. It is concluded that the annihilation mechanism is concerned with interactive effects of temperature rise due to electron irradiation and collective oscillation resulted from inelastic scattering of electron beam.

  17. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Edwards, D.J.; Singh, B.N.; Bilde-Sørensen, Jørgen

    2005-01-01

    The formation of ‘cleared’ channels in neutron irradiated metals and alloys have been frequently reported for more than 40 years. So far, however, no unambiguous and conclusive evidence showing as to how and where these channels are initiated has emerged. In the following we present experimental ...

  18. Segregation and precipitation in iron-chromium alloys during thermal ageing and irradiation

    International Nuclear Information System (INIS)

    Iron-Chromium alloys have a peculiar thermodynamic and diffusion behavior which is due to their magnetic properties. The alloy decomposition under thermal ageing has been studied in this thesis. An atomistic kinetic model has been performed in this aim in which we have modeled in details the chemical species thermodynamic and diffusion properties. In particular, the evolution of elements diffusion properties which the ferro-paramagnetic transition has been introduced in the model. Simulated decompositions have been compared with experiments for a large range of concentrations and temperatures. A good agreement between simulations and experiments was observed and these comparisons have highlighted the ferro to paramagnetic transition key role in the concentrated alloys kinetic decomposition. This study has also evidenced that the elements diffusion at phases interfaces is responsible for the alloy decomposition kinetic in long lasting.We have also started a study of the alloy radiation induced segregation. For that purpose, atomistic kinetic model has been performed modeling defects migration through a perfect planar sink. It have been shown, I agreement with former studies, that chromium tends to segregate in the vicinity of sinks at low temperatures and deplete at high temperature. (author)

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

    Science.gov (United States)

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

    2014-05-01

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

  20. Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Long, Fei; Daymond, Mark R., E-mail: mark.daymond@queensu.ca; Yao, Zhongwen [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, Ontario K7L 3N6 (Canada); Kirk, Marquis A. [Material Science Division, Argonne National Laboratory, Illinois 60439 (United States)

    2015-03-14

    The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.

  1. Vloga in nastanek mikrostrukturnih sestavin M-A v zvarnih spojih maloogljičnih visokotrdnostnih konstrukcijskih jekel: The role and formation of martensite-austenite constituents in HSLA welded joints:

    OpenAIRE

    Praunseis, Zdravko; Toyoda, Masao; Križman, Alojz; Ohata, Mitsuru

    2001-01-01

    The existence of martensite-austenite constituents in the weld metal and heat-affected zone seriously reduces the fracture toughness of the welded joint. Therefore, we have investigated the formation of the martensite-austenite constituents when high-strength low-alloy steel is welded with a high heat input or using multi-pass welding. This paper deals with the effects of martensite-austenite constituents on the fracture toughness, the metallurgical features of the martensite-austenite consti...

  2. Irradiation effect of swift heavy ion for Zr{sub 50}Cu{sub 40}Al{sub 10} bulk glassy alloy

    Energy Technology Data Exchange (ETDEWEB)

    Onodera, Naoto; Ishii, Akito; Ishii, Kouji; Iwase, Akihiro [Department of Materials Science, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Yokoyama, Yoshihiko [Institute for Materials Research, Tohoku University, 2-1-1, Katahira, Aoba-ku, Sendai 980-8577 (Japan); Saitoh, Yuichi [Japan Atomic Energy Agency (JAEA), Takasaki Advanced Radiation Research Institute, 1233, Watanuki-machi, Takasaki, Gunma 370-1292 (Japan); Ishikawa, Norito [Japan Atomic Energy Agency (JAEA), Tokai Research and Development Center, Naka-ku, Ibaraki 319-1195 (Japan); Yabuuchi, Atsushi [Research Organization for the 21st Century, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan); Hori, Fuminobu, E-mail: horif@mtr.osakafu-u.ac.jp [Department of Materials Science, Osaka Prefecture University, 1-1, Gakuen-cho, Naka-ku, Sakai, Osaka 599-8531 (Japan)

    2013-11-01

    It has been reported that heavy ion irradiation causes softening in some cases of Zr-based bulk metallic glass alloys. However, the fundamental mechanisms of such softening have not been clarified yet. In this study, Zr{sub 50}Cu{sub 40}Al{sub 10} bulk glassy alloys were irradiated with heavy ions of 10 MeV I at room temperature. The maximum fluence was 3 × 10{sup 14} ions/cm{sup 2}. The positron annihilation measurements have performed before and after irradiation to investigate changes in free volume. We discuss the relationship between the energy loss and local open volume change after 10 MeV I irradiation compared with those obtained for 200 MeV Xe and 5 MeV Al. The energy loss analysis in ion irradiation for the positron lifetime has revealed that the decreasing trend of positron lifetime is well expressed as a function of total electronic energy deposition rather than total elastic energy deposition. It means that the positron lifetime change by the irradiation has a relationship with the inelastic collisions with electrons during heavy ion irradiation.

