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Sample records for irradiated austenitic steels

  1. Weldability of neutron irradiated austenitic stainless steels

    NASA Astrophysics Data System (ADS)

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

    1999-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2010-05-01

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

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

    SciTech Connect

    Stoller, R.E.

    1987-12-01

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

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

    SciTech Connect

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

    2008-01-21

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

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

    SciTech Connect

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

    1995-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  7. Irradiation assisted stress corrosion cracking of austenitic stainless steels

    SciTech Connect

    Was, G.S.; Atzmon, M.

    1990-06-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2000-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

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

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

    SciTech Connect

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

    2010-06-01

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

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

    SciTech Connect

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

    1999-05-21

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

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

    SciTech Connect

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

    1980-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2004-08-01

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

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

    SciTech Connect

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

    2004-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-05-01

    The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ0.2). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ0.2, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young's modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in "soft" grains with a high Schmid factor located near "stiff" grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to <0 0 1>||TA (tensile axis) and <1 0 1>||TA were twin free and grain with <1 1 1>||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

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

    NASA Astrophysics Data System (ADS)

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

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

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

    SciTech Connect

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

    1998-03-01

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

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

    DOE PAGESBeta

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

    2015-02-24

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

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

    SciTech Connect

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

    2015-02-24

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2007-02-01

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

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

    SciTech Connect

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

    1999-10-01

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

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

    SciTech Connect

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

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

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

    SciTech Connect

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

    1996-10-01

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

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

    SciTech Connect

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

    1996-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2011-02-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

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

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

    SciTech Connect

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

    2006-01-31

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

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

    NASA Astrophysics Data System (ADS)

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

    2007-03-01

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

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

    SciTech Connect

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

    1996-04-01

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

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

    SciTech Connect

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

    1986-01-01

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

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

    SciTech Connect

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

    2012-11-01

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

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

    SciTech Connect

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

    2006-01-31

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

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

    NASA Astrophysics Data System (ADS)

    Schibli, Raluca; Schäublin, Robin

    2013-11-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-03-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-02-01

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

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

    PubMed

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

    2016-12-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic steel 18Cr10NiTi pre-implanted at 295 K with deuterium ions in the dose range from 8 × 10(14) to 2.7 × 10(18) D/cm(2). The kinetics of structural transformation development in the steel layer was traced from deuterium thermodesorption spectra as a function of deuterium concentration. Three characteristic regions with different low rates of deuterium amount desorption as the implantation dose increases were revealed: I-the linear region of low implantation doses (up to 1 × 10(17) D/cm(2)); II-the nonlinear region of medium implantation doses (1 × 10(17) to 8 × 10(17) D/cm(2)); III-the linear region of high implantation doses (8 × 10(17) to 2.7 × 10(18) D/cm(2)). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps. Each of the discrete regions of deuterium retention coefficient variation corresponds to different implanted-matter states formed during deuterium ion implantation. The low-dose region is characterized by formation of deuterium-vacancy complexes and solid-solution phase state of deuterium in the steel. The total concentration of the accumulated deuterium in this region varies between 2.5 and 3 at.%. The medium-dose region is characterized by the radiation-induced action on the steel in the presence of deuterium with the resulting formation of the energy-stable nanosized crystalline structure of steel, having a developed network of intercrystalline boundaries. The basis for this developed network of intercrystalline boundaries is provided by the amorphous state, which manifests itself in the thermodesorption spectra as a widely temperature-scale extended region of deuterium desorption (structure formation with a varying activation energy). The total concentration of the accumulated deuterium in the region of medium implantation doses makes 7 to 8 at.%. The

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

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

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

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

    SciTech Connect

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

    1999-10-26

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

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

    SciTech Connect

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

    1999-07-16

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

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

    NASA Astrophysics Data System (ADS)

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

    2001-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2000-07-01

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

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

    SciTech Connect

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

    2010-02-16

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

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

    NASA Astrophysics Data System (ADS)

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

    1996-10-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2000-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1991-03-01

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

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

    NASA Astrophysics Data System (ADS)

    Maksimkin, O. P.

    2016-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2002-12-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1992-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2009-08-01

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

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

    SciTech Connect

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

    2012-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2004-08-01

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

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

    SciTech Connect

    Glass, R.S.

    1985-09-01

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

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

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

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

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

    NASA Astrophysics Data System (ADS)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

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

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

    SciTech Connect

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

    1990-10-01

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

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

    SciTech Connect

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

    1998-03-01

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

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

    SciTech Connect

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

    1997-04-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-09-01

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

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

    SciTech Connect

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

    1997-04-01

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

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

    NASA Astrophysics Data System (ADS)

    Maziasz, P. J.

    1992-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1992-09-01

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

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

    SciTech Connect

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

    1980-01-01

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

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

    SciTech Connect

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

    1997-08-01

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

  10. High Mn austenitic stainless steel

    DOEpatents

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

    2010-07-13

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

  11. Tensile properties of austenitic stainless steels and their weld joints after irradiation by the ORR-spectrally-tailoring experiment

    NASA Astrophysics Data System (ADS)

    Jitsukawa, S.; Maziasz, P. J.; Ishiyama, T.; Gibson, L. T.; Hishinuma, A.

    1992-09-01

    Tensile specimens of the Japanese heat of PCA (JPCA) and type 316 stainless steels were irradiated in spectrally tailored capsules in the Oak Ridge Research Reactor (ORR) to a peak dose of 7.4 dpa and a peak helium level of 105 appm in the temperature range between 328 and 673 K. Specimens of type 316 steel with weld joints produced by tungsten inert gas (TIG) and electron beam (EB) welding techniques were also included. Irradiation caused both increases in flow stress and decreases in elongation. Weld joint specimens exhibited both lower strength and elongation after irradiation. The reduction of area (RA) for the TIG weld joint specimens decreased by a factor of 5 compared to unirradiated base metal specimens, however, they still fractured in a ductile mode. The EB weld joints maintained RA levels similar to that of the unirradiated base metal specimens. Post-radiation ductilities of weld joints and base metal specimens of these steels should be adequate for their application to next generation fusion experimental devices, such as the International Tokamak Experimental Reactor (ITER).

  12. Microstructural evolution of welded austenitic stainless steels irradiated in the spectrally-tailored ORR experiment at 400$deg;C*1

    NASA Astrophysics Data System (ADS)

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

    1991-03-01

    Microstructural evolution of austenitic stainless steels and their welds has been examined after spectrally-tailored neutron irradiation. JPCA and 316W, containing 0.24 and 0.08 wt% of titanium, respectively, were electron-beam welded. TEM disks taken from these weld joints were irradiated in the ORR (Oak Ridge Research Reactor), to 7.4 dpa and almost 100 appm He. Base metal specimens of 316R with very low titanium content (0.005 wt%) were also irradiated. Specimens were examined by precision immersion densitometry before TEM observation. Only the 316R base metal showed measurable swelling by density change. Cavity swelling, determined by TEM observations in the base metals, was 0.29% for 316R, 0.06% for 316W and 0.03% for JPCA. Titanium effectively suppressed the cavity swelling of the base metals. The cellular microstructure of fusion zone remained after this irradiation both in JPCA and 316W with uniform distribution of cavities. Welding did not degrade the swelling resistance as measured either by immersion densitometry or TEM.

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

    SciTech Connect

    Was, G.S.; Atzmon, M.

    1990-06-01

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

  14. Explosive Surface Hardening of Austenitic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Kovacs-Coskun, T.

    2016-04-01

    In this study, the effects of explosion hardening on the microstructure and the hardness of austenitic stainless steel have been studied. The optimum explosion hardening technology of austenitic stainless steel was researched. In case of the explosive hardening used new idea mean indirect hardening setup. Austenitic stainless steels have high plasticity and can be easily cold formed. However, during cold processing the hardening phenomena always occurs. Upon the explosion impact, the deformation mechanism indicates a plastic deformation and this deformation induces a phase transformation (martensite). The explosion hardening enhances the mechanical properties of the material, includes the wear resistance and hardness. In case of indirect hardening as function of the setup parameters specifically the flayer plate position the hardening increased differently. It was find a relationship between the explosion hardening setup and the hardening level.

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

    SciTech Connect

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

    1999-10-01

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

  16. Austenitic stainless steels for cryogenic service

    SciTech Connect

    Dalder, E.N.C.; Juhas, M.C.

    1985-09-19

    Presently available information on austenitic Fe-Cr-Ni stainless steel plate, welds, and castings for service below 77 K are reviewed with the intent (1) of developing systematic relationships between mechanical properties, composition, microstructure, and processing, and (2) of assessing the adequacy of these data bases in the design, fabrication, and operation of engineering systems at 4 K.

  17. Effect of titanium doping on accumulation and annealing of radiation defects in austenitic steel 16Cr15Ni3Mo(0-1)Ti at low temperature (80 K) electron irradiation

    NASA Astrophysics Data System (ADS)

    Arbuzov, V. L.; Danilov, S. E.

    2016-02-01

    The effect of titanium doping on accumulation and annealing of radiation defects was investigated in austenitic stainless steel 16Cr15Ni3Mo under low temperature (80 K) electron irradiation. Steel has been taken in the quenched, aged and separation of solid solution states. The data obtained on the accumulation of radiation defects and their evolution during isochronous annealing. The types of defects and its complexes and the activation energy of the processes taking place with their participation is identified. The mechanisms of radiation- induced processes and the effect of titan doping is discussed.

  18. Cast alumina forming austenitic stainless steels

    DOEpatents

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

    2013-04-30

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

  19. Tailoring plasticity of austenitic stainless steels for nuclear applications: Review of mechanisms controlling plasticity of austenitic steels below 400 °C

    NASA Astrophysics Data System (ADS)

    Meric de Bellefon, G.; van Duysen, J. C.

    2016-07-01

    AISI 304 and 316 austenitic stainless steels were invented in the early 1900s and are still trusted by materials and mechanical engineers in numerous sectors because of their good combination of strength, ductility, and corrosion resistance, and thanks to decades of experience and data. This article is part of an effort focusing on tailoring the plasticity of both types of steels to nuclear applications. It provides a synthetic and comprehensive review of the plasticity mechanisms in austenitic steels during tensile tests below 400 °C. In particular, formation of twins, extended stacking faults, and martensite, as well as irradiation effects and grain rotation are discussed in details.

  20. Pitting corrosion resistant austenite stainless steel

    DOEpatents

    van Rooyen, D.; Bandy, R.

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

  1. Embrittlement of austenitic stainless steel welds

    SciTech Connect

    David, S.A.; Vitek, J.M.; Alexander, D.J.

    1995-06-01

    To prevent hot-cracking, austenitic stainless steel welds generally contain a small percent of delta ferrite. Although ferrite has been found to effectively prevent hot-cracking, it can lead to embrittlement of welds when exposed to elevated temperatures. The aging behavior of type-308 stainless steel weld has been examined over a range of temperatures 475--850 C for times up to 10,000 hrs. Upon aging, and depending on the temperature range, the unstable ferrite may undergo a variety of solid state transformations. These phase changes creep-rupture and Charpy impact properties.

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

    NASA Astrophysics Data System (ADS)

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

    2015-07-01

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

  3. The chemical composition of precipitated austenite in 9Ni steel

    NASA Astrophysics Data System (ADS)

    Fultz, B.; Kim, J. I.; Kim, Y. H.; Morris, J. W.

    1986-06-01

    Analytical scanning transmission electron microscopy and a novel Mössbauer spectrometry technique were used to measure the chemical composition of austenite particles which precipitate during intercritical tempering of 9Ni steel. Both techniques showed an enrichment of Ni, Mn, Cr, and Si in the austenite. A straightforward analysis involving data on both austenite composition and austenite formation kinetics suggests that the growth of austenite particles is controlled by a 3-dimensional diffusion process. The segregation of solutes to the austenite accounts for much of its stability against the martensitic transformation at low temperatures. Composition inhomogeneities develop in austenite particles after long temperings; the central regions of the particles are lean in solutes and are first to undergo the martensitic transformation. However, changes in solute concentrations of the austenite during long temperings seem too small to account for the large changes in austenite stability. It appears that some of the stability of precipitated austenite must be microstructural in origin.

  4. Instabilities in stabilized austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Ayer, Raghavan; Klein, C. F.; Marzinsky, C. N.

    1992-09-01

    The effect of aging on the precipitation of grain boundary phases in three austenitic stainless steels (AISI 347, 347AP, and an experimental steel stabilized with hafnium) was investigated. Aging was performed both on bulk steels as well as on samples which were subjected to a thermal treatment to simulate the coarse grain region of the heat affected zone (HAZ) during welding. Aging of the bulk steels at 866 K for 8000 hours resulted in the precipitation of Cr23C6 carbides, σ, and Fe2Nb phases; the propensity for precipitation was least for the hafnium-stabilized steel. Weld simulation of the HAZ resulted in dissolution of the phases present in the as-received 347 and 347AP steels, leading to grain coarsening. Subsequent aging caused extensive grain boundary Cr23C6 carbides and inhomogeneous matrix precipitation. In addition, steel 347AP formed a precipitate free zone (PFZ) along the grain boundaries. The steel containing hafnium showed the best microstructural stability to aging and welding.

  5. Embrittlement of austenitic stainless steel welds

    SciTech Connect

    David, S.A.; Vitek, J.M.

    1997-12-31

    The microstructure of type-308 austenitic stainless steel weld metal containing {gamma} and {delta} and ferrite is shown. Typical composition of the weld metal is Cr-20.2, Ni-9.4, Mn-1.7, Si-0.5, C-0.05, N-0.06 and balance Fe (in wt %). Exposure of austenitic stainless steel welds to elevated temperatures can lead to extensive changes in the microstructural features of the weld metal. On exposure to elevated temperatures over a long period of time, a continuous network of M{sub 23}C{sub 6} carbide forms at the austenite/ferrite interface. Upon aging at temperatures between 550--850 C, ferrite in the weld has been found to be unstable and transforms to sigma phase. These changes have been found to influence mechanical behavior of the weld metal, in particular the creep-rupture properties. For aging temperatures below 550 C the ferrite decomposes spinodally into {alpha} and {alpha}{prime} phases. In addition, precipitation of G-phase occurs within the decomposed ferrite. These transformations at temperatures below 550 C lead to embrittlement of the weld metal as revealed by the Charpy impact properties.

  6. Weldment for austenitic stainless steel and method

    DOEpatents

    Bagnall, Christopher; McBride, Marvin A.