  3. One-step and rapid synthesis of high quality alloyed quantum dots (CdSe-CdS) in aqueous phase by microwave irradiation with controllable temperature

    International Nuclear Information System (INIS)

    In this paper, we presented a seed-mediated approach for rapid synthesis of high quality alloyed quantum dots (CdSe-CdS) in aqueous phase by microwave irradiation with controllable temperature in 1 h. In the synthesis, CdSe seeds were first formed by the reaction of NaHSe and Cd2+, and then alloyed quantum dots (CdSe-CdS) were rapidly produced by releasing of sulfide ions from 3-mercaptopropionic acid as sulfide source with microwave irradiation. The alloyed quantum dots synthesized had good optical properties, the quantum yield was up to 25%, and the full width at half maximum of the emission spectrum peak was about 28 nm. The as-prepared alloyed CdSe-CdS QDs were characterized by XRD, XPS and ICP-AES in order to explore the structure and component of the alloyed nanocrystals and the reaction mechanism. We speculate that the alloyed CdSe-CdS quantum dots may exist a gradient internal structure according to our preliminary results

  4. Effect of Primary Factor on Cavitation Resistance of Some Austenitic Metals

    Institute of Scientific and Technical Information of China (English)

    WANG Zai-you; ZHU Jin-hua

    2003-01-01

    The cavitation resistance of six kinds of austenitic metals was investigated using a rotating disc rig. The research results show that cavitation resistance of the austenitic metals is obviously raised due to cavitation-induced martensite and greatly influenced by mechanism of martensitic transformation. The cavitation resistance of two stress-induced martensite Fe-Mn-Si-Cr shape memory alloys is much better than that of three strain-induced martensite austenitic stainless steels. The Fe-Mn-Si-Cr shape memory alloys possess excellent cavitation resistance mainly because of their excellent elasticity in local small-zone. The first principal factor for cavitation resistance of metastable austenitic metals is unloaded rebounding depth, and the second one is energy dissipation resulted from cavitation-induced martensite.

  5. Phase stability in two-phase alloy Cu-4 at.% Ti under neutron irradiation

    International Nuclear Information System (INIS)

    X-ray diffraction analysis was used to investigate structural changes of phases in equilibrium two-phase copper-titanium alloy (4 at.%) depending on neutron fluence. It was established that lattice volume of α-phase increases linearly with fluence growth and lattice volume of orthorhombic β-phase-along th curve with the maximum at 1.6x1022nm-2 fluence. It is shown that the maximum of alloy hardening is observed at the maximal difference of α-and β-phase volumes

  6. Investigation of the thermo-mechanical behavior of neutron-irradiated Fe-Cr alloys by self-consistent plasticity theory

    Science.gov (United States)

    Xiao, Xiazi; Terentyev, Dmitry; Yu, Long; Bakaev, A.; Jin, Zhaohui; Duan, Huiling

    2016-08-01

    The thermo-mechanical behavior of non-irradiated (at 223 K, 302 K and 573 K) and neutron irradiated (at 573 K) Fe-2.5Cr, Fe-5Cr and Fe-9Cr alloys is studied by a self-consistent plasticity theory, which consists of constitutive equations describing the contribution of radiation defects at grain level, and the elastic-viscoplastic self-consistent method to obtain polycrystalline behaviors. Attention is paid to two types of radiation-induced defects: interstitial dislocation loops and solute rich clusters, which are believed to be the main sources of hardening in Fe-Cr alloys at medium irradiation doses. Both the hardening mechanism and microstructural evolution are investigated by using available experimental data on microstructures, and implementing hardening rules derived from atomistic data. Good agreement with experimental data is achieved for both the yield stress and strain hardening of non-irradiated and irradiated Fe-Cr alloys by treating dislocation loops as strong thermally activated obstacles and solute rich clusters as weak shearable ones.

  7. Expanded austenite in nitrided layers deposited on austenitic and super austenitic stainless steel grades

    International Nuclear Information System (INIS)

    In this work nitrided layers deposited on austenitic and super austenitic stainless steels were analyzed through optical microscopy and X-rays diffraction analysis (XRD). It was observed that the formation of N supersaturated phase, called expanded austenite, has promoted significant increment of hardness (> 1000HV). XRD results have indicated the anomalous displacement of the diffracted peaks, in comparison with the normal austenite. This behavior, combined with peaks broadening, it was analyzed in different nitriding temperatures which results showed good agreement with the literature. (author)

  8. Characteristics of the aluminum alloy plasma produced by a 1064 nm Nd:YAG laser with different irradiances

    Indian Academy of Sciences (India)

    W F Luo; X X Zhao; Q B Sun; C X Gao; J Tang; H J Wang; W Zhao

    2010-06-01

    The plasma generated by 1064 nm Nd:YAG laser irradiation of aluminum alloy in air at atmospheric pressure was studied spectroscopically. The electron density inferred by measuring the Stark-broadened line profile of Si(I) 288.16 nm decreases with increasing distance from the target surface. The electron temperature was determined using the Boltzmann plot method with nine strong neutral aluminum lines. Due to the thermal conduction towards the solid target and radiative cooling of the plasma as well as conversion of thermal energy into kinetic energy, the electron temperature decreases both at the plasma edge and close to the target surface. Electron density and electron temperature were also studied as functions of laser power density. At the same time, the validity of the assumption of local thermodynamic equilibrium and the effect of self-absorption were discussed in light of the results obtained.

  9. Temperature dependence of irradiation hardening due to dislocation loops and precipitates in RPV steels and model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kotrechko, S. [G.V. Kurdyumov Institute for Metal Physics of NAS of the Ukraine, Kiev, Ukraine, (Ukraine); Dubinko, V. [NSC Kharkov Institute of Physics and Technology, NAS of the Ukraine, Kharkov (Ukraine); Stetsenko, N. [G.V. Kurdyumov Institute for Metal Physics of NAS of the Ukraine, Kiev, Ukraine, (Ukraine); Terentyev, D. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol 2400 (Belgium); He, Xinfu [China Institute of Atomic Energy, 102413 Beijing (China); Sorokin, M. [National Research Centre ‘Kurchatov Institute‘, Kurchatov Sq. 1, 123182 Moscow (Russian Federation)

    2015-09-15

    A relative contribution to irradiation hardening caused by dislocation loops and solute-rich precipitates is established for RPV steels of WWER-440 and WWER-1000 reactors, based on TEM measurements and mechanical testing at reactor operating temperature of 563 K. The pinning strength factors evaluated for loops and precipitates are shown to be much lower than those obtained for model alloys based on the room temperature testing as well as those evaluated by means of atomistic simulations in the temperature range of 300–600 K. This discrepancy is explained in the framework of a model of thermally activated dislocation motion, which takes into account the difference in temperature and strain rate employed in atomistic simulations and in mechanical testing.

  10. Hot ductility of austenitic and duplex stainless steels under hot rolling conditions

    OpenAIRE

    Kömi, J. (Jenni)

    2001-01-01

    Abstract The effects of restoration and certain elements, nitrogen, sulphur, calcium and Misch metal, on the hot ductility of austenitic, high-alloyed austenitic and duplex stainless steels have been investigated by means of hot rolling, hot tensile, hot bending and stress relaxation tests. The results of these different testing methods indicated that hot rolling experiments using stepped specimens is the most effective way to investigate the relationship between the s...

  11. Deformation-induced austenite grain rotation and transformation in TRIP-assisted steel

    OpenAIRE

    Tirumalasetty, G.K.; van Huis, M.A.; Kwakernaak, C.; Sietsma, J.; Sloof, W.G.; Zandbergen, H. W.

    2012-01-01

    Uniaxial straining experiments were performed on a rolled and annealed Si-alloyed TRIP (transformation-induced plasticity) steel sheet in order to assess the role of its microstructure on the mechanical stability of austenite grains with respect to martensitic transformation. The transformation behavior of individual metastable austenite grains was studied both at the surface and inside the bulk of the material using electron back-scattered diffraction (EBSD) and X-ray diffraction (XRD) by de...

  12. Experimental Determination of the Primary Solidification Phase dependency on the solidification velocity for 17 different austenitic stainless steel compositions

    DEFF Research Database (Denmark)

    Laursen, Birthe Nørgaard; Olsen, Flemming Ove; Yardy, John;

    1997-01-01

    to the austenite phase.Most stainless steels are weldable by conventional welding techniques. However, during laser weldng the solidification velocities can be very much higher than by conventional welding techniques. By increasing the solidification velocity to a critical value known as the transition velocity......, the primary solidification phase is found to change from ferrite to austenite.A novel laser remelting technique has been modified to enable the transition velocity for laser welded austenitic stainless steels to be deermined experimentally and on the basis of results from 17 different alloy compositions...... an equation for the calculation of the transition velocity from alloy composition is proposed....

  13. Effet d'un enrichissement en nickel sur la stabilite mecanique de l'austenite de reversion lorsque soumise a de la fatigue oligocyclique

    Science.gov (United States)

    Godin, Stephane

    The effect of nickel enrichment on the mechanical stability of the reversed austenite contained in martensitic stainless steels 13%Cr-4%Ni and 13%Cr-6%Ni was investigated. The main objective of the study was to observe their microstructure and to compare the dynamic behaviour of the reversed austenite. Tempers made at different temperatures showed that the 6% Ni alloy began to form more austenite and at a lower temperature. SEM and TEM analysis were used to see the austenite and measure its chemical composition. It has been observed that it was richer in Ni than the surrounding martensite. This enrichment increased with tempering temperature and caused an impoverishment of the surrounding martensite. The study also showed that the chemical composition of the austenite formed at the peak (maximum) of both alloys was similar. For a same tempering, this suggests Ni can help to form more austenite but this austenite is not necessarily richer in Ni. The analysis also showed that the austenite was predominantly lamellar and located at the interface and/or inside the martensite laths. Low cycle fatigue tests have shown that the austenite of the 6% Ni alloy was the most mechanically stable even if its Ni content was lower than the 4% Ni alloy austenite. This behaviour was explained by a thinner and narrower morphology of this phase. For a different content of Ni and different quantity of austenite, the most mechanically stable one was in the 4% Ni alloy. It turned out that its reversed austenite was thinner and its surrounding martensite was a bit harder than the 6% Ni alloy austenite. The effect of Ni enrichment of an alloy would be beneficial regarding the mechanical stability if a suitable tempering is made. This tempering must form a thin lamellar austenite in a sufficiently hard martensite. More Ni in the austenite would not necessarily raise the mechanical stability. It could contribute but it seems that it is not be the main factor governing the mechanical stability

  14. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-03-28

    Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

  15. Effects of HVEM irradiation on ordered phases in Ni-Ti

    International Nuclear Information System (INIS)

    Various ordered phases in the Ni-Ti system were subjected to electron irradiation in the Berkeley HVEM. Austenitic NiTi (B2 structure) disorders and turns amorphous with room-temperature irradiations at accelerating potentials between 1 and 1.5 MeV. Total doses for the onset of amorphiticity range between 0.7 x 1022 and 3 x 1022 e.cm-2 (0.4 to 1.0dpa). At 90K the dose requirement decreases to 4 x 1020e.cm-2 (approx. 10-2dpa). Martensitic NiTi (distorted monoclinic structure) readily detwins and transforms to austenite when irradiated for short times (approx. 10 seconds). Vapor-deposited amorphous films were crystallized to produce NiTi, Phase X (ordered nickel-rich phase with unknown structure) and Ni3Ti (DO24 structure). Upon electron irradiation, NiTi and Phase X disorder and become amorphous, while Ni3Ti disorders but does not turn amorphous with doses up to 4 x 1022e.cm-2 at 90K. These results are discussed in terms of the requirement of a critical concentration of defects and their relative mobilities. Brimhall's solubility criteria for amorphization of ordered alloys by ion bombardment is apparantly applicable to electron-induced crystalline to amorphous transitions in this alloy