    1985-01-01

    For making defect-free welds for joining two austenitic stainless steel mers, using gas tungsten-arc welding, a thin foil-like iron member is placed between the two steel members to be joined, prior to making the weld, with the foil-like iron member having a higher melting point than the stainless steel members. When the weld is formed, there results a weld nugget comprising melted and then solidified portions of the joined members with small portions of the foil-like iron member projecting into the solidified weld nugget. The portions of the weld nugget proximate the small portions of the foil-like iron member which project into the weld nugget are relatively rich in iron. This causes these iron-rich nugget portions to display substantial delta ferrite during solidification of the weld nugget which eliminates weld defects which could otherwise occur. This is especially useful for joining austenitic steel members which, when just below the solidus temperature, include at most only a very minor proportion of delta ferrite.

  7. Dependence of steady-state radiation swelling rate of l 0.1C-16Cr-15Ni-2Mo-2Mn-Ti-Si austenitic steel on dpa rate and irradiation temperature

    NASA Astrophysics Data System (ADS)

    Kozlov, А. V.; Portnykh, I. А.

    2009-04-01

    A large number of swelling measurement data on the 0.1C-16Cr-15Ni-2Mo-2Mn-Ti-Si austenitic steel used as a fuel cladding at temperatures 640-870 К in the BN-600 fast reactor were analyzed. It was found that within irradiation temperatures 690-830 К a steady-state swelling dose rate was from 0.45%/dpa to 1.1%/dpa. By the statistical model of point defect migration for the 0.1C-16Cr-15Ni-2Mo-2Mn-Ti-S steel the dependence of the steady-state swelling rate on the irradiation temperature and displacement rate was calculated. The calculation data were consistent with the experimental data.

  8. Effect of gamma irradiation on stress corrosion behavior of austenitic stainless steel under ITER-relevant conditions

    NASA Astrophysics Data System (ADS)

    Jones, R. H.; Henager, C. H.

    1992-09-01

    Stress corrosion crack growth tests were conducted on Type 316 SS and PCA sensitized to 5 C/cm2 at 100°C in deionized water with 10 ppm Cl-. A constant K test specimen was cylically loaded at 1 Hz with an R of 0.5 and a δK of 11 MPa√m in an autoclave immersed in a 60Co source. Tests were conducted at 0, 2.3 × 102, and 6.5 × 105 rad/h. The average crack velocities were found to be 2.0 and 1.5×105 mm/cycle for the Type 316 SS and PCA, respectively, in the absence of gamma irradiation and 1.3 and 0.74×10-5 mm/cycle, respectively, at both gamma fluxes. Gamma irradiation may have shifted the potential to more reducing rather than more oxidizing, as observed by others in high-temperature water with low O2 activity. This study suggests that there is no significant detrimental effect of gamma irradiation on the subcritical crack growth behavior of unirradiated Type 316 SS and PCA at ITER-relevant conditions.

  9. Microstructural studies of advanced austenitic steels

    SciTech Connect

    Todd, J. A.; Ren, Jyh-Ching

    1989-11-15

    This report presents the first complete microstructural and analytical electron microscopy study of Alloy AX5, one of a series of advanced austenitic steels developed by Maziasz and co-workers at Oak Ridge National Laboratory, for their potential application as reheater and superheater materials in power plants that will reach the end of their design lives in the 1990's. The advanced steels are modified with carbide forming elements such as titanium, niobium and vanadium. When combined with optimized thermo-mechanical treatments, the advanced steels exhibit significantly improved creep rupture properties compared to commercially available 316 stainless steels, 17--14 Cu--Mo and 800 H steels. The importance of microstructure in controlling these improvements has been demonstrated for selected alloys, using stress relaxation testing as an accelerated test method. The microstructural features responsible for the improved creep strengths have been identified by studying the thermal aging kinetics of one of the 16Ni--14Cr advanced steels, Alloy AX5, in both the solution annealed and the solution annealed plus cold worked conditions. Time-temperature-precipitation diagrams have been developed for the temperature range 600 C to 900 C and for times from 1 h to 3000 h. 226 refs., 88 figs., 10 tabs.

  10. Wear behavior of austenite containing plate steels

    NASA Astrophysics Data System (ADS)

    Hensley, Christina E.

    As a follow up to Wolfram's Master of Science thesis, samples from the prior work were further investigated. Samples from four steel alloys were selected for investigation, namely AR400F, 9260, Hadfield, and 301 Stainless steels. AR400F is martensitic while the Hadfield and 301 stainless steels are austenitic. The 9260 exhibited a variety of hardness levels and retained austenite contents, achieved by heat treatments, including quench and tempering (Q&T) and quench and partitioning (Q&P). Samples worn by three wear tests, namely Dry Sand/Rubber Wheel (DSRW), impeller tumbler impact abrasion, and Bond abrasion, were examined by optical profilometry. The wear behaviors observed in topography maps were compared to the same in scanning electron microscopy micrographs and both were used to characterize the wear surfaces. Optical profilometry showed that the scratching abrasion present on the wear surface transitioned to gouging abrasion as impact conditions increased (i.e. from DSRW to impeller to Bond abrasion). Optical profilometry roughness measurements were also compared to sample hardness as well as normalized volume loss (NVL) results for each of the three wear tests. The steels displayed a relationship between roughness measurements and observed wear rates for all three categories of wear testing. Nanoindentation was used to investigate local hardness changes adjacent to the wear surface. DSRW samples generally did not exhibit significant work hardening. The austenitic materials exhibited significant hardening under the high impact conditions of the Bond abrasion wear test. Hardening in the Q&P materials was less pronounced. The Q&T microstructures also demonstrated some hardening. Scratch testing was performed on samples at three different loads, as a more systematic approach to determining the scratching abrasion behavior. Wear rates and scratch hardness were calculated from scratch testing results. Certain similarities between wear behavior in scratch testing

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

    SciTech Connect

    M Hackett; G Was

    2005-08-12

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

  12. Comminuting irradiated ferritic steel

    DOEpatents

    Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

    1985-01-01

    Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

  13. Irradiation effects in ferritic steels

    NASA Astrophysics Data System (ADS)

    Lechtenberg, Thomas

    1985-08-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2006-12-01

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

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

    SciTech Connect

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

    2006-12-01

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

  16. Radiation resistant austenitic stainless steel alloys

    DOEpatents

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

    1987-02-11

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

  17. Radiation resistant austenitic stainless steel alloys

    DOEpatents

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

    1989-01-01

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

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

    DOEpatents

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

    1977-03-08

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

  19. The effect of dose rate on the response of austenitic stainless steels to neutron radiaiton

    SciTech Connect

    Allen, T. R.; Cole, J I.; Trybus, Carole L.; Porter, D. L.; Tsai, Hanchung; Garner, Francis A.; Kenik, E A.; Yoshitake, T.; Ohta, Joji

    2006-01-01

    Depending on reactor design and component location, austenitic stainless steels may experience significantly different irradiation dose rates in the same reactor. Understanding the effect of dose rate on radiation performance is important to predicting component lifetime. This study examined the effect of dose rate on swelling, grain boundary segregation, and tensile properties in austenitic stainless steels through the examination of components retrieved from the Experimental Breeder Reactor-II (EBR-II) following its shutdown. Annealed 304 stainless steel, stress-relieved 304 stainless steel, 12% cold-worked 316 stainless steel, and 20% cold-worked 316 stainless steel were irradiated over a dose range of 1-56 dpa at temperatures from 371 to 440 C and dose rates from 0.5 to 5.8 ? 10*7 dpa/s. Density and tensile properties were measured for 304 and 316 stainless steel. Changes in grain boundary composition were examined for 304 stainless steel. Swelling appears to increase at lower dose rates in both 304 and 316 stainless steel, although the effect was not always statistically significant. Grain boundary segregation also appears to increase at lower dose rate in 304 stainless steel. For the range of dose rates examined, no measurable dose rate effect on tensile properties was noted for any of the steels.

  20. Austenitic stainless steel for high temperature applications

    DOEpatents

    Johnson, Gerald D.; Powell, Roger W.

    1985-01-01

    This invention describes a composition for an austenitic stainless steel which has been found to exhibit improved high temperature stress rupture properties. The composition of this alloy is about (in wt. %): 12.5 to 14.5 Cr; 14.5 to 16.5 Ni; 1.5 to 2.5 Mo; 1.5 to 2.5 Mn; 0.1 to 0.4 Ti; 0.02 to 0.08 C; 0.5 to 1.0 Si; 0.01 maximum, N; 0.02 to 0.08 P; 0.002 to 0.008 B; 0.004-0.010 S; 0.02-0.05 Nb; 0.01-0.05 V; 0.005-0.02 Ta; 0.02-0.05 Al; 0.01-0.04 Cu; 0.02-0.05 Co; 0.03 maximum, As; 0.01 maximum, O; 0.01 maximum, Zr; and with the balance of the alloy being essentially iron. The carbon content of the alloy is adjusted such that wt. % Ti/(wt. % C+wt. % N) is between 4 and 6, and most preferably about 5. In addition the sum of the wt. % P+wt. % B+wt. % S is at least 0.03 wt. %. This alloy is believed to be particularly well suited for use as fast breeder reactor fuel element cladding.

  1. Weldable, age hardenable, austenitic stainless steel

    DOEpatents

    Brooks, J.A.; Krenzer, R.W.

    1975-07-22

    An age hardenable, austenitic stainless steel having superior weldability properties as well as resistance to degradation of properties in a hydrogen atmosphere is described. It has a composition of from about 24.0 to about 34.0 weight percent (w/o) nickel, from about 13.5 to about 16.0 w/o chromium, from about 1.9 to about 2.3 w/o titanium, from about 1.0 to about 1.5 w/ o molybdenum, from about 0.01 to about 0.05 w/o carbon, from about 0 to about 0.25 w/o manganese, from about 0 to about 0.01 w/o phosphorous and preferably about 0.005 w/o maximum, from about 0 to about 0.010 w/o sulfur and preferably about 0.005 w/o maximum, from about 0 to about 0.25 w/o silicon, from about 0.1 to about 0.35 w/o aluminum, from about 0.10 to about 0.50 w/o vanadium, from about 0 to about 0.0015 w/o boron, and the balance essentially iron. (auth)

  2. Solidification behavior of austenitic stainless steel filler metals

    SciTech Connect

    David, S.A.; Goodwin, G.M.; Braski, D.N.

    1980-02-01

    Thermal analysis and interrupted solidification experiments on selected austenitic stainless steel filler metals provided an understanding of the solidification behavior of austenitic stainless steel welds. The sequences of phase separations found were for type 308 stainless steel filler metal, L + L + delta + L + delta + ..gamma.. ..-->.. ..gamma.. + delta, and for type 310 stainless steel filler metal, L ..-->.. L + ..gamma.. ..-->.. ..gamma... In type 308 stainless steel filler metal, ferrite at room temperature was identified as either the untransformed primary delta-ferrite formed during the initial stages of solidification or the residual ferrite after Widmanstaetten austenite precipitation. Microprobe and scanning transmission electron microscope microanalyses revealed that solute extensively redistributes during the transformation of primary delta-ferrite to austenite, leading to enrichment and stabilization of ferrite by chromium. The type 310 stainless steel filler metal investigated solidifies by the primary crystallization of austenite, with the transformation going to completion at the solidus temperature. In our samples residual ferrite resulting from solute segregation was absent at the intercellular or interdendritic regions.

  3. Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

    SciTech Connect

    Maziasz, P.J.

    1983-01-01

    Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation.

  4. Influence of boron on void swelling in model austenitic steels

    NASA Astrophysics Data System (ADS)

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

    2004-08-01

    Model austenitic steels based on Fe-15Cr-16Ni with additions of 0.25Ti, 500 appm B, or 0.25Ti-500 appm B were irradiated in FFTF/MOTA over a wide range of dose rates at ˜400 °C. In addition to the effect of dose rate on swelling, it was desired to study the effect of boron addition to produce variations in He/dpa ratio. A strong effect of dose rate was observed, so strong that the relatively small distances separating the boron-free and doped alloys introduced a complication into the experiment. For specimens irradiated within the core, boron addition had no significant effect. For irradiations conducted near or outside the core edge, swelling appeared to be either enhanced or decreased by boron. The variability was a consequence of a strong dose rate effect overwhelming the influence of boron and helium. It is shown that helium exerted little influence relative to other important factors in these alloys.

  5. The mechanical stability of precipitated austenite in 9Ni steel

    NASA Astrophysics Data System (ADS)

    Fultz, B.; Morris, J. W.

    1985-12-01

    The strains inherent to the martensitic transformation of austenite particles in 9Ni steel create dislocation structures in the tempered martensite. These dislocation structures were studied by the complementary techniques of X-ray line profile analysis and transmission electron microscopy. The energy required to form these dislocation structures affects the thermodynamics of the transformation. We propose that changes in these dislocation structures reduce the “mechanical stability” of the austenite particles as they grow larger during isothermal tempering.

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

    SciTech Connect

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

    1996-12-01

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

  7. Austenite recrystallization and carbonitride precipitation in niobium microalloyed steels

    SciTech Connect

    Speer, J.G.; Hansen, S.S. )

    1989-01-01

    The response of austenite to thermomechanical treatment is investigated in two series of niobium microalloyed steels. Optical and electron metallographic techniques were used to follow the austenite recystallizaiton and carbonitride precipitation reactions in these steels. The first series of steels contained a constant level of 0.05Nb, with carbon levels varying from 0.008 to 0.25 pct. It was found that a lower carbon concentration results in faster austenite recrystallization due to a smaller carbonitride supersaturation which leads to a reduced precipitate nucleation rate. The second series of steels was designed with a constant carbonitride supersaturation by simultaneously varying the Nb and C concentrations while maintaining a constant solubility product. In these steels, the recrystallization kinetics increase as the volume fraction of Nb(C,N) is reduced and/or as the precipitate coarsening rate is increased. The volume fraction of carbonitrides increases as the Nb:(C + 12/14 N) ratio approaches the stoichiometric ratio of approximately 8:1. An experiment to determine whether Nb atoms dissolved in the austenite could exert a significant solute-drag effect on the recrystallization reaction indicated that 0.20Nb in solution could reduce the rate of recrystallization compared to a Nb-free C-Mn steel.