  16. Thermodynamic stability of austenitic Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2014-07-01

    Full Text Available The performed research was aimed at determining thermodynamic stability of structures of Ni-Mn-Cu cast iron castings. Examined were 35 alloys. The castings were tempered at 900 °C for 2 hours. Two cooling speeds were used: furnace-cooling and water-cooling. In the alloys with the nickel equivalent value less than 20,0 %, partial transition of austenite to martensite took place. The austenite decomposition ratio and the related growth of hardness was higher for smaller nickel equivalent value and was clearly larger in annealed castings than in hardened ones. Obtaining thermodynamically stable structure of castings requires larger than 20,0 % value of the nickel equivalent.

  17. Mechanical properties and microstructure of neutron irradiated cold worked Al-6063 alloy

    International Nuclear Information System (INIS)

    The impact of neutron irradiation on the mechanical properties and fracture morphology of cold worked Al-6063 were studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens (50 mm long and 6 mm wide gauge sections) were punched out from an Al-6063 23% cold worked tubes, which had been exposed to prolonged neutron irradiation of up to 4.5 x 1025 thermal neutrons/m2 (E -3 s-1. The uniform elongation and the ultimate tensile strength increase as functions of fluence. Metallographic examination and fractography reveal a decrease in the local area reduction of the final fracture necking. This reduction is accompanied with a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. No voids could be observed up to the maximum fluence. The dislocation density of cold worked Al decreases with the thermal neutron fluence. Prolonged annealing of unirradiated cold worked Al-6063 at 52 C revealed similar results. It thus appears that under our irradiation conditions the temperature during irradiation is the major factor influencing the mechanical properties and the microstructure during irradiation. (orig.)

  18. Microstructural evolutions of zirconium alloys under irradiation. Link with the irradiation growth phenomenon; Evolutions microstructurales des alliages de zirconium sous irradiation. Liens avec le phenomene de croissance

    Energy Technology Data Exchange (ETDEWEB)

    Simonot, C.

    1995-07-18

    This study deals with the irradiation-induced growth and microstructural evolutions of Zircaloy-4 type materials (ZrSn{sub 1.2-1.7} Fe{sub 0.18-0.24} Cr{sub 0.07-0.13} O{sub 0.09-0.15}), used as cladding and guide-tubes in PWR`s fuel assemblies. The main objective was to obtain a better understanding of the growth acceleration which may occur at high doses for the recrystallized metallurgical state. The elongation values of stress-free tubes irradiated at 400 deg in experimental reactors give clear indication of accelerated growth after a critical dose. Microstructural investigations reveal some large basal plane dislocation loops with vacancy character, which is an unexpected defect configuration for an hexagonal material with a c/a ratio less than the ideal value. In addition, a significant redistribution of iron and chromium solute elements comes from the dissolution of the initial Zr(Fe,Cr){sub 2} phases. In a guide-tube irradiated to high dose at 320 deg in a power reactor, a large density of these c-component loops was also observed in coincidence with a large iron re-solution due to the progressive partial amorphization of Laves phases. By contrast, as long as a negligible amount of iron is available in the matrix (start of progressive) amorphization at 350 deg or complete amorphization without any chemical change at 280 deg, only prism plane loops with interstitial and vacancy character are observed and the steady-state growth rate is low. A mechanism taking into account the Diffusional Anisotropy Difference of the radiation induced point defects seems to be the most suitable to explain the correlations between microstructural evolutions and growth rates. However it does not allow to predict the dose necessary for the formation of the basal plane loops responsible for the growth acceleration. (Abstract Truncated)

  19. Mechanistic understanding of irradiation-induced corrosion of zirconium alloys in nuclear power plants: Stimuli, status, and outlook

    International Nuclear Information System (INIS)

    Failures in the basic materials used in nuclear power plants continue to be costly and insidious, despite increasing industry vigilance to catch failures before they degrade safety. For instance, the overall costs to the US industry from materials problems could amount to as much as $10 billion annually. Moreover, estimates indicate that the cost of a pipe failure in a nuclear plant is one hundred times greater than the cost of a similar failure in a coal-fired plant. There are important practical stimuli and much scope for further understanding of the effects of irradiation on Zr-alloys (and other materials used in nuclear installations) by careful experimentation. Moreover, these studies need to address the effect of irradiation on all components of heterogeneous systems: the metal, the oxide and the environment, and especially those processes recurring at the interphases between these components. The present paper is aimed at providing specialists with some systematic information on the subject and with important considerations on the key items for further experimentation

  20. Combined nano-SIMS/AFM/EBSD analysis and atom probe tomography, of carbon distribution in austenite/ε-martensite high-Mn steels.

    Science.gov (United States)

    Seol, Jae-Bok; Lee, B-H; Choi, P; Lee, S-G; Park, C-G

    2013-09-01

    We introduce a new experimental approach for the identification of the atomistic position of interstitial carbon in a high-Mn binary alloy consisting of austenite and ε-martensite. Using combined nano-beam secondary ion mass spectroscopy, atomic force microscopy and electron backscatter diffraction analyses, we clearly observe carbon partitioning to austenite. Nano-beam secondary ion mass spectroscopy and atom probe tomography studies also reveal carbon trapping at crystal imperfections as identified by transmission electron microscopy. Three main trapping sites can be distinguished: phase boundaries between austenite and ε-martensite, stacking faults in austenite, and prior austenite grain boundaries. Our findings suggest that segregation and/or partitioning of carbon can contribute to the austenite-to-martensite transformation of the investigated alloy.