  8. Influence of Martensite Fraction on the Stabilization of Austenite in Austenitic-Martensitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Huang, Qiuliang; De Cooman, Bruno C.; Biermann, Horst; Mola, Javad

    2016-05-01

    The influence of martensite fraction ( f α') on the stabilization of austenite was studied by quench interruption below M s temperature of an Fe-13Cr-0.31C (mass pct) stainless steel. The interval between the quench interruption temperature and the secondary martensite start temperature, denoted as θ, was used to quantify the extent of austenite stabilization. In experiments with and without a reheating step subsequent to quench interruption, the variation of θ with f α' showed a transition after transformation of almost half of the austenite. This trend was observed regardless of the solution annealing temperature which influenced the martensite start temperature. The transition in θ was ascribed to a change in the type of martensite nucleation sites from austenite grain and twin boundaries at low f α' to the faults near austenite-martensite (A-M) boundaries at high f α'. At low temperatures, the local carbon enrichment of such boundaries was responsible for the enhanced stabilization at high f α'. At high temperatures, relevant to the quenching and partitioning processing, on the other hand, the pronounced stabilization at high f α' was attributed to the uniform partitioning of the carbon stored at A-M boundaries into the austenite. Reduction in the fault density of austenite served as an auxiliary stabilization mechanism at high temperatures.

  9. Stress corrosion cracking of austenitic stainless steel core internal welds.

    SciTech Connect

    Chung, H. M.; Park, J.-H.; Ruther, W. E.; Sanecki, J. E.; Strain, R. V.; Zaluzec, N. J.

    1999-04-14

    Microstructural analyses by several advanced metallographic techniques were conducted on austenitic stainless steel mockup and core shroud welds that had cracked in boiling water reactors. Contrary to previous beliefs, heat-affected zones of the cracked Type 304L, as well as 304 SS core shroud welds and mockup shielded-metal-arc welds, were free of grain-boundary carbides, which shows that core shroud failure cannot be explained by classical intergranular stress corrosion cracking. Neither martensite nor delta-ferrite films were present on the grain boundaries. However, as a result of exposure to welding fumes, the heat-affected zones of the core shroud welds were significantly contaminated by oxygen and fluorine, which migrate to grain boundaries. Significant oxygen contamination seems to promote fluorine contamination and suppress thermal sensitization. Results of slow-strain-rate tensile tests also indicate that fluorine exacerbates the susceptibility of irradiated steels to intergranular stress corrosion cracking. These observations, combined with previous reports on the strong influence of weld flux, indicate that oxygen and fluorine contamination and fluorine-catalyzed stress corrosion play a major role in cracking of core shroud welds.

  10. Bainitic stabilization of austenite in low alloy sheet steels

    NASA Astrophysics Data System (ADS)

    Brandt, Mitchell L.

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

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

    SciTech Connect

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

    2015-09-01

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

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

    NASA Astrophysics Data System (ADS)

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

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

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

    NASA Astrophysics Data System (ADS)

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

    2001-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2016-01-01

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

  15. Nanostructured nickel-free austenitic stainless steel/hydroxyapatite composites.

    PubMed

    Tulinski, Maciej; Jurczyk, Mieczyslaw

    2012-11-01

    In this work Ni-free austenitic stainless steels with nanostructure and their nanocomposites with hydroxyapatite are presented and characterized by means of X-ray diffraction and optical profiling. The samples were synthesized by mechanical alloying, heat treatment and nitriding of elemental microcrystalline powders with addition of hydroxyapatite (HA). In our work we wanted to introduce into stainless steel hydroxyapatite ceramics that have been intensively studied for bone repair and replacement applications. Such applications were chosen because of their high biocompatibility and ability to bond to bone. Since nickel-free austenitic stainless steels seem to have better mechanical properties, corrosion resistance and biocompatibility compared to 316L stainless steels, it is possible that composite made of this steel and HA could improve properties, as well. Mechanical alloying and nitriding are very effective technologies to improve the corrosion resistance of stainless steel. Similar process in case of nanocomposites of stainless steel with hydroxyapatite helps achieve even better mechanical properties and corrosion resistance. Hence nanocrystalline nickel-free stainless steels and nickel-free stainless steel/hydroxyapatite nanocomposites could be promising bionanomaterials for use as a hard tissue replacement implants, e.g., orthopedic implants. In such application, the surface roughness and more specifically the surface topography influences the proliferation of cells (e.g., osteoblasts). PMID:23421285

  16. An alternative to the crystallographic reconstruction of austenite in steels

    SciTech Connect

    Bernier, Nicolas; Bracke, Lieven; Malet, Loïc; Godet, Stéphane

    2014-03-01

    An alternative crystallographic austenite reconstruction programme written in Matlab is developed by combining the best features of the existing models: the orientation relationship refinement, the local pixel-by-pixel analysis and the nuclei identification and spreading strategy. This programme can be directly applied to experimental electron backscatter diffraction mappings. Its applicability is demonstrated on both quenching and partitioning and as-quenched lath-martensite steels. - Highlights: • An alternative crystallographic austenite reconstruction program is developed. • The method combines a local analysis and a nuclei identification/spreading strategy. • The validity of the calculated orientation relationship is verified on a Q and P steel. • The accuracy of the reconstructed microtexture is investigated on a martensite steel.

  17. Accurate modelling of anisotropic effects in austenitic stainless steel welds

    NASA Astrophysics Data System (ADS)

    Nowers, O. D.; Duxbury, D. J.; Drinkwater, B. W.

    2014-02-01

    The ultrasonic inspection of austenitic steel welds is challenging due to the formation of highly anisotropic and heterogeneous structures post-welding. This is due to the intrinsic crystallographic structure of austenitic steel, driving the formation of dendritic grain structures on cooling. The anisotropy is manifested as both a `steering' of the ultrasonic beam and the back-scatter of energy due to the macroscopic granular structure of the weld. However, the quantitative effects and relative impacts of these phenomena are not well-understood. A semi-analytical simulation framework has been developed to allow the study of anisotropic effects in austenitic stainless steel welds. Frequency-dependent scatterers are allocated to a weld-region to approximate the coarse grain-structures observed within austenitic welds and imaged using a simulated array. The simulated A-scans are compared against an equivalent experimental setup demonstrating excellent agreement of the Signal to Noise (S/N) ratio. Comparison of images of the simulated and experimental data generated using the Total Focusing Method (TFM) indicate a prominent layered effect in the simulated data. A superior grain allocation routine is required to improve upon this.

  18. Accurate modelling of anisotropic effects in austenitic stainless steel welds

    SciTech Connect

    Nowers, O. D.; Duxbury, D. J.; Drinkwater, B. W.

    2014-02-18

    The ultrasonic inspection of austenitic steel welds is challenging due to the formation of highly anisotropic and heterogeneous structures post-welding. This is due to the intrinsic crystallographic structure of austenitic steel, driving the formation of dendritic grain structures on cooling. The anisotropy is manifested as both a ‘steering’ of the ultrasonic beam and the back-scatter of energy due to the macroscopic granular structure of the weld. However, the quantitative effects and relative impacts of these phenomena are not well-understood. A semi-analytical simulation framework has been developed to allow the study of anisotropic effects in austenitic stainless steel welds. Frequency-dependent scatterers are allocated to a weld-region to approximate the coarse grain-structures observed within austenitic welds and imaged using a simulated array. The simulated A-scans are compared against an equivalent experimental setup demonstrating excellent agreement of the Signal to Noise (S/N) ratio. Comparison of images of the simulated and experimental data generated using the Total Focusing Method (TFM) indicate a prominent layered effect in the simulated data. A superior grain allocation routine is required to improve upon this.

  19. Hardenability of austenite in a dual-phase steel

    SciTech Connect

    Sarwar, M.; Priestner, R.

    1999-06-01

    A low-carbon, low-alloy steel was intercritically heat treated and thermomechanically processed to study the martensitic hardenability of austenite present. Rolling of the two-phase ({alpha} + {gamma}) microstructure elongated austenite particles and reduced their martensitic hardenability because the {alpha}/{gamma} interface where new ferrite forms during cooling was increased by the particle elongation. The martensite particles obtained in rolled material were also elongated or fibered in the rolling direction. Therefore, the thermomechanical processing of a two-phase ({alpha} + {gamma}) mixture has the detrimental effect of increasing the quenching power needed to yield a specific amount of martensite.

  20. Oxidation resistant high creep strength austenitic stainless steel

    DOEpatents

    Brady, Michael P.; Pint, Bruce A.; Liu, Chain-Tsuan; Maziasz, Philip J.; Yamamoto, Yukinori; Lu, Zhao P.

    2010-06-29

    An austenitic stainless steel displaying high temperature oxidation and creep resistance has a composition that includes in weight percent 15 to 21 Ni, 10 to 15 Cr, 2 to 3.5 Al, 0.1 to 1 Nb, and 0.05 to 0.15 C, and that is free of or has very low levels of N, Ti and V. The alloy forms an external continuous alumina protective scale to provide a high oxidation resistance at temperatures of 700 to 800.degree. C. and forms NbC nanocarbides and a stable essentially single phase fcc austenitic matrix microstructure to give high strength and high creep resistance at these temperatures.

  1. Phase control of austenitic chrome-nickel steel

    SciTech Connect

    Korkh, M. K. Davidov, D. I. Korkh, J. V. Rigmant, M. B. Nichipuruk, A. P. Kazantseva, N. V.

    2015-10-27

    The paper presents the results of the comparative study of the possibilities of different structural and magnetic methods for detection and visualization of the strain-induced martensitic phase in low carbon austenitic chromium-nickel steel. Results of TEM, SEM, optical microscopy, atomic and magnetic force microscopy, and magnetic measurements are presented. Amount of the magnetic strain-induced martensite was estimated. We pioneered magnetic force microscopic images of the single domain cluster distribution of the strain-induced martensite in austenite-ferrite materials.

  2. Intermetallic strengthened alumina-forming austenitic steels for energy applications

    NASA Astrophysics Data System (ADS)

    Hu, Bin

    In order to achieve energy conversion efficiencies of >50 % for steam turbines/boilers in power generation systems, materials required are strong, corrosion-resistant at high temperatures (>700°C), and economically viable. Austenitic steels strengthened with Laves phase and Ni3Al precipitates, and alloyed with aluminum to improve oxidation resistance, are potential candidate materials for these applications. The creep resistance of these alloys is significantly improved through intermetallic strengthening (Laves-Fe 2Nb + L12-Ni3Al precipitates) without harmful effects on oxidation resistance. This research starts with microstructural and microchemical analyses of these intermetallic strengthened alumina-forming austenitic steels in a scanning electron microscope. The microchemistry of precipitates, as determined by energy-dispersive x-ray spectroscopy and transmission electron microscope, is also studied. Different thermo-mechanical treatments were carried out to these stainless steels in an attempt to further improve their mechanical properties. The microstructural and microchemical analyses were again performed after the thermo-mechanical processing. Synchrotron X-ray diffraction was used to measure the lattice parameters of these steels after different thermo-mechanical treatments. Tensile tests at both room and elevated temperatures were performed to study mechanical behaviors of this novel alloy system; the deformation mechanisms were studied by strain rate jump tests at elevated temperatures. Failure analysis and post-mortem TEM analysis were performed to study the creep failure mechanisms of these alumina-forming austenitic steels after creep tests. Experiments were carried out to study the effects of boron and carbon additions in the aged alumina-forming austenitic steels.

  3. Materials compatibility of hydride storage materials with austenitic stainless steels

    SciTech Connect

    Clark, E.A.

    1992-09-21

    This task evaluated the materials compatibility of LaNi[sub 5-x]Al[sub x] (x= 0.3, 0.75) hydrides and palladium coated kieselguhr with austenitic stainless steel in hydrogen and tritium process environments. Based on observations of retired prototype hydride storage beds and materials exposure testing samples designed for this study, no materials compatibility problem was indicated. Scanning electron microscopy observations of features on stainless steel surfaces after exposure to hydrides are also commonly found on as-received materials before hydriding. These features are caused by either normal heat treating and acid cleaning of stainless steel or reflect the final machining operation.

  4. Materials compatibility of hydride storage materials with austenitic stainless steels

    SciTech Connect

    Clark, E.A.

    1992-09-21

    This task evaluated the materials compatibility of LaNi{sub 5-x}Al{sub x} (x= 0.3, 0.75) hydrides and palladium coated kieselguhr with austenitic stainless steel in hydrogen and tritium process environments. Based on observations of retired prototype hydride storage beds and materials exposure testing samples designed for this study, no materials compatibility problem was indicated. Scanning electron microscopy observations of features on stainless steel surfaces after exposure to hydrides are also commonly found on as-received materials before hydriding. These features are caused by either normal heat treating and acid cleaning of stainless steel or reflect the final machining operation.

  5. Materials compatibility of hydride storage materials with austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Clark, E. A.

    1992-09-01

    This task evaluated the materials compatibility of LaNi(5-x)Al(x) (x= 0.3, 0.75) hydrides and palladium coated kieselguhr with austenitic stainless steel in hydrogen and tritium process environments. Based on observations of retired prototype hydride storage beds and materials exposure testing samples designed for this study, no materials compatibility problem was indicated. Scanning electron microscopy observations of features on stainless steel surfaces after exposure to hydrides are also commonly found on as-received materials before hydriding. These features are caused by either normal heat treating and acid cleaning of stainless steel or reflect the final machining operation.

  6. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  7. A review of compatibility of IFR fuel and austenitic stainless steel

    SciTech Connect

    Keiser, D.D. Jr.

    1996-11-01

    Interdiffusion experiments have been conducted to investigate the compatibility of various austenitic stainless steels with U-Pu-Zr alloys, which are alloys to be employed as fuel for the Integral Fast Reactor being developed by Argonne National Laboratory. These tests have also studied the compatibility of austenitic stainless steels with fission products, like the minor actinides (Np and Am) and lanthanides (Ce and Nd), that are generated during the fission process in an IFR. This paper compares the results of these investigations in the context of fuel-cladding compatibility in IFR fuel elements, specifically focusing on the relative Interdiffusion behavior of the components and the types of phases that develop based on binary phase diagrams. Results of Interdiffusion tests are assessed in the light of observations derived from post-test examinations of actual irradiated fuel elements.

  8. Characterization of microstructure and texture across dissimilar super duplex/austenitic stainless steel weldment joint by austenitic filler metal

    SciTech Connect

    Eghlimi, Abbas; Shamanian, Morteza; Eskandarian, Masoomeh; Zabolian, Azam; Szpunar, Jerzy A.

    2015-08-15

    The evolution of microstructure and texture across an as-welded dissimilar UNS S32750 super duplex/UNS S30403 austenitic stainless steel joint welded by UNS S30986 (AWS A5.9 ER309LMo) austenitic stainless steel filler metal using gas tungsten arc welding process was evaluated by optical micrography and EBSD techniques. Due to their fabrication through rolling process, both parent metals had texture components resulted from deformation and recrystallization. The weld metal showed the highest amount of residual strain and had large austenite grain colonies of similar orientations with little amounts of skeletal ferrite, both oriented preferentially in the < 001 > direction with cub-on-cube orientation relationship. While the super duplex stainless steel's heat affected zone contained higher ferrite than its parent metal, an excessive grain growth was observed at the austenitic stainless steel's counterpart. At both heat affected zones, austenite underwent some recrystallization and formed twin boundaries which led to an increase in the fraction of high angle boundaries as compared with the respective base metals. These regions showed the least amount of residual strain and highest amount of recrystallized austenite grains. Due to the static recrystallization, the fraction of low degree of fit (Σ) coincident site lattice boundaries, especially Σ3 boundaries, was increased in the austenitic stainless steel heat affected zone, while the formation of subgrains in the ferrite phase increased the content of < 5° low angle boundaries at that of the super duplex stainless steel. - Graphical abstract: Display Omitted - Highlights: • Extensive grain growth in the HAZ of austenitic stainless steel was observed. • Intensification of < 100 > orientated grains was observed adjacent to both fusion lines. • Annealing twins with Σ3 CSL boundaries were formed in the austenite of both HAZ. • Cub-on-cube OR was observed between austenite and ferrite in the weld metal.

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

    NASA Astrophysics Data System (ADS)

    Dong, Dingqian; Chen, Fei; Cui, Zhenshan

    2016-01-01

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

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

    NASA Astrophysics Data System (ADS)

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

    2012-10-01

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

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

    SciTech Connect

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

    1997-04-01

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

  12. Magneto-mechanical effects in two steels with metastable austenite

    SciTech Connect

    Fultz, B.; Fior, G.O.; Chang, G.M.; Kopa, R.; Morris, J.W. Jr.

    1985-03-01

    Magneto-mechanical effects, or effects of magnetic fields on mechanical propeties of materials, are reported for two steels containing austenite that undergoes an fcc..-->..bcc martensitic transformation during plastic deformation. Stress-strain curves from tensile tests of AISI 304 stainless steels and 9Ni steels were measured in magnetic fields as large as 18 T at temperatures of 4/sup 0/K, 77/sup 0/K and room temperature. Even in 18 T magnetic fields at cryogenic temperatures the magneto-mechanical effects were small, but they were reproducible and scaled with the strength of the magnetic field and the amount of transformation. Magneto-mechanical effects in steels with metastable austenite provide a unique means of determining how martensitic transformations affect mechanical behavior. The fcc..-->..bcc transformation makes an important contribution to the work hardening of both AISI 304 and 9Ni steel, so the more rapid transformation during magnetic exposure results in a higher strength and a reduced elongation of tensile specimens. In AISI 304 stainless steel a reduced flow stress in the magnetic field was found at small plastic strains.

  13. Development of corrosion-resistant improved Al-doped austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi

    2011-10-01

    Aluminum-doped type 316L SS (316L/Al) has been developed for the purpose of suppressing the degradation of corrosion resistance induced by irradiation in austenitic stainless steels (SSs). The electrochemical corrosion properties of this material were estimated after Ni-ion irradiation at a temperature range from 330 °C to 550 °C. When irradiated at 550 °C up to 12 dpa, 316L/Al showed high corrosion resistance in the vicinity of grain boundaries (GBs) and in grains, while severe GB etching and local corrosion in grains were observed in irradiated 316L and 316 SS. It is supposed that aluminum enrichment was enhanced by high-temperature irradiation at GBs and in grains, to compensate for lost corrosion resistance induced by chromium depletion.

  14. Grain Boundary Strengthening in High Mn Austenitic Steels

    NASA Astrophysics Data System (ADS)

    Kang, Jee-Hyun; Duan, Shanghong; Kim, Sung-Joon; Bleck, Wolfgang

    2016-05-01

    The Hall-Petch relationship is investigated to find the yield strengths of two high Mn austenitic steels. The Hall-Petch coefficient is found to depend on the overall C concentration and cooling rate, which suggests that the C concentration at the grain boundaries is an important factor. The pile-up model suggests that C raises the stress for the dislocation emission, while the ledge model predicts that C increases the density of ledges which act as dislocation sources.

  15. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-01-01

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  16. Manganese-stabilized austenitic stainless steels for fusion applications

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1990-08-07

    An austenitic stainless steel that is comprised of Fe, Cr, Mn, C but no Ni or Nb and minimum N. To enhance strength and fabricability minor alloying additions of Ti, W, V, B and P are made. The resulting alloy is one that can be used in fusion reactor environments because the half-lives of the elements are sufficiently short to allow for handling and disposal.

  17. Mechanical properties of irradiated 9Cr-2WVTa steel

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.; Rieth, M.

    1998-09-01

    An Fe-9Cr-2W-0.25V-0.07Ta-0.1C (9Cr-2WVTa) steel has excellent strength and impact toughness before and after irradiation in the Fast Flux Test Facility and the High Flux Reactor (HFR). The ductile-brittle transition temperature (DBTT) increased only 32 C after 28 dpa at 365 C in FFTF, compared to a shift of {approx}60 C for a 9Cr-2WV steel--the same as the 9Cr-2WVTa steel but without tantalum. This difference occurred despite the two steels having similar tensile but without tantalum. This difference occurred despite the two steels having similar tensile properties before and after irradiation. The 9Cr-2WVTa steel has a smaller prior-austenite grain size, but otherwise microstructures are similar before irradiation and show similar changes during irradiation. The irradiation behavior of the 9Cr-2WVTa steel differs from the 9Cr-2WV steel and other similar steels in two ways: (1) the shift in DBTT of the 9Cr-2WVTa steel irradiated in FFTF does not saturate with fluence by {approx}28 dpa, whereas for the 9Cr-2WV steel and most similar steels, saturation occurs at <10 dpa, and (2) the shift in DBTT for 9Cr-2WVTa steel irradiated in FFTF and HFR increased with irradiation temperature, whereas it decreased for the 9Cr-2WV steel, as it does for most similar steels. The improved properties of the 9Cr-2WVTa steel and the differences with other steels were attributed to tantalum in solution.

  18. Evaluation of Microstructure and Mechanical Properties in Dissimilar Austenitic/Super Duplex Stainless Steel Joint

    NASA Astrophysics Data System (ADS)

    Rahmani, Mehdi; Eghlimi, Abbas; Shamanian, Morteza

    2014-10-01

    To study the effect of chemical composition on microstructural features and mechanical properties of dissimilar joints between super duplex and austenitic stainless steels, welding was attempted by gas tungsten arc welding process with a super duplex (ER2594) and an austenitic (ER309LMo) stainless steel filler metal. While the austenitic weld metal had vermicular delta ferrite within austenitic matrix, super duplex stainless steel was mainly comprised of allotriomorphic grain boundary and Widmanstätten side plate austenite morphologies in the ferrite matrix. Also the heat-affected zone of austenitic base metal comprised of large austenite grains with little amounts of ferrite, whereas a coarse-grained ferritic region was observed in the heat-affected zone of super duplex base metal. Although both welded joints showed acceptable mechanical properties, the hardness and impact strength of the weld metal produced using super duplex filler metal were found to be better than that obtained by austenitic filler metal.

  19. Researches upon the cavitation erosion behaviour of austenite steels

    NASA Astrophysics Data System (ADS)

    Bordeasu, I.; Popoviciu, M. O.; Mitelea, I.; Salcianu, L. C.; Bordeasu, D.; Duma, S. T.; Iosif, A.

    2016-02-01

    Paper analyzes the cavitation erosion behavior of two stainless steels with 100% austenitic structure but differing by the chemical composition and the values of mechanical properties. The research is based on the MDE(t) and MDER(t) characteristic curves. We studied supplementary the aspect of the eroded areas by other to different means: observations with performing optical microscopes and roughness measurements. The tests were done in the T2 vibratory facility in the Cavitation Laboratory of the Timisoara Polytechnic University. The principal purpose of the study is the identification of the elements influencing significantly the cavitation erosion resistance. It was established the effect of the principal chemical components (determining the proportion of the structural components in conformity the Schaffler diagram) upon the cavitation erosion resistance. The results of the researches present the influence of the proportion of unstable austenite upon cavitation erosion resistance. The stainless steel with the great proportion of unstable austenite has the best behavior. The obtained conclusion are important for the metallurgists which realizes the stainless steels used for manufacturing the runners of hydraulic machineries (turbines and pumps) with increased resistance to cavitation attack.

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

    NASA Astrophysics Data System (ADS)

    Gan, J.; Was, G. S.

    2001-08-01

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

  1. Market Opportunities for Austenitic Stainless Steels in SO2 Scrubbers

    NASA Astrophysics Data System (ADS)

    Michels, Harold T.

    1980-10-01

    Recent U.S. federal legislation has created new opportunities for SO2 scrubbers because all coals, even low-sulfur western coals, will probably require scrubbing to remove SO2 from gaseous combustion products. Scrubbing, the chemical absorption of SO2 by vigorous contact with a slurry—usually lime or limestone—creates an aggressive acid-chloride solution. This presents a promising market for pitting-resistant austenitic stainless steels, but there is active competition from rubber and fiberglass-lined carbon steel. Since the latter are favored on a first-cost basis, stainless steels must be justified on a cost/performance or life-cost basis. Nickel-containing austenitic alloys are favored because of superior field fabricability. Ferritic stainless steels have little utility in this application because of limitations in weldability and resulting poor corrosion resistance. Inco corrosion test spools indicate that molybdenum-containing austenitic alloys are needed. The leanest alloys for this application are 316L and 317L. Low-carbon grades of stainless steel are specified to minimize corrosion in the vicinity of welds. More highly alloyed materials may be required in critical areas. At present, 16,000 MW of scrubber capacity is operational and 17,000 MW is under construction. Another 29,000 MW is planned, bringing the total to 62,000 MW. Some 160,000 MW of scrubber capacity is expected to be placed in service over the next 10 years. This could translate into a total potential market of 80,000 tons of alloy plate for new power industry construction in the next decade. Retrofitting of existing power plants plus scrubbers for other applications such as inert gas generators for oil tankers, smelters, municipal incinerators, coke ovens, the pulp and paper industry, sulfuric acid plants, and fluoride control in phosphoric acid plants will add to this large market.

  2. Austenite Static Recrystallization Kinetics in Microalloyed B Steels

    NASA Astrophysics Data System (ADS)

    Larrañaga-Otegui, Ane; Pereda, Beatriz; Jorge-Badiola, Denis; Gutiérrez, Isabel

    2016-04-01

    Boron is added to steels to increase hardenability, substituting of more expensive elements. Moreover, B acts as a recrystallization delaying element when it is in solid solution. However, B can interact with N and/or C to form nitrides and carbides at high temperatures, limiting its effect on both phase transformation and recrystallization. On the other hand, other elements like Nb and Ti are added due to the retarding effect that they exert on the austenite softening processes, which results in pancaked austenite grains and refined room microstructures. In B steels, Nb and Ti are also used to prevent B precipitation. However, the complex interaction between these elements and its effect on the austenite microstructure evolution during hot working has not been investigated in detail. The present work is focused on the effect the B exerts on recrystallization when added to microalloyed steels. Although B on its own leads to retarded static recrystallization kinetics, when Nb is added a large delay in the static recrystallization times is observed in the 1273 K to 1373 K (1000 °C to 1100 °C) temperature range. The effect is larger than that predicted by a model developed for Nb-microalloyed steels, which is attributed to a synergistic effect of both elements. However, this effect is not so prominent for Nb-Ti-B steels. The complex effect of the composition on recrystallization kinetics is explained as a competition between the solute drag and precipitation pinning phenomena. The effect of the microalloying elements is quantified, and a new model for the predictions of recrystallization kinetics that accounts for the B and Nb+B synergetic effects is proposed.

  3. Austenite Static Recrystallization Kinetics in Microalloyed B Steels

    NASA Astrophysics Data System (ADS)

    Larrañaga-Otegui, Ane; Pereda, Beatriz; Jorge-Badiola, Denis; Gutiérrez, Isabel

    2016-06-01

    Boron is added to steels to increase hardenability, substituting of more expensive elements. Moreover, B acts as a recrystallization delaying element when it is in solid solution. However, B can interact with N and/or C to form nitrides and carbides at high temperatures, limiting its effect on both phase transformation and recrystallization. On the other hand, other elements like Nb and Ti are added due to the retarding effect that they exert on the austenite softening processes, which results in pancaked austenite grains and refined room microstructures. In B steels, Nb and Ti are also used to prevent B precipitation. However, the complex interaction between these elements and its effect on the austenite microstructure evolution during hot working has not been investigated in detail. The present work is focused on the effect the B exerts on recrystallization when added to microalloyed steels. Although B on its own leads to retarded static recrystallization kinetics, when Nb is added a large delay in the static recrystallization times is observed in the 1273 K to 1373 K (1000 °C to 1100 °C) temperature range. The effect is larger than that predicted by a model developed for Nb-microalloyed steels, which is attributed to a synergistic effect of both elements. However, this effect is not so prominent for Nb-Ti-B steels. The complex effect of the composition on recrystallization kinetics is explained as a competition between the solute drag and precipitation pinning phenomena. The effect of the microalloying elements is quantified, and a new model for the predictions of recrystallization kinetics that accounts for the B and Nb+B synergetic effects is proposed.

  4. Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of LWR and Test Reactor Spectral Environments

    SciTech Connect

    Garner, Francis A.; Oliver, Brian M.; Greenwood, Lawrence R.; Edwards, Danny J.; Bruemmer, Stephen M.; Grossbeck, Martin L.

    2002-03-31

    In fission and fusion reactor environments stainless steels generate significant amounts of helium and hydrogen by transmutation. The primary sources of helium are boron and nickel, interacting with both fast and especially thermal neutrons. Hydrogen arises primarily from fast neutron reactions, but is also introduced into steels at often much higher levels by other environmental processes. Although essentially all of the helium is retained in the steel, it is commonly assumed that most of the hydrogen is not retained. It now appears that under some circumstances, significant levels of hydrogen can be retained, especially when helium-nucleated cavities become a significant part of the microstructure. A variety of stainless steel specimens have been examined from various test reactors, PWRs and BWRs. These specimens were exposed to a wide range of neutron spectra with different thermal/fast neutron ratios. Pure nickel and pure iron have also been examined. It is shown that all major features of the retention of helium and hydrogen can be explained in terms of the composition, thermal/fast neutron ratio and the presence or absence of helium-nucleated cavities. In some cases, the hydrogen retention is very large and can exceed that generated by transmutation, with the additional hydrogen arising from either environmental sources and/or previously unidentified radioisotope sources that may come into operation at high neutron exposures.