  1. Change of Cr atoms distribution in Fe85Cr15 alloy caused by 250 keV He+ ion irradiation to different doses

    International Nuclear Information System (INIS)

    Highlights: • Effect of He-ion irradiation dose on Fe85Cr15 alloy. • Irradiation-induced clustering of Cr atoms. • Irradiation-caused reorientation of the surface magnetization vector. • Irradiation-caused increase of Fe-site spin-density. - Abstract: Redistribution of Cr atoms in a Fe85Cr15 alloy caused by its irradiation with 250 keV He+ ions to different doses, D = 8 ⋅ 1016, 16 ⋅ 1016 and 48 ⋅ 1016 ions/cm2 was investigated by means of conversion electrons Mössbauer spectroscopy. The redistribution was expressed in terms of the Warren–Cowley short-range order parameters α1, α2 and α12 pertaining to the first (1NN), second (2NN) and both i.e. 1NN + 2NN shells, respectively. Clear evidence was found, both for non-irradiated and irradiated samples that the actual distribution of Cr atoms is characteristic of the shell, and for a given shell it depends on the irradiation dose. In particular, α1 is positive, hence indicates an under population of Cr atoms in 1NN with respect to the random case, α2 is negative, giving evidence thereby that 2NN is overpopulated by Cr atoms, and α12 is weakly positive. Under the applied irradiation the number of Cr atoms in both neighbor shells decreased signifying thereby a clustering of Cr atoms. The underlying decrease of Cr concentration within the 1NN–2NN volume around the probe Fe atoms was estimated at 1.5 at.% ranging between 2.1 for the lowest and 0.8 at.% for the highest dose

  2. Time dependency of the hydrophilicity and hydrophobicity of metallic alloys subjected to femtosecond laser irradiations

    International Nuclear Information System (INIS)

    Surfaces of metallic alloys were laser-processed with femtosecond laser pulses of 800 nm, with different power densities. The effect of time on the wettability of these surfaces was investigated. A multi-scale roughness made of undulations was created after the laser processing. This specific surface topography allowed the occurrence of a Wenzel's state. This state clearly explains the high hydrophilicity and hydrophobicity observed respectively one day after laser treatment and several days later. The change from hydrophilicity to hydrophobicity occurred over time and is due to surface chemistry modifications. The creation of new hydrophobic functional groups on aluminum alloy surface, for example, was proposed to be responsible for the hydrophobic behavior observed on these surfaces.

  3. Time dependency of the hydrophilicity and hydrophobicity of metallic alloys subjected to femtosecond laser irradiations

    Science.gov (United States)

    Bizi-bandoki, P.; Valette, S.; Audouard, E.; Benayoun, S.

    2013-05-01

    Surfaces of metallic alloys were laser-processed with femtosecond laser pulses of 800 nm, with different power densities. The effect of time on the wettability of these surfaces was investigated. A multi-scale roughness made of undulations was created after the laser processing. This specific surface topography allowed the occurrence of a Wenzel's state. This state clearly explains the high hydrophilicity and hydrophobicity observed respectively one day after laser treatment and several days later. The change from hydrophilicity to hydrophobicity occurred over time and is due to surface chemistry modifications. The creation of new hydrophobic functional groups on aluminum alloy surface, for example, was proposed to be responsible for the hydrophobic behavior observed on these surfaces.

  4. Irradiation experiments for the US/Japan collaborative testing program in HFIR and ORR

    International Nuclear Information System (INIS)

    The experiments in the US/Japan collaborative testing program for HFIR and ORR irradiate austenitic stainless steel candidate alloys for use as first-wall and blanket structural materials in fusion reactors. They will be irradiated with mixed-spectrum neutrons and with spectral tailoring to achieve helium-to-displacement-per-atom (he/dpa) ratios predicted for fusion reactor service. The assembly of JP-7 and -8 was completed. Capsules JP-2 through -8 are being irradiated in the HFIR-Capsule JP-1 completed its irradiation for a total of 336.40 d at 100 MW reactor power. The fabrication and assembly of ORR-MFE-6J are in progress. The assembly of MFE-7J is nearing completion. The specimens are expected to be delivered for loading by the end of March

  5. Effects of fluence and fluence rate of proton irradiation upon magnetism in Fe{sub 65}Ni{sub 35} Invar alloy

    Energy Technology Data Exchange (ETDEWEB)

    Matsushita, Masafumi, E-mail: matsushita.masafumi.me@ehime-u.ac.jp [Department of Mechanical Engineering, Ehime University, 3-Bunkyocho, Matsuyama 790-8977 (Japan); Wada, Hideki [Department of Mechanical Engineering, Ehime University, 3-Bunkyocho, Matsuyama 790-8977 (Japan); Matsushima, Yasushi [Department of Physics, Okayama University, 2-naka-tsushima, Kitaku, Okayama 700-8530 (Japan)

    2015-11-15

    Curie temperature, T{sub C}, of the Fe-Ni Invar alloys increase due to irradiation with electron and some kinds of ions. In this study, proton irradiation effects upon magnetism in an Fe{sub 65}Ni{sub 35} alloy have been investigated. It is found that the increment of T{sub C,} ∆T{sub C}, increases with increasing fluence. The magnetic hysteresis curve of the alloy was found to be unaffected by irradiation. Comparing ∆T{sub C} and the calculated energy transfer from the ions to the sample, it seemed that ∆T{sub C} was found to be related to the number of vacancies formed in nuclear collision events. In addition, ∆T{sub C} was influenced by the fluence rate, i.e., the deposited energy per unit time. - Highlights: • Proton irradiation effect on T{sub C} of Fe{sub 65}Ni{sub 35} was investigated. • Increment of T{sub C}, ∆T{sub C}, was confirmed in ion passed through and stopped samples. • The relationships among ∆T{sub C} and the deposited energy and vacancies were discussed. • It was reasonable to consider that ∆T{sub C} was related to the number of vacancies. • ∆T{sub C} was influenced by fluence rate, i.e. the energy deposition rate.