  5. Solidification and solid state transformations of austenitic stainless steel welds

    SciTech Connect

    Brooks, J A; Williams, J C; Thompson, A W

    1982-05-01

    The microstructure of austenitic stainless steel welds can contain a large variety of ferrite morphologies. It was originally thought that many of these morphologies were direct products of solidification. Subsequently, detailed work on castings suggested the structures can solidify either as ferrite or austenite. However, when solidification occurs by ferrite, a large fraction of the ferrite transforms to austenite during cooling via a diffusion controlled transformation. It was also shown by Arata et al that welds in a 304L alloy solidified 70-80% as primary ferrite, a large fraction of which also transformed to austenite upon cooling. More recently it was suggested that the cooling rates in welds were sufficiently high that diffusionless transformations were responsible for several commonly observed ferrite morphologies. However, other workers have suggested that even in welds, delta ..-->.. ..gamma.. transformations are diffusion controlled. A variety of ferrite morphologies have more recently been characterized by Moisio and coworkers and by David. The purpose of this paper is to provide further understanding of the evaluation of the various weld microstructures which are related to both the solidification behavior and the subsequent solid state transformations. To accomplish this, both TEM and STEM (Scanning Transmission Electron Microscopy) techniques were employed.

  6. Microstructural observations of HFIR-irratiated austenitic stainless steels including welds from JP9-16

    SciTech Connect

    Sawai, T.; Shiba, K.; Hishinuma, A.

    1996-04-01

    Austenitic stainless steels, including specimens taken from various electron beam (EB) welds, have been irradiated in HFIR Phase II capsules, JP9-16. Fifteen specimens irradiated at 300, 400, and 500{degrees}C up to 17 dpa are so far examined by a transmission electron microscope (TEM). In 300{degrees}C irradiation, cavities were smaller than 2nm and different specimens showed little difference in cavity microstructure. At 400{degrees}C, cavity size was larger, but still very small (<8 nm). At 500{degrees}C, cavity size reached 30 nm in weld metal specimens of JPCA, while cold worked JPCA contained a small (<5 nm) cavities. Inhomogeneous microstructural evolution was clearly observed in weld-metal specimens irradiated at 500{degrees}C.

  7. Reducing tool wear when machining austenitic stainless steels

    SciTech Connect

    Magee, J.H.; Kosa, T.

    1998-07-01

    Austenitic stainless steels are considered more difficult to machine than carbon steels due to their high work hardening rate, large spread between yield and ultimate tensile strength, high toughness and ductility, and low thermal conductivity. These characteristics can result in a built-up edge or excessive tool wear during machining, especially when the cutting speed is too high. The practical solution is to lower the cutting speed until tool life reaches an acceptable level. However, lower machining speed negatively impacts productivity. Thus, in order to overcome tool wear at relatively high machining speeds for these alloys, on-going research is being performed to improve cutting fluids, develop more wear-resistant tools, and to modify stainless steels to make them less likely to cause tool wear. This paper discusses compositional modifications to the two most commonly machined austenitic stainless steels (Type 303 and 304) which reduced their susceptibility to tool wear, and allowed these grades to be machined at higher cutting speeds.

  8. High temperature properties of an austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Nikulin, I.; Kaibyshev, R.; Skorobogatykh, V.

    2010-07-01

    Tensile properties of the 18Cr-9Ni-W-Nb-V-N austenitic stainless steel were studied at strain rates ranging from 6.7×10-6 to 1.3×10-2 s-1 in the temperature interval 20-740°C. It was found that this steel exhibits jerky flow at temperatures ranging from 530 to 680°C and an initial strain rate of 1.3×10-3 s-1. This phenomenon was interpreted in terms of Portevin-Le Chatelier (PLC) effect occurring due to dynamic strain aging (DSA). PLC yields significant increase in high temperature strength of this steel due to extending of plateau on temperature dependence of yield strength (YS) and ultimate tensile strength (UTS) to higher temperatures. As a result, YS and UTS remain virtually unchanged with increasing temperature from 350 to 740°C. Role of additives of tungsten and vanadium in DSA and high temperatures strength of the austenitic stainless steel is discussed.

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

    SciTech Connect

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

    1996-10-01

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

  10. Effects of Retained Austenite Stability and Volume Fraction on Deformation Behaviors of TRIP Steels

    SciTech Connect

    Choi, Kyoo Sil; Soulami, Ayoub; Liu, Wenning N.; Sun, Xin; Khaleel, Mohammad A.

    2010-10-02

    In this paper, the separate effects of austenite stability and its volume fraction on the deformation behaviors of transformation-induced plasticity (TRIP) steels are investigated based on the microstructure-based finite element modeling method. The effects of austenite stability on the strength, ductility and formability of TRIP steels are first examined based on the microstructure of a commercial TRIP 800 steel. Then, the separate effects of the austenite volume fraction on the overall deformation behaviors of TRIP steels are examined based on the various representative volume elements (RVEs). The computational results suggest that the higher austenite stability is helpful to increase the ductility and formability, but not the UTS. However, the increase of austenite volume fraction alone is not helpful in improving the performance of TRIP steels. This may indicate that various other material factors should also be concurrently adjusted during thermo-mechanical manufacturing process in a way to increase the performance of TRIP steels, which needs further investigation.

  11. Hydrogen vibrations in austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Danilkin, S. A.; Delafosse, D.; Fuess, H.; Gavriljuk, V. G.; Ivanov, A.; Magnin, T.; Wipf, H.

    The vibrational modes of hydrogen in fcc Fe-25Cr-20Ni stainless steel with a hydrogen content of 0.33at.% were studied by neutron spectroscopy. Hydrogen doping was performed at 810K in a hydrogen-gas atmosphere of 190bar. Neutron spectra were taken at 2K and 77K with the spectrometer IN1-BeF (ILL, Grenoble). The spectra show the fundamental hydrogen vibration at 130 meV and the second harmonics at 260 meV. The frequencies are higher than in other fcc hydrides. In spite of the cubic symmetry of the octahedral hydrogen positions and the low hydrogen content, the inelastic hydrogen peak has a relatively large width and an asymmetric shape.

  12. Microstructures of laser deposited 304L austenitic stainless steel

    SciTech Connect

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

    2000-05-22

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

  13. Development of Alumina-Forming Austenitic Stainless Steels

    SciTech Connect

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

    2009-01-01

    This paper presents the results of the continued development of creep-resistant, alumina-forming austenitic (AFA) stainless steel alloys, which exhibit a unique combination of excellent oxidation resistance via protective alumina (Al2O3) scale formation and high-temperature creep strength through the formation of stable nano-scale MC carbides and intermetallic precipitates. Efforts in fiscal year 2009 focused on the characterization and understanding of long-term oxidation resistance and tensile properties as a function of alloy composition and microstructure. Computational thermodynamic calculations of the austenitic matrix phase composition and the volume fraction of MC, B2-NiAl, and Fe2(Mo,Nb) base Laves phase precipitates were used to interpret oxidation behavior. Of particular interest was the enrichment of Cr in the austenitic matrix phase by additions of Nb, which aided the establishment and maintenance of alumina. Higher levels of Nb additions also increased the volume fraction of B2-NiAl precipitates, which served as an Al reservoir during long-term oxidation. Ageing studies of AFA alloys were conducted at 750 C for times up to 2000 h. Ageing resulted in near doubling of yield strength at room temperature after only 50 h at 750 C, with little further increase in yield strength out to 2000 h of ageing. Elongation was reduced on ageing; however, levels of 15-25% were retained at room temperature after 2000 h of total ageing.

  14. Retained austenite thermal stability in a nanostructured bainitic steel

    SciTech Connect

    Avishan, Behzad; Garcia-Mateo, Carlos; Yazdani, Sasan; Caballero, Francisca G.

    2013-07-15

    The unique microstructure of nanostructured bainite consists of very slender bainitic ferrite plates and high carbon retained austenite films. As a consequence, the reported properties are opening a wide range of different commercial uses. However, bainitic transformation follows the T{sub 0} criteria, i.e. the incomplete reaction phenomena, which means that the microstructure is not thermodynamically stable because the bainitic transformation stops well before austenite reaches an equilibrium carbon level. This article aims to study the different microstructural changes taking place when nanostructured bainite is destabilized by austempering for times well in excess of that strictly necessary to end the transformation. Results indicate that while bainitic ferrite seems unaware of the extended heat treatment, retained austenite exhibits a more receptive behavior to it. - Highlights: • Nanostructured bainitic steel is not thermodynamically stable. • Extensive austempering in these microstructures has not been reported before. • Precipitation of cementite particles is unavoidable at longer austempering times. • TEM, FEG-SEM and XRD analysis were used for microstructural characterization.

  15. Corrosion resistance of kolsterised austenitic 304 stainless steel

    SciTech Connect

    Abudaia, F. B. Khalil, E. O. Esehiri, A. F. Daw, K. E.

    2015-03-30

    Austenitic stainless suffers from low wear resistance in applications where rubbing against other surfaces is encountered. This drawback can be overcome by surface treatment such as coating by hard materials. Other treatments such as carburization at relatively low temperature become applicable recently to improve hardness and wear resistance. Carburization heat treatment would only be justified if the corrosion resistance is unaffected. In this work samples of 304 stainless steels treated by colossal supersaturation case carburizing (known as Kolsterising) carried out by Bodycote Company was examined for pitting corrosion resistance at room temperature and at 50 °C. Comparison with results obtained for untreated samples in similar testing conditions show that there is no deterioration in the pitting resistance due to the Kolsterising heat treatment. X ray diffraction patterns obtained for Kolsterising sample showed that peaks correspond to the austenite phase has shifted to lower 2θ values compared with those of the untreated sample. The shift is an indication for expansion of austenite unit cells caused by saturation with diffusing carbon atoms. The XRD of Kolsterising samples also revealed additional peaks appeared in the patterns due to formation of carbides in the kolsterised layer. Examination of these additional peaks showed that these peaks are attributed to a type of carbide known as Hagg carbide Fe{sub 2}C{sub 5}. The absence of carbides that contain chromium means that no Cr depletion occurred in the layer and the corrosion properties are maintained. Surface hardness measurements showed large increase after Kolsterising heat treatment.

  16. Corrosion resistance of kolsterised austenitic 304 stainless steel

    NASA Astrophysics Data System (ADS)

    Abudaia, F. B.; Khalil, E. O.; Esehiri, A. F.; Daw, K. E.

    2015-03-01

    Austenitic stainless suffers from low wear resistance in applications where rubbing against other surfaces is encountered. This drawback can be overcome by surface treatment such as coating by hard materials. Other treatments such as carburization at relatively low temperature become applicable recently to improve hardness and wear resistance. Carburization heat treatment would only be justified if the corrosion resistance is unaffected. In this work samples of 304 stainless steels treated by colossal supersaturation case carburizing (known as Kolsterising) carried out by Bodycote Company was examined for pitting corrosion resistance at room temperature and at 50 °C. Comparison with results obtained for untreated samples in similar testing conditions show that there is no deterioration in the pitting resistance due to the Kolsterising heat treatment. X ray diffraction patterns obtained for Kolsterising sample showed that peaks correspond to the austenite phase has shifted to lower 2θ values compared with those of the untreated sample. The shift is an indication for expansion of austenite unit cells caused by saturation with diffusing carbon atoms. The XRD of Kolsterising samples also revealed additional peaks appeared in the patterns due to formation of carbides in the kolsterised layer. Examination of these additional peaks showed that these peaks are attributed to a type of carbide known as Hagg carbide Fe2C5. The absence of carbides that contain chromium means that no Cr depletion occurred in the layer and the corrosion properties are maintained. Surface hardness measurements showed large increase after Kolsterising heat treatment.

  17. Heavy hydrogen isotopes penetration through austenitic and martensitic steels

    NASA Astrophysics Data System (ADS)

    Dolinski, Yu.; Lyasota, I.; Shestakov, A.; Repritsev, Yu.; Zouev, Yu.

    2000-12-01

    Experimental results are presented of deuterium and tritium permeability through samples of nickel, austenitic steel (16Cr-15Ni-3Mo-Ti), and martensitic steel DIN 1.4914 (MANET) exposed to a gaseous phase. Experiments were carried out at the RFNC-VNHTF installation, which has the capability of measuring the permeability of hydrogen isotopes by mass spectrometry over a temperature range of 293-1000 K, hydrogen isotope pressure ranges of 50-1000 Pa. Sample disks (30 and 40 mm diam.) can be assembled in the test chamber by electron-beam welding or mounted (30-mm diam. disks) on gaskets. Diffusion and permeability dependencies on temperature and pressure are determined and corresponding activation energies are presented.

  18. Strength of nanostructured austenitic steel 316LN at cryogenic temperatures

    NASA Astrophysics Data System (ADS)

    Czarkowski, P.; Krawczynska, A. T.; Brynk, T.; Nowacki, M.; Lewandowska, M.; Kurzydłowski, K. J.

    2014-12-01

    The aim of this work was to investigate the effect of nano-refinement on the properties of austenitic steel. The material with the initial grain size of 40-50pm was subjected to hydrostatic extrusion at a room temperature to the total accumulated strain exceeding 1. The microstructure developed was investigated by Transmission Electron Microscopy (TEM) and Focus Ion Beam (FIB). The strength of the extruded samples was tested at 293K, 77K and 4.2K by means of cryostat for static tensile tests. The results show that the hydrostatically extruded steel 316LN has excellent strength in cryogenic conditions, which make this material interesting for applications in cryogenic devices.

  19. Tensile behavior of irradiated manganese-stabilized stainless steel

    SciTech Connect

    Klueh, R.L.

    1996-10-01

    Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

  20. Ion beam nitriding of single and polycrystalline austenitic stainless steel

    SciTech Connect

    Abrasonis, G.; Riviere, J.P.; Templier, C.; Declemy, A.; Pranevicius, L.; Milhet, X.

    2005-04-15

    Polycrystalline and single crystalline [orientations (001) and (011)] AISI 316L austenitic stainless steel was implanted at 400 deg. C with 1.2 keV nitrogen ions using a high current density of 0.5 mA cm{sup -2}. The nitrogen distribution profiles were determined using nuclear reaction analysis (NRA). The structure of nitrided polycrystalline stainless steel samples was analyzed using glancing incidence and symmetric x-ray diffraction (XRD) while the structure of the nitrided single crystalline stainless steel samples was analyzed using x-ray diffraction mapping of the reciprocal space. For identical treatment conditions, it is observed that the nitrogen penetration depth is larger for the polycrystalline samples than for the single crystalline ones. The nitrogen penetration depth depends on the orientation, the <001> being more preferential for nitrogen diffusion than <011>. In both type of samples, XRD analysis shows the presence of the phase usually called 'expanded' austenite or {gamma}{sub N} phase. The lattice expansion depends on the crystallographic plane family, the (001) planes showing an anomalously large expansion. The reciprocal lattice maps of the nitrided single crystalline stainless steel demonstrate that during nitriding lattice rotation takes place simultaneously with lattice expansion. The analysis of the results based on the presence of stacking faults, residual compressive stress induced by the lattice expansion, and nitrogen concentration gradient indicates that the average lattice parameter increases with the nitrided layer depth. A possible explanation of the anomalous expansion of the (001) planes is presented, which is based on the combination of faster nitriding rate in the (001) oriented grains and the role of stacking faults and compressive stress.

  1. Austenite Formation in a Cold-Rolled Semi-austenitic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Celada Casero, Carola; San Martín, David

    2014-04-01

    The progress of the martensite ( α') to austenite ( γ) phase transformation has been thoroughly investigated at different temperatures during the continuous heating of a cold-rolled precipitation hardening metastable stainless steel at a heating rate of 0.1 K/s. Heat-treated samples have been characterized using different experimental complementary techniques: high-resolution dilatometry, magnetization, and thermoelectric power (TEP) measurements, micro-hardness-Vickers testing, optical/scanning electron microscopy, and tensile testing. The two-step transformation behavior observed is thought to be related to the presence of a pronounced chemical banding in the initial microstructure. This banding has been characterized using electron probe microanalysis. Unexpectedly, dilatometry measurements seem unable to locate the end of the transformation accurately, as this technique does not detect the second step of this transformation (last 20 pct of it). It is shown that once the starting ( A S) and finishing ( A F) transformation temperatures have been estimated by magnetization measurements, the evolution of the volume fractions of austenite and martensite can be evaluated by TEP or micro-hardness measurement quite reliably as compared to magnetization measurements. The mechanical response of the material after being heated to temperatures close to A S, A F, and ( A F - A S)/2 is also discussed.

  2. Hydrogen-related phase transformations in austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Narita, N.; Altstetter, C. J.; Birnbaum, H. K.

    1982-08-01

    The effect of hydrogen and stress (strain) on the stability of the austenite phase in stainless steels was investigated. Hydrogen was introduced by severe cathodic charging and by elevated temperature equilibration with high pressure H2 gas. Using X-ray diffraction and magnetic techniques, the behavior of two “stable” type AISI310 steels and an “unstable” type AISI304 steel was studied during charging and during the outgassing period following charging. Transformation from the fcc γ phase to an expanded fcc phase, γ*, and to the hcp ɛ phase occurred during cathodic charging. Reversion of the γ* and e phases to the original γ structure and formation of the bcc α structure were examined, and the kinetics of these processes was studied. The γ* phase was shown to be ferromagnetic with a subambient Curie temperature. The γ⇆ɛ phase transition was studied after hydrogen charging in high pressure gas, as was the formation of a during outgassing. These results are interpreted as effects of hydrogen and stress (strain) on the stability of the various phases. A proposed psuedo-binary phase diagram for the metal-hydrogen system was proposed to account for the formation of the γ* phase. The relation of these phase changes to hydrogen embrittlement and stress corrosion cracking of stainless steel is discussed.

  3. Austenite Grain Growth and the Surface Quality of Continuously Cast Steel

    NASA Astrophysics Data System (ADS)

    Dippenaar, Rian; Bernhard, Christian; Schider, Siegfried; Wieser, Gerhard

    2014-04-01

    Austenite grain growth does not only play an important role in determining the mechanical properties of steel, but certain surface defects encountered in the continuous casting industry have also been attributed to the formation of large austenite grains. Earlier research has seen innovative experimentation, the development of metallographic techniques to determine austenite grain size and the building of mathematical models to simulate the conditions pertaining to austenite grain growth during the continuous casting of steel. Oscillation marks and depressions in the meniscus region of the continuously casting mold lead to retarded cooling of the strand surface, which in turn results in the formation of coarse austenite grains, but little is known about the mechanism and rate of formation of these large austenite grains. Relevant earlier research will be briefly reviewed to put into context our recent in situ observations of the delta-ferrite to austenite phase transition. We have confirmed earlier evidence that very large delta-ferrite grains are formed very quickly in the single-phase region and that these large delta-ferrite grains are transformed to large austenite grains at low cooling rates. At the higher cooling rates relevant to the early stages of the solidification of steel in a continuously cast mold, delta-ferrite transforms to austenite by an apparently massive type of transformation mechanism. Large austenite grains then form very quickly from this massive type of microstructure and on further cooling, austenite transforms to thin ferrite allotriomorphs on austenite grain boundaries, followed by Widmanstätten plate growth, with almost no regard to the cooling rate. This observation is important because it is now well established that the presence of a thin ferrite film on austenite grain boundaries is the main cause of reduction in hot ductility. Moreover, this reduction in ductility is exacerbated by the presence of large austenite grains.

  4. Retained austenite characteristics in thermomechanically processed Si-Mn transformation-induced plasticity steels

    SciTech Connect

    Hanzaki, A.Z.; Hodgson, P.D.; Yue, S.

    1997-11-01

    It is well known that a significant amount of retained austenite can be obtained in steels containing high additions (>1 pct) of Si, where bainite is the predominant microconstituent. Furthermore, retained austenite with optimum characteristics (volume fraction, composition, morphology, size, and distribution), when present in ferrite plus bainite microstructures, can potentially increase strength and ductility, such that formability and final properties are greatly improved. These beneficial properties can be obtained largely by transformation-induced plasticity (TRIP). In this work, the effect of a microalloy addition (0.035 pct Nb) in a 0.22 pct C-1.55 pct Si-1.55 pct Mn TRIP steel was investigated. Niobium was added to enable the steel to be processed by a variety of thermomechanical processing (TMP) routes, thus allowing the effects of prior austenite grain size, austenite recrystallization temperature, Nb in austenite solid solution, and Nb as a precipitate to be studied. The results, which were compared with those of the same steel without Nb, indicate that the retained austenite volume fraction is strongly influenced by both prior austenite grain size and the state of Nb in austenite. Promoting Nb(CN) precipitation by the change in TMP conditions resulted in a decrease in the V{sub RA}. These findings are rationalized by considering the effects of changes in the TMP conditions on the subsequent transformation characteristics of the parent austenite.

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

    NASA Astrophysics Data System (ADS)

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

    2015-06-01

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

  6. Assessment of void swelling in austenitic stainless steel PWR core internals.

    SciTech Connect

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

  7. The stability of precipitated austenite and the toughness of 9Ni steel

    NASA Astrophysics Data System (ADS)

    Fultz, B.; Kim, J. I.; Kim, Y. H.; Kim, H. J.; Fior, G. O.; Morris, J. W.

    1985-12-01

    A correlation was confirmed between the good low temperature Charpy toughness of 9Ni steel and the stability of its precipitated austenite against the martensitic transformation. Changes in the microstructure during isothermal tempering were studied in detail. The austenite/martensite interface is originally quite coherent over ˜100 A distances. With further tempering, however, the dislocation structure at the austenite/martensite interface changes, and this change may be related to the increased instability of the austenite particles. The reduction in austenite carbon concentration does not seem large enough to account for the large reduction in austenite stability with tempering time. The strains inherent to the transformation of austenite particles create dislocation structures in the tempered martensite. The large deterioration of the Charpy toughness of overtempered material is attributed, in part, to these dislocation structures.

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

    SciTech Connect

    Woollin, P.

    1996-12-31

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

  9. Mechanical Properties of Austenitic Stainless Steel Made by Additive Manufacturing

    PubMed Central

    Luecke, William E; Slotwinski, John A

    2014-01-01

    Using uniaxial tensile and hardness testing, we evaluated the variability and anisotropy of the mechanical properties of an austenitic stainless steel, UNS S17400, manufactured by an additive process, selective laser melting. Like wrought materials, the mechanical properties depend on the orientation introduced by the processing. The recommended stress-relief heat treatment increases the tensile strength, reduces the yield strength, and decreases the extent of the discontinuous yielding. The mechanical properties, assessed by hardness, are very uniform across the build plate, but the stress-relief heat treatment introduced a small non-uniformity that had no correlation to position on the build plate. Analysis of the mechanical property behavior resulted in four conclusions. (1) The within-build and build-to-build tensile properties of the UNS S17400 stainless steel are less repeatable than mature engineering structural alloys, but similar to other structural alloys made by additive manufacturing. (2) The anisotropy of the mechanical properties of the UNS S17400 material of this study is larger than that of mature structural alloys, but is similar to other structural alloys made by additive manufacturing. (3) The tensile mechanical properties of the UNS S17400 material fabricated by selective laser melting are very different from those of wrought, heat-treated 17-4PH stainless steel. (4) The large discontinuous yielding strain in all tests resulted from the formation and propagation of Lüders bands. PMID:26601037

  10. Modified Monkman-Grant relationship for austenitic stainless steel foils

    NASA Astrophysics Data System (ADS)

    Osman Ali, Hassan; Tamin, Mohd Nasir

    2013-02-01

    Characteristics of creep deformation for austenitic stainless steel foils are examined using the modified Monkman-Grant equation. A series of creep tests are conducted on AISI 347 steel foils at 700 °C and different stress levels ranging from 54 to 221 MPa. Results showed that at lower stress levels below 110 MPa, the creep life parameters ɛ, ɛr, tr can be expressed using the modified Monkman-Grant equation with exponent m'= 0.513. This indicates significant deviation of the creep behavior from the first order reaction kinetics theory for creep (m' = 1.0). The true tertiary creep damage in AISI 347 steel foil begins after 65.9% of the creep life of the foil has elapsed at stress levels above 150 MPa. At this high stress levels, Monkman-Grant ductility factor λ' saturates to a value of 1.3 with dislocation-controlled deformation mechanisms operating. At low stress levels, λ' increases drastically (λ'=190 at 54 MPa) when slow diffusion-controlled creep is dominant.

  11. A new constitutive model for nitrogen austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Fréchard, S.; Lichtenberger, A.; Rondot, F.; Faderl, N.; Redjaïmia, A.; Adoum, M.

    2003-09-01

    Quasi-static, quasi-dynamic and dynamic compression tests have been performed on a nitrogen alloyed austenitic stainless steel. For all strain rates, a high strain hardening rate and a good ductility have been achieved. In addition, this steel owns a great strain rate sensitivity. The temperature sensitivity bas been determined between 20°C and 400°C. Microstructural analysis has been performed after different loading conditions in relation to the behaviour of the material. Johnson-Cook and Zerilli-Armstrong models have been selected to fit the experimental data into constitutive equations. These models do not reproduce properly the behaviour of this type of steel over the complete range. A new constitutive model that fits very well all the experimental data at different strain, strain rate and temperature has been determined. The model is based on empirical considerations on the separated influence of the main parameters. Single Taylor tests have been realized to validate the models. Live observations of the specimen during impact have been achieved using a special CCD camera set-up. The overall profile at different times are compared to numerical predictions using LS-DYNA code.

  12. Mechanical Properties of Austenitic Stainless Steel Made by Additive Manufacturing.

    PubMed

    Luecke, William E; Slotwinski, John A

    2014-01-01

    Using uniaxial tensile and hardness testing, we evaluated the variability and anisotropy of the mechanical properties of an austenitic stainless steel, UNS S17400, manufactured by an additive process, selective laser melting. Like wrought materials, the mechanical properties depend on the orientation introduced by the processing. The recommended stress-relief heat treatment increases the tensile strength, reduces the yield strength, and decreases the extent of the discontinuous yielding. The mechanical properties, assessed by hardness, are very uniform across the build plate, but the stress-relief heat treatment introduced a small non-uniformity that had no correlation to position on the build plate. Analysis of the mechanical property behavior resulted in four conclusions. (1) The within-build and build-to-build tensile properties of the UNS S17400 stainless steel are less repeatable than mature engineering structural alloys, but similar to other structural alloys made by additive manufacturing. (2) The anisotropy of the mechanical properties of the UNS S17400 material of this study is larger than that of mature structural alloys, but is similar to other structural alloys made by additive manufacturing. (3) The tensile mechanical properties of the UNS S17400 material fabricated by selective laser melting are very different from those of wrought, heat-treated 17-4PH stainless steel. (4) The large discontinuous yielding strain in all tests resulted from the formation and propagation of Lüders bands. PMID:26601037

  13. Welding techniques for high alloy austenitic stainless steel

    SciTech Connect

    Gooch, T.G.; Woollin, P.

    1996-11-01

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

  14. Formability analysis of austenitic stainless steel-304 under warm conditions

    NASA Astrophysics Data System (ADS)

    Lade, Jayahari; Singh, Swadesh Kumar; Banoth, Balu Naik; Gupta, Amit Kumar

    2013-12-01

    A warm deep drawing process of austenitic stainless steel-304 (ASS-304) of circular blanks with coupled ther mal analysis is studied in this article. 65 mm blanks were deep drawn at different temperatures and thickness distribution is experimentally measured after cutting the drawn component into two halves. The process is simulated using explicit fin ite element code LS-DYNA. A Barlat 3 parameter model is used in the simulation, as the material is anisotropic up to 30 0°C. Material properties for the simulation are determined at different temperatures using a 5 T UTM coupled with a furn ace. In this analysis constant punch speed and variable blank holder force (BHF) is applied to draw cups without wrinkle.

  15. Strain oxidation cracking of austenitic stainless steels at 610 C

    SciTech Connect

    Calvar, M. Le; Scott, P.M.; Magnin, T.; Rieux, P.

    1998-02-01

    Strain oxidation cracking of both forged and welded austenitic stainless steels (SS) was studied. Creep and slow strain rate tests (SSRT) were performed in vacuum, air, and a gas furnace environment (air + carbon dioxide [CO{sub 2}] + water [H{sub 2}O]). Results showed cracking was environmentally dependent. Almost no cracking was observed in vacuum, whereas intergranular cracking was observed with increasing severity in passing from an air to a gas furnace environment. The most severe cracking was associated with formation of a less protective film formed in the gas furnace environment (air: haematite-like M{sub 2}O{sub 3} oxide; gas furnace environment: spinel M{sub 3}O{sub 4} oxide). Cracking depended strongly on the carbon content and the sensitization susceptibility of the material: the higher the carbon content, the more susceptible the alloy. This cracking was believed to be similar to other oxidation-induced cracking phenomena.

  16. Fatigue crack growth in metastable austenitic stainless steels

    SciTech Connect

    Mei, Z.; Chang, G.; Morris, J.W. Jr.