  6. Development of ODS ferritic-martensitic steels for application to high temperature and irradiation environment; Developpement d'une nouvelle nuance martensitique ODS pour utilisation sous rayonnement a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Lambard, V

    2000-07-01

    Iron oxide dispersion strengthened alloys are candidate for nuclear fuel cladding. Therefore, it is crucial to control their microstructure in order to optimise their mechanical properties at temperatures up to 700 deg C. The industrial candidates, ODS ferritic alloys, present an anisotropic microstructure which induces a weakening of mechanical properties in transversal direction as well as the precipitation of brittle phases under thermal aging and irradiation. For this purpose, we tried to develop a material with isotropic properties. We studied several 9Cr-1Mo ferritic/martensitic alloys, strengthened or not by oxide dispersion. The mechanical alloying was performed by attribution and powders were consolidated by hot extrusion. In this work, different metallurgical characterisation techniques and modelling were used to optimise a new martensitic ODS alloy. Microstructural and chemical characterization of matrix has been done. The effect of austenitizing and isochronal tempering treatments on microstructure and hardness has been studied. Oxide distribution, size and chemical composition have been studied before and after high temperature thermal treatment. The study of phase transformation upon heating has permitted the extrapolation to the equilibrium temperature formation of austenite. Phase transformation diagrams upon cooling have been determined and the transformation kinetics have been linked to austenite grain size by a simple relation. Fine grain size is unfavourable for the targeted application, so a particular thermal treatment inducing a coarser grain structure has been developed. Finally, tensile properties have been determined for the different microstructures. (author)

  7. Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    Radiation induced segregation (RIS) has been frequently reported in structural materials such as austenitic, ferritic, and ferritic-martensitic stainless steels (SS) that have been widely used in light water reactors (LWRs). RIS has been linked to secondary degradation effects in SS including irradiation induced stress corrosion cracking (IASCC). Earlier studies on thermal segregation in Fe based alloys found that metalloids elements such as P, S, Si, Ge, Sn etc. embrittle the materials when enrichment was observed at grain boundaries (GBs). RIS of Fe-Cr- Ni based austenitic steels has been modeled in the U.S. 2015 fiscal year (FY2015), which identified the pre-enrichment due to thermal segregation can have an important role on the subsequent RIS. The goal of this work is to develop thermal segregation models for alloying elements in steels for future integration with RIS modeling.

  8. Study of the precipitation and of the hardening microscopic mechanisms under irradiation in dilute ferritic alloys; Etude de la precipitation et des mecanismes microscopiques de durcissement sous irradiation dans des alliages ferritiques dilues

    Energy Technology Data Exchange (ETDEWEB)

    Mathon, M.H

    1995-07-01

    The copper precipitation plays a significant role in the embrittlement process of reactor vessel steels under neutron irradiation at 300 deg C. In order to understand the copper precipitation mechanisms, we have studied model ferritic binary FeCu and ternary alloys FeCuX (X=Mn,Ni, Cr, P). These materials have been either Irradiated with 2.5 MeV electrons In the 175-360 deg C temperature range or thermal aged at 500 deg C. The evolution of materials has been followed by resistivity measurements under irradiation, by small angle neutron scattering and by Vickers microhardness measurements. We have shown the similarity of copper precipitation under thermally ageing at 500 deg C and electron Irradiation at 300 deg C, in FeCu{sub 1,34%}. This result confirms that the main effect of electronic irradiation is to accelerate precipitation. Nevertheless, we have observed that irradiation induces an additional contribution to hardening attributed to point defect clusters. Concerning the ternary alloys, we observed that at 300 deg C the addition of a third element has no significant effect on the copper precipitation kinetic under irradiation but that at lower temperature manganese slows down precipitation kinetic. In order to reproduce the experimental results obtained on FeCu{sub 1,34%} by using a cluster kinetics model, we have to suppose that the precipitation is heterogeneous and controlled by interface reactions for the small size clusters. In addition, neutron or electron irradiated industrial steels have been studied by small angle neutron scattering. The results revealed the presence of nano-metric solute clusters which contain few copper atoms and which are not linked to the formation of displacement cascades. (author)

  9. Transmission electron microscopy investigation of the microstructure of Fe-Cr alloys induced by neutron and ion irradiation at 300 °C

    Science.gov (United States)

    Hernández-Mayoral, M.; Heintze, C.; Oñorbe, E.

    2016-06-01

    Four Fe-Cr binary alloys, with Cr content from 2.5 up to 12wt%, were neutron or ion irradiated up to a dose of 0.6 dpa at 300 °C. The microstructural response to irradiation has been characterised using Transmission Electron Microscopy (TEM). Both, neutrons and ions, gave rise to the formation of dislocation loops. The most striking difference between ion and neutron irradiation is the distribution of these loops in the sample. Except for the lowest Cr content, loops are distributed mainly along grain boundaries and dislocations in the neutron irradiated samples. The inhomogeneous distribution of dislocation loops could be related to the presence of α‧ precipitates in the matrix. In contrast, a homogeneous distribution is observed in all ion irradiated samples. This important difference is attributed to the orders of magnitude difference in dose rate between these two irradiation conditions. Moreover, the density of loops depends non-monotonically on Cr content in case of neutron irradiation, while it seems to increase with Cr content for ion implantation. Differences are also observed in terms of cluster size, with larger sizes for neutron irradiation than for ion implantation, again pointing towards an effect of the dose rate.