    1988-06-01

    The research reported here is an investigation of the influence of the mechanically induced martensitic transformation on the fatigue crack growth rate in 304-type steels. The alloys 304L and 304LN were used to test the influence of composition, the testing temperatures 298 K and 77 K were used to study the influence of test temperature, and various load ratios (R) were used to determine the influence of the load ratio. It was found that decreasing the mechanical stability of the austenite by changing composition or lowering temperature decreases the fatigue crack growth rate. The R-ratio effect is more subtle. The fatigue crack growth rate increases with increasing R-ratio, even though this change increases the martensite transformation. Transformation-induced crack closure can explain the results in the threshold regime, but cannot explain the R-ratio effect at higher cyclic stress intensities. 26 refs., 6 figs.

  17. Austenite Formation Kinetics During Rapid Heating in a Microalloyed Steel

    SciTech Connect

    BURNETT,M.E.; DYKHUIZEN,RONALD C.; KELLEY,J. BRUCE; PUSKAR,JOSEPH D.; ROBINO,CHARLES V.

    1999-09-07

    The model parameters for the normalized 1054V1 material were compared to parameters previously generated for 1026 steel, and the transformation behavior was relatively consistent. Validation of the model predictions by heating into the austenite plus undissolved ferrite phase field and rapidly quenching resulted in reasonable predictions when compared to the measured volume fractions from optical metallography. The hot rolled 1054V1 material, which had a much coarser grain size and a non-equilibrium volume fraction of pearlite, had significantly different model parameters and the on heating transformation behavior of this material was less predictable with the established model. The differences in behavior is consistent with conventional wisdom that normalized micro-structure produce a more consistent response to processing, and it reinforces the need for additional work in this area.

  18. Structure and properties of high-temperature austenitic steels for superheater tubes

    NASA Astrophysics Data System (ADS)

    Blinov, V. M.

    2009-12-01

    The structure and properties of high-temperature austenitic steels intended for superheater tubes are analyzed. Widely used Kh18N10T (AISI 304) and Kh16N13M3 (AISI 316) steels are found not to ensure a stable austenitic structure and stable properties during long-term thermal holding under stresses. The hardening of austenitic steels by fine particles of vanadium and niobium carbides and nitrides and γ'-phase and Fe2W and Fe2Mo Laves phase intermetallics is considered. The role of Cr23C6 chromium carbides, the σ phase, and coarse precipitates of an M 3B2 phase and a boron-containing eutectic in decreasing the time to failure and the stress-rupture strength of austenitic steels is established. The mechanism of increasing the stress-rupture strength of steels by boron additions is described. The chemical compositions, mechanical properties, stress-rupture strength, and creep characteristics of Russian and foreign austenitic steels used or designed for superheater tubes intended for operation under stress conditions at temperatures above 600°C are presented. The conditions are found for increasing the strength, plasticity, and thermodeformation stability of austenite in steels intended for superheater tubes operating at 700°C under high stresses for a long time.

  19. Magnetic properties of single crystalline expanded austenite obtained by plasma nitriding of austenitic stainless steel single crystals.

    PubMed

    Menéndez, Enric; Templier, Claude; Garcia-Ramirez, Pablo; Santiso, José; Vantomme, André; Temst, Kristiaan; Nogués, Josep

    2013-10-23

    Ferromagnetic single crystalline [100], [110], and [111]-oriented expanded austenite is obtained by plasma nitriding of paramagnetic 316L austenitic stainless steel single crystals at either 300 or 400 °C. After nitriding at 400 °C, the [100] direction appears to constitute the magnetic easy axis due to the interplay between a large lattice expansion and the expected decomposition of the expanded austenite, which results in Fe- and Ni-enriched areas. However, a complex combination of uniaxial (i.e., twofold) and biaxial (i.e., fourfold) in-plane magnetic anisotropies is encountered. It is suggested that the former is related to residual stress-induced effects while the latter is associated to the in-plane projections of the cubic lattice symmetry. Increasing the processing temperature strengthens the biaxial in-plane anisotropy in detriment of the uniaxial contribution, in agreement with a more homogeneous structure of expanded austenite with lower residual stresses. In contrast to polycrystalline expanded austenite, single crystalline expanded austenite exhibits its magnetic easy axes along basic directions. PMID:24028676

  20. Austenite Formation from Martensite in a 13Cr6Ni2Mo Supermartensitic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Bojack, A.; Zhao, L.; Morris, P. F.; Sietsma, J.

    2016-05-01

    The influence of austenitization treatment of a 13Cr6Ni2Mo supermartensitic stainless steel (X2CrNiMoV13-5-2) on austenite formation during reheating and on the fraction of austenite retained after tempering treatment is measured and analyzed. The results show the formation of austenite in two stages. This is probably due to inhomogeneous distribution of the austenite-stabilizing elements Ni and Mn, resulting from their slow diffusion from martensite into austenite and carbide and nitride dissolution during the second, higher temperature, stage. A better homogenization of the material causes an increase in the transformation temperatures for the martensite-to-austenite transformation and a lower retained austenite fraction with less variability after tempering. Furthermore, the martensite-to-austenite transformation was found to be incomplete at the target temperature of 1223 K (950 °C), which is influenced by the previous austenitization treatment and the heating rate. The activation energy for martensite-to-austenite transformation was determined by a modified Kissinger equation to be approximately 400 and 500 kJ/mol for the first and the second stages of transformation, respectively. Both values are much higher than the activation energy found during isothermal treatment in a previous study and are believed to be effective activation energies comprising the activation energies of both mechanisms involved, i.e., nucleation and growth.

  1. Effect of initial microstructure on austenite formation kinetics in high-strength experimental microalloyed steels

    NASA Astrophysics Data System (ADS)

    López-Martínez, Edgar; Vázquez-Gómez, Octavio; Vergara-Hernández, Héctor Javier; Campillo, Bernardo

    2015-12-01

    Austenite formation kinetics in two high-strength experimental microalloyed steels with different initial microstructures comprising bainite-martensite and ferrite-martensite/austenite microconstituents was studied during continuous heating by dilatometric analysis. Austenite formation occurred in two steps: (1) carbide dissolution and precipitation and (2) transformation of residual ferrite to austenite. Dilatometric analysis was used to determine the critical temperatures of austenite formation and continuous heating transformation diagrams for heating rates ranging from 0.03°C•s-1 to 0.67°C•s-1. The austenite volume fraction was fitted using the Johnson-Mehl-Avrami-Kolmogorov equation to determine the kinetic parameters k and n as functions of the heating rate. Both n and k parameters increased with increasing heating rate, which suggests an increase in the nucleation and growth rates of austenite. The activation energy of austenite formation was determined by the Kissinger method. Two activation energies were associated with each of the two austenite formation steps. In the first step, the austenite growth rate was controlled by carbon diffusion from carbide dissolution and precipitation; in the second step, it was controlled by the dissolution of residual ferrite to austenite.

  2. A review on nickel-free nitrogen containing austenitic stainless steels for biomedical applications.

    PubMed

    Talha, Mohd; Behera, C K; Sinha, O P

    2013-10-01

    The field of biomaterials has become a vital area, as these materials can enhance the quality and longevity of human life. Metallic materials are often used as biomaterials to replace structural components of the human body. Stainless steels, cobalt-chromium alloys, commercially pure titanium and its alloys are typical metallic biomaterials that are being used for implant devices. Stainless steels have been widely used as biomaterials because of their very low cost as compared to other metallic materials, good mechanical and corrosion resistant properties and adequate biocompatibility. However, the adverse effects of nickel ions being released into the human body have promoted the development of "nickel-free nitrogen containing austenitic stainless steels" for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also much improves steel properties. Here we review the harmful effects associated with nickel and emphatically the advantages of nitrogen in stainless steel, as well as the development of nickel-free nitrogen containing stainless steels for medical applications. By combining the benefits of stable austenitic structure, high strength, better corrosion and wear resistance and superior biocompatibility in comparison to the currently used austenitic stainless steel (e.g. 316L), the newly developed nickel-free high nitrogen austenitic stainless steel is a reliable substitute for the conventionally used medical stainless steels. PMID:23910251

  3. HYDROGEN-ASSISTED FRACTURE IN FORGED TYPE 304L AUSTENITIC STAINLESS STEEL

    SciTech Connect

    Switzner, Nathan; Neidt, Ted; Hollenbeck, John; Knutson, J.; Everhart, Wes; Hanlin, R.; Bergen, R.; Balch, D. K.

    2012-09-06

    Austenitic stainless steels generally have good resistance to hydrogen-assisted fracture; however, structural designs for high-pressure gaseous hydrogen are constrained by the low strength of this class of material. Forging is used to increase the low strength of austenitic stainless steels, thus improving the efficiency of structural designs. Hydrogen-assisted racture, however, depends on microstructural details associated with manufacturing. In this study, hydrogen-assisted fracture of forged type 304L austenitic stainless steel is investigated. Microstructural variation in multi-step forged 304L was achieved by forging at different rates and temperatures, and by process annealing. High internal hydrogen content in forged type 304L austenitic stainless steel is achieved by thermal precharging in gaseous hydrogen and results in as much as 50% reduction of tensile ductility.

  4. Microstructural Evolution During Friction Surfacing of Austenitic Stainless Steel AISI 304 on Low Carbon Steel

    NASA Astrophysics Data System (ADS)

    Khalid Rafi, H.; Kishore Babu, N.; Phanikumar, G.; Prasad Rao, K.

    2013-01-01

    Austenitic stainless steel AISI 304 coating was deposited over low carbon steel substrate by means of friction surfacing and the microstructural evolution was studied. The microstructural characterization of the coating was carried out by optical microscopy (OM), electron back scattered diffraction (EBSD), and transmission electron microscopy (TEM). The coating exhibited refined grains (average size of 5 μm) as compared to the coarse grains (average size of 40 μm) in as-received consumable rod. The results from the microstructural characterization studies show that discontinuous dynamic recrystallization (DDRX) is the responsible mechanism for grain evolution as a consequence of severe plastic deformation.

  5. Hot compression deformation behavior of AISI 321 austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Haj, Mehdi; Mansouri, Hojjatollah; Vafaei, Reza; Ebrahimi, Golam Reza; Kanani, Ali

    2013-06-01

    The hot compression behavior of AISI 321 austenitic stainless steel was studied at the temperatures of 950-1100°C and the strain rates of 0.01-1 s-1 using a Baehr DIL-805 deformation dilatometer. The hot deformation equations and the relationship between hot deformation parameters were obtained. It is found that strain rate and deformation temperature significantly influence the flow stress behavior of the steel. The work hardening rate and the peak value of flow stress increase with the decrease of deformation temperature and the increase of strain rate. In addition, the activation energy of deformation ( Q) is calculated as 433.343 kJ/mol. The microstructural evolution during deformation indicates that, at the temperature of 950°C and the strain rate of 0.01 s-1, small circle-like precipitates form along grain boundaries; but at the temperatures above 950°C, the dissolution of such precipitates occurs. Energy-dispersive X-ray analyses indicate that the precipitates are complex carbides of Cr, Fe, Mn, Ni, and Ti.

  6. Deuterium retention in ITER-grade austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Nemanič, Vincenc; Žumer, Marko; Zajec, Bojan

    2008-11-01

    In view of the construction of ITER, it is essential to confirm that the retention of tritium by the large interior surface area of stainless steel will not become an issue for safety or operating inventory reasons. Retention of deuterium in ITER-grade austenitic stainless steel samples was studied during t = 24 h exposures to pure gaseous deuterium at p = 0.01 mbar and 0.1 mbar and T = 100 °C, 250 °C and 400 °C, respectively. The required high sensitivity for distinguishing hydrogen isotopes involved in the process (H2, HD and D2) was gained after suppression of the native hydrogen concentration by a thermal treatment at T = 400 °C for t = 200 h. The quantity of retained deuterium was determined by measuring the absolute pressure change during the deuterium exposure and subsequent mass spectrometry revealing an intense isotope exchange reaction. The retained amount of 2.6 × 1016 D cm-2 was the highest at T = 400 °C and p = 0.1 mbar and noticeably less at lower deuterium pressure and temperature. Our results, when compared with similar tritium exposures, do not exceed the limits set in the generic safety analysis for the ITER. They manifest that an extremely high sensitivity for deuterium absorption and release can be gained with a precise pressure measuring technique, otherwise attributed exclusively to tritium scintillation methods.

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

    NASA Astrophysics Data System (ADS)

    Gan, Jian

    1999-11-01

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

  8. The formation of twinned austenite in Fe-10Cr-10Ni-2W maraging steel

    SciTech Connect

    Suk, J.I.; Hong, S.H.; Nam, S.W. )

    1991-12-01

    The precipitation hardening mechanisms in high strength maraging steels have been studied in detail by many investigators, but limited information is available on the formation of austenite during aging. Some investigations have been concerned with the understanding of the effect of reverted austenite formed during aging on the mechanical properties. However, only a few investigations have been reported on the morphology and crystallographic feature of austenite. Shiang and Wayman first reported the twin-related and coupled morphology of Widmanstatten austenite plates which were frequently observed in maraging steel. In addition, Ameyama et al. reported the morphology and crystallographic features of austenite formed in ferrite grain during aging in a two-phase stainless steel, and found that each side of the austenite pair of twins satisfies the Kurdjumov-Sachs (K-S) orientation relationship with the parent phase. The morphology and crystallographic features of the reverted austenite formed during aging of Fe-10Cr-10Ni-2W stainless maraging steel have been investigated in this paper. The major strengthening precipitate in Fe-10Cr-10Ni-2W maraging steels has been identified as the rod-shaped {eta}-Ni{sub 3}Ti phase in our previous study. The peculiar morphology of the austenite, i.e., twinned austenite, also has been found in our studies of maraging steel in the Fe-10Cr-10Ni-2W lath martensite. In addition, computer simulation of the diffraction pattern is used to confirm the orientation relationships, such as the Kurdjumov-Sachs (K-S) relationship, the Nishiyama-Wasserman (N-W) relationship and the twin relationship by comparisons with the experimentaly observed results.

  9. Study of biocompatibility of medical grade high nitrogen nickel-free austenitic stainless steel in vitro.

    PubMed

    Li, Menghua; Yin, Tieying; Wang, Yazhou; Du, Feifei; Zou, Xingzheng; Gregersen, Hans; Wang, Guixue

    2014-10-01

    Adverse effects of nickel ions being released into the living organism have resulted in development of high nitrogen nickel-free austenitic stainless steels for medical applications. Nitrogen not only replaces nickel for austenitic structure stability but also improves steel properties. The cell cytocompatibility, blood compatibility and cell response of high nitrogen nickel-free austenitic stainless steel were studied in vitro. The mechanical properties and microstructure of this stainless steel were compared to the currently used 316L stainless steel. It was shown that the new steel material had comparable basic mechanical properties to 316L stainless steel and preserved the single austenite organization. The cell toxicity test showed no significant toxic side effects for MC3T3-E1 cells compared to nitinol alloy. Cell adhesion testing showed that the number of MC3T3-E1 cells was more than that on nitinol alloy and the cells grew in good condition. The hemolysis rate was lower than the national standard of 5% without influence on platelets. The total intracellular protein content and ALP activity and quantification of mineralization showed good cell response. We conclude that the high nitrogen nickel-free austenitic stainless steel is a promising new biomedical material for coronary stent development. PMID:25175259

  10. Austenite layer and precipitation in high Co-Ni maraging steel.

    PubMed

    Wang, Chenchong; Zhang, Chi; Yang, Zhigang

    2014-12-01

    In high Co-Ni maraging steel, austenite has a great effect on the fracture toughness of the steel and the precipitated carbides are the main strengthening phase. In this study, both austenite layers and precipitation were observed and their formation theory was analyzed by Thermo-Calc simulation and several reported results. TEM and HRTEM observation results showed that the thickness of the austenite layers was about 5-10 nm and the length of the needle-like precipitated carbides was less than 10nm. The carbides maintained coherent or semi-coherent relation with the matrix. PMID:25129424

  11. Structure and Mechanical Properties of Nitrogen Austenitic Steel after Ultrasonic Forging

    NASA Astrophysics Data System (ADS)

    Narkevich, N. A.; Tolmachev, A. I.; Vlasov, I. V.; Surikova, N. S.

    2016-03-01

    Electron microscopy and X-ray diffraction have been used to investigate a nitrogen 07Kh17AG18 steel with an austenitic structure after the surface deformation treatment—ultrasonic forging. During ultrasonic forging, an austenitic structure transforms into a new structure with an elevated concentration of deformation-induced stacking faults, a lot of deformation microtwins, ɛ-martensite crystals. The austenite lattice parameter is found to be decreased in the surface layer. After ultrasonic forging, nitrided steel exhibits enhanced strength properties with retained high plasticity.

  12. Austenite precipitation during tempering in 16Cr-2Ni martensitic stainless steels

    SciTech Connect

    Balan, K.P.; Reddy, A.V.; Sarma, D.S.

    1998-09-04

    The 16Cr-2Ni steel when quenched from austenitizing temperature of 1,323K results in the formation of a complex microstructure consisting of the inherited {delta}-ferrite, martensite and retained austenite with a few undissolved M{sub 23}C{sub 6} carbides. There do not appear to be many reports on tempering behavior of 16Cr-2Ni steel through microstructural characterization using transmission electron microscopy. A comprehensive study is under progress to examine the structure-fracture-property relationship on 16Cr-2Ni steel and the microstructural changes that occur on tempering the steel are dealt with in this paper.

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

    NASA Astrophysics Data System (ADS)

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

    2013-10-01

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

  14. Defect microstructures in neutron-irradiated copper and stainless steel

    SciTech Connect

    Zinkle, S.J.; Sindelar, R.L.

    1987-09-01

    The defect microstructures of copper and type 304L austenitic stainless steel have been examined following neutron irradiation under widely different conditions. Less than 0.2% of the defect clusters in steel irradiated at 120/sup 0/C with moderated fission neutrons were resolvable as stacking fault tetrahedra (SFT). The fraction of defect clusters identified as SFT in copper varied from approx.10% for a low-dose 14-MeV neutron irradiation at 25/sup 0/C to approx.50% for copper irradiated to 1.3 dpa in a moderated fission spectrum at 182/sup 0/C. The mean cluster size in copper was about 2.6 nm for both cases, despite the large differences in irradiation conditions. The mean defect cluster size in the irradiated steel was about 1.8 nm. The absence of SFT in stainless steel may be due to the generation of 35 appm He during the irradiation, which caused the vacancies to form helium-filled cavities instead of SFT. 20 refs.

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

    SciTech Connect

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

    1997-04-01

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

  16. Implications of radiation-induced reductions in ductility to the design of austenitic stainless steel structures

    SciTech Connect

    Lucas, G.E.; Billone, M.; Pawel, J.E.; Hamilton, M.L.

    1995-12-31

    In the dose and temperature range anticipated for ITER, austenitic stainless steels exhibit significant hardening with a concomitant loss in work hardening and uniform elongation. However, significant post-necking ductility may still be retained. When uniform elongation (e{sub u}) is well defined in terms of a plastic instability criterion, e{sub u} is found to sustain reasonably high values out to about 7 dpa in the temperature range 250-350 C, beyond which it decreases to about 0.3% for 316LN. This loss of ductility has significant implications to fracture toughness and the onset of new failure modes associated with hear instability. However, the retention of a significant reduction in area at failure following irradiation indicates a less severe degradation of low-cycle fatigue life in agreement with a limited amount of data obtained to date. Suggestions are made for incorporating these results into design criteria and future testing programs.

  17. Long term corrosion resistance of alumina forming austenitic stainless steels in liquid lead

    NASA Astrophysics Data System (ADS)

    Ejenstam, Jesper; Szakálos, Peter

    2015-06-01

    Alumina forming austenitic steels (AFA) and commercial stainless steels have been exposed in liquid lead with 10-7 wt.% oxygen at 550 °C for up to one year. It is known that chromia forming austenitic stainless steels, such as 316L and 15-15 Ti, have difficulties forming protective oxides in liquid lead at temperatures above 500 °C, which is confirmed in this study. By adding Al to austenitic steels, it is in general terms possible to increase the corrosion resistance. However this study shows that the high Ni containing AFA alloys are attacked by the liquid lead, i.e. dissolution attack occurs. By lowering the Ni content in AFA alloys, it is possible to achieve excellent oxidation properties in liquid lead. Following further optimization of the microstructural properties, low Ni AFA alloys may represent a promising future structural steel for lead cooled reactors.

  18. Fundamental study of the austenite formation and decomposition in low-silicon, aluminum added TRIP steels

    NASA Astrophysics Data System (ADS)

    Garcia-Gonzalez, Jose Enrique

    2005-11-01

    TRIP (Transformation Induced Plasticity) steels are under development for automotive applications that require high strength and excellent formability. Conventional TRIP steels consist of a multiphase microstructure comprised of a ferrite matrix with a dispersion of bainite and metastable retained austenite. The high ductility exhibited by these steels results from the transformation of the metastable retained austenite to martensite during straining. In conventional TRIP steel processing, the multiphase microstructure is obtained by controlled cooling from the alpha + gamma region to an isothermal holding temperature. During this holding, bainite forms and carbon is rejected out into the austenite, which lowers the Ms temperature and stabilizes the austenite to room temperature. In this research project, a fundamental study of a low-Si, Mo-Nb added cold rolled TRIP steel with and without Al additions was conducted. In this study, the recrystallization of cold-rolled ferrite, the formation of austenite during intercritical annealing and the characteristics of the decomposition of the intercritically annealed austenite by controlled cooling rates were systematically assessed. Of special interest were: (i) the effect of the initial hot band microstructure, (ii) the formation of epitaxial ferrite during cooling from the intercritical annealing temperature to the isothermal holding temperature, (iii) the influence of the intercritically annealed austenite on the formation of bainite during the isothermal holding temperature, and (iv) the influence of the processing variables on the type, amount, composition and stability of the retained austenite. During this research study, techniques such as OM, SEM, EBSD, TEM, XRD and Magnetometry were used to fully characterize the microstructures. Furthermore, a Gleeble 3500 unit at US Steel Laboratories was used for dilatometry studies and to simulate different CGL processing routes, from which specimens were obtained to evaluate

  19. Numerical simulation and experimental investigation of laser dissimilar welding of carbon steel and austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Nekouie Esfahani, M. R.; Coupland, J.; Marimuthu, S.

    2015-07-01

    This study reports an experimental and numerical investigation on controlling the microstructure and brittle phase formation during laser dissimilar welding of carbon steel to austenitic stainless steel. The significance of alloying composition and cooling rate were experimentally investigated. The investigation revealed that above a certain specific point energy the material within the melt pool is well mixed and the laser beam position can be used to control the mechanical properties of the joint. The heat-affected zone within the high-carbon steel has significantly higher hardness than the weld area, which severely undermines the weld quality. A sequentially coupled thermo-metallurgical model was developed to investigate various heat-treatment methodology and subsequently control the microstructure of the HAZ. Strategies to control the composition leading to dramatic changes in hardness, microstructure and service performance of the dissimilar laser welded fusion zone are discussed.

  20. Amorphous stainless steel coatings prepared by reactive magnetron-sputtering from austenitic stainless steel targets

    NASA Astrophysics Data System (ADS)

    Cusenza, Salvatore; Schaaf, Peter

    2009-01-01

    Stainless steel films were reactively magnetron sputtered in argon/methane gas flow onto oxidized silicon wafers using austenitic stainless-steel targets. The deposited films of about 200 nm thickness were characterized by conversion electron Mössbauer spectroscopy, magneto-optical Kerr-effect, X-ray diffraction, scanning electron microscopy, Rutherford backscattering spectrometry, atomic force microscopy, corrosion resistance tests, and Raman spectroscopy. These complementary methods were used for a detailed examination of the carburization effects in the sputtered stainless-steel films. The formation of an amorphous and soft ferromagnetic phase in a wide range of the processing parameters was found. Further, the influence of the substrate temperature and of post vacuum-annealing were examined to achieve a comprehensive understanding of the carburization process and phase formation.

  1. Material Parameters for Creep Rupture of Austenitic Stainless Steel Foils

    NASA Astrophysics Data System (ADS)

    Osman, H.; Borhana, A.; Tamin, M. N.

    2014-08-01

    Creep rupture properties of austenitic stainless steel foil, 347SS, used in compact recuperators have been evaluated at 700 °C in the stress range of 54-221 MPa to establish the baseline behavior for its extended use. Creep curves of the foil show that the primary creep stage is brief and creep life is dominated by tertiary creep deformation with rupture lives in the range of 10-2000 h. Results are compared with properties of bulk specimens tested at 98 and 162 MPa. Thin foil 347SS specimens were found to have higher creep rates and higher rupture ductility than their bulk specimen counterparts. Power law relationship was obtained between the minimum creep rate and the applied stress with stress exponent value, n = 5.7. The value of the stress exponent is indicative of the rate-controlling deformation mechanism associated with dislocation creep. Nucleation of voids mainly occurred at second-phase particles (chromium-rich M23C6 carbides) that are present in the metal matrix by decohesion of the particle-matrix interface. The improvement in strength is attributed to the precipitation of fine niobium carbides in the matrix that act as obstacles to the movement of dislocations.

  2. Dissimilar Friction Stir Welding Between UNS S31603 Austenitic Stainless Steel and UNS S32750 Superduplex Stainless Steel

    NASA Astrophysics Data System (ADS)

    Theodoro, Maria Claudia; Pereira, Victor Ferrinho; Mei, Paulo Roberto; Ramirez, Antonio Jose

    2015-02-01

    In order to verify the viability of dissimilar UNS S31603 austenitic and UNS S32750 superduplex stainless steels joined by friction stir welding, 6-mm-thick plates were welded using a PCBN-WRe tool. The welded joints were performed in position control mode at rotational speeds of 100 to 300 rpm and a feed rate of 100 mm/min. The joints performed with 150 and 200 rpm showed good appearance and no defects. The metallographic analysis of both joints showed no internal defects and that the material flow pattern is visible only in the stirred zone (SZ) of the superduplex steel. On the SZ top, these patterns are made of regions of different phases (ferrite and austenite), and on the bottom and central part of the SZ, these patterns are formed by alternated regions of different grain sizes. The ferrite grains in the superduplex steel are larger than those in the austenitic ones along the SZ and thermo-mechanically affected zone, explained by the difference between austenite and ferrite recrystallization kinetics. The amount of ferrite islands present on the austenitic steel base metal decreased near the SZ interface, caused by the dissolving of the ferrite in austenitic matrix. No other phases were found in both joints. The best weld parameters were found to be 200 rpm rotation speed, 100 mm/min feed rate, and tool position control.

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

    DOEpatents

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

    1987-04-28

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

  4. Texture evolution of warm-rolled and annealed 304L and 316L austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Lindell, D.

    2015-04-01

    The brass-to-copper rolling texture transition is observed during warm rolling austenitic stainless steels. In the current paper austenitic stainless steels 304L and 316L have been subjected to warm rolling at 700°C to 90% reduction. The evolution of microstructure and texture during subsequent annealing has been studied using dilatometry and electron backscatter diffraction. Recrystallisation texture for 304L was primarily cube with some retained rolling texture while 316L only had retained rolling texture. The different behaviour between the two steels is believed to originate from differences in molybdenum content.

  5. Stages of austenitization of cold-worked low-carbon steel in intercritical temperature range

    NASA Astrophysics Data System (ADS)

    Panov, D. O.; Simonov, Y. N.; Spivak, L. V.; Smirnov, A. I.

    2015-08-01

    Austenization processes in 10Kh3G3MF low-carbon steel in the initially cold-worked state are investigated during its continuous heating in an intercritical temperature range. The austenization of this steel has three stages, which is shown by dilatometry, differential scanning calorimetry, and transmission electron microscopy. The thermokinetic diagram of the austenite formation in 10Kh3G3MF steel is constructed. Critical points A c1 and A c2 and temperature ranges of austenite formation at every stage of the α → γ transformation at heating rates of 0.6-400 K/s are determined.

  6. Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

    NASA Astrophysics Data System (ADS)

    Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

    2014-03-01

    Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H+ ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix-Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix-Gao plot. The bulk hardness of the irradiated region, H0, estimated by the Nix-Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H0. Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials.

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

    SciTech Connect

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

    1988-01-01

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

  8. Retained Austenite in SAE 52100 Steel Post Magnetic Processing and Heat Treatment

    SciTech Connect

    Pappas, Nathaniel R; Watkins, Thomas R; Cavin, Odis Burl; Jaramillo, Roger A; Ludtka, Gerard Michael

    2007-01-01

    Steel is an iron-carbon alloy that contains up to 2% carbon by weight. Understanding which phases of iron and carbon form as a function of temperature and percent carbon is important in order to process/manufacture steel with desired properties. Austenite is the face center cubic (fcc) phase of iron that exists between 912 and 1394 C. When hot steel is rapidly quenched in a medium (typically oil or water), austenite transforms into martensite. The goal of the study is to determine the effect of applying a magnetic field on the amount of retained austenite present at room temperature after quenching. Samples of SAE 52100 steel were heat treated then subjected to a magnetic field of varying strength and time, while samples of SAE 1045 steel were heat treated then subjected to a magnetic field of varying strength for a fixed time while being tempered. X-ray diffraction was used to collect quantitative data corresponding to the amount of each phase present post processing