  10. Fabrication and irradiation of HFIR-MFE-JP-17, -18, and -19 target irradiation capsules

    International Nuclear Information System (INIS)

    The objective of this work is to fabricate and irradiate capsules for testing magnetic fusion energy (MFE) reactor candidate first-wall materials in the High Flux Isotope Reactor (HFIR) target positions. Japanese and US test specimens are being irradiated to determine fracture toughness of austenitic stainless steels and a few chromium ferritic steels and high heat flux alloys after irradiation to 3 dpa at temperatures of 60-125 and 250-300C. Fabrication and irradiation of three new uninstrumented HFIR target capsules for testing 12.5-mm-diameter stainless steel fracture toughness specimens to a damage level of approximately 3 displacements per atom (dpa) at temperatures of 60-125 and 250-300C are proceeding satisfactorily. Two low temperature capsules of identical design, designated HFIR-MFE-JP-18 and HFIR-MFE-JP-19, each contain 32 fracture toughness specimens directly cooled by reactor cooling water. Irradiation of these two capsules is nearing completion. A single helium-filled elevated temperature capsule, designated HFIR-MFE-JP-17, will contain a stack of 86 fracture toughness specimens. Additional neutronic calculations were required for this experiment to insure that it will not cause unacceptable neutron flux shifting and hot spots in HFIR fuel regions. Irradiation of this capsule is scheduled to start in late November, 1991. Included in each capsule are companion transmission electron microscopy (TEM) and SS-3 tensile specimens for correlation of microstructural, tensile, and fracture toughness properties

  11. Nitrogen bearing austenitic stainless steels for surgical implants

    Energy Technology Data Exchange (ETDEWEB)

    Tschiptschin, A.P.; Aidar, C.H.; Alonso-Falleiros, N. [Sao Paulo Univ. (Brazil). Escola Politecnica; Neto, F.B. [Instituto de Pesquisas Tecnologicas, Sao Paulo (Brazil)

    1999-07-01

    Nitrogen addition promotes substantial improvements on general and localized corrosion performance of stainless steels. In recent times high nitrogen (up to 0.6 wt%) and Mn bearing super austenitic stainless steel has been studied for medical applications due to its low Ni content, the so called body friendly alloys. 18%Cr, 0.4%N and 15%Mn stainless steels were cast either from electrolytic or commercial master alloys in induction furnace, forged, solubilized at 1423K for 3 hours and water quenched. Delta ferrite and carbide precipitate free structures were observed. (orig.)

  12. Effect of hydrogen on change carrier dissipation in 60Co irradiated by γ-quanta and non-alloyed n-type GaAs

    International Nuclear Information System (INIS)

    The pretreatment in hydrogen plasma (the hydrogenation) influences on the charge carrier dissipation processes in the non-alloyed gallium arsenide of n-type with no = (5...7) centre dot 1015 cm-3 and μo = (5...6) centre dot 10 13 cm2 / (V centre dot c) irradiated by γ-quantum 60Co was studied. The comparison of experimental dependence μ (T) with the designed one in the temperature range 77...291 K for non-hydrogenized and hydrogenized non irradiated and γ-quantum irradiated crystals was carried out. It is shown that the main dissipative mechanism that determine the charged carrier mobility in the non hydrogenized material is the dissipation on the charged centers - the radiation defects in the γ-quantum irradiated GaAs; the presence of double ionized defects is possible

  13. Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Busby, Jeremy T [ORNL; Gussev, Maxim N [ORNL

    2011-04-01

    Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be

  14. Failure of austenitic stainless steel tubes during steam generator operation

    Directory of Open Access Journals (Sweden)

    M. Głowacka

    2012-12-01

    Full Text Available Purpose: of this study is to analyze the causes of premature failure of steam generator coil made of austenitic stainless steel. Special attention is paid to corrosion damage processes within the welded joints.Design/methodology/approach: Examinations were conducted several segments of the coil made of seamless cold-formed pipes Ø 23x2.3 mm, of austenitic stainless steel grade X6CrNiTi18-10 according to EN 10088-1:2007. The working time of the device was 6 months. The reason for the withdrawal of the generator from the operation was leaks in the coil tube caused by corrosion damage. The metallographic investigations were performed with the use of light microscope and scanning electron microscope equipped with the EDX analysis attachment.Findings: Examinations of coil tubes indicated severe corrosion damages as pitting corrosion, stress corrosion cracking, and intergranular corrosion within base material and welded joints. Causes of corrosion was defined as wrong choice of austenitic steel grade, improper welding technology, lack of quality control of water supply and lack of surface treatment of stainless steel pipes.Research limitations/implications: It was not known the quality of water supply of steam generator and this was the reason for some problems in the identification of corrosion processes.Practical implications: Based on the obtained research results and literature studies some recommendations were formulated in order to avoid failures in the application of austenitic steels in the steam generators. These recommendations relate to the selection of materials, processing technology and working environment.Originality/value: Article clearly shows that attempts to increase the life time of evaporator tubes and steam coils by replacing non-alloy or low alloy structural steel by austenitic steel, without regard to restrictions on its use, in practice often fail.

  15. Influence of neutron irradiation on H diffusion in Zr-2.5Nb alloy

    International Nuclear Information System (INIS)

    Deuterium diffusion in Zr-2.5Nb pressure-tube material was measured in- and out-of-flux in the U-2 loop of the NRU reactor at Chalk River Laboratories. The results show that deuterium diffusivity at a neutron flux of about 5x1017 n.m-2s-1 is similar to that measured out-of-flux. This result suggests that deuterium diffusivities determined from out-reactor tests are sufficient for modeling deuterium ingress and its redistribution in pressure tubes and other reactor components made from zirconium alloys. (author)

  16. Effect of implantation defects on the corrosion of austenitic stainless steels in pressurized water reactor primary medium

    International Nuclear Information System (INIS)

    Internal parts of pressurized water reactor (PWR) vessels are often made of austenitic stainless steels (304L and 316L). These structural materials are exposed to an oxidizing medium under irradiation and mechanical stresses. Under these conditions, they can suffer damages by IASCC (Irradiation-Assisted Stress Corrosion Cracking). The first step in this cracking phenomenon is the initiation, which implies the breakdown of the passive layer. The nature and the structure of the oxide film formed on these steels are key factors in initiation of IASCC cracks. In this context, the objective of this work is first to better understand the oxidation mechanisms of stainless steels in primary medium and second to study the effects of irradiation induced defects on the oxide film formed on stainless steels in primary medium. Xenon ions and protons, were implanted in 316L-type austenitic stainless steel samples, respectively at an energy of 240 and 230 keV in order to simulate the irradiation defects. Implanted and non-implanted samples were exposed in a corrosion loop at 325 C to an aqueous medium containing 1000 ppm of boron, 2 ppm of lithium and 1,19.10-3 mol.L-1 of dissolved hydrogen. The samples were analyzed by TEM before and after exposure to primary medium in order to characterize both the defects generated by the implantation and the nature, structure, and morphology of the formed oxide. Comparing implanted and non-implanted samples has shown that the nature and the density of defects in the alloy subsurface played an important role on the composition (mainly on the content of Cr and Mo) and on the thickness of the inner layer. The study of the oxidation kinetics by coupling two ion beam analysis techniques (NRA and RBS) has revealed different behavior between the two types of samples: non-implanted and implanted. Tracer experiments (using D and 18O) were conducted to study the growth mechanism of the inner oxide layer and the associated transport mechanisms. The

  17. Shape memory alloy thaw sensors

    Science.gov (United States)

    Shahinpoor, Mohsen; Martinez, David R.

    1998-01-01

    A sensor permanently indicates that it has been exposed to temperatures exceeding a critical temperature for a predetermined time period. An element of the sensor made from shape memory alloy changes shape when exposed, even temporarily, to temperatures above the Austenitic temperature of the shape memory alloy. The shape change of the SMA element causes the sensor to change between two readily distinguishable states.

  18. Application Feasibility of PRE 50 grade Super Austenitic Stainless Steel as a Steam Generator Tubing

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Soo [Yonsei University, Seoul (Korea, Republic of); Kim, Young sik [Andong National University, Andong (Korea, Republic of); Kim, Taek Jun; Kim, Sun Tae; Park, Hui Sang [Yonsei University, Seoul (Korea, Republic of)

    1997-07-01

    The aim of this study is to evaluate the properties of the super austenitic stainless steel, SR-50A for application as steam generator tubing material. The microstructure, mechanical properties, corrosion properties, were analyzed and the results were compared between super austenitic stainless steel and Alloy 600 and Alloy 690. Super austenitic stainless steel, SR-50A is superior to Alloy 600, Alloy 690 and Alloy 800 in the mechanical properties(tensile strength, yield strength, and elongation). It was investigated that thermal conductivity of SR-50A was higher than Alloy 600. As a result of thermal treatment on super stainless steel, SR-50A, caustic SCC resistance was increased and its resistance was as much as Alloy 600TT and Alloy 690TT. In this study, optimum thermal treatment condition to improve the caustic corrosion properties was considered as 650 deg C or 550 deg C 15 hours. However, it is necessary to verify the corrosion mechanism and to prove the above results in the various corrosive environments. 27 refs., 6 tabs., 59 figs. (author)

  19. Microstructural studies of advanced austenitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Todd, J. A.; Ren, Jyh-Ching [University of Southern California, Los Angeles, CA (USA). Dept. of Materials Science

    1989-11-15

    This report presents the first complete microstructural and analytical electron microscopy study of Alloy AX5, one of a series of advanced austenitic steels developed by Maziasz and co-workers at Oak Ridge National Laboratory, for their potential application as reheater and superheater materials in power plants that will reach the end of their design lives in the 1990's. The advanced steels are modified with carbide forming elements such as titanium, niobium and vanadium. When combined with optimized thermo-mechanical treatments, the advanced steels exhibit significantly improved creep rupture properties compared to commercially available 316 stainless steels, 17--14 Cu--Mo and 800 H steels. The importance of microstructure in controlling these improvements has been demonstrated for selected alloys, using stress relaxation testing as an accelerated test method. The microstructural features responsible for the improved creep strengths have been identified by studying the thermal aging kinetics of one of the 16Ni--14Cr advanced steels, Alloy AX5, in both the solution annealed and the solution annealed plus cold worked conditions. Time-temperature-precipitation diagrams have been developed for the temperature range 600 C to 900 C and for times from 1 h to 3000 h. 226 refs., 88 figs., 10 tabs.

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

    Science.gov (United States)

    Kimoto, Takayoshi

    1993-08-01

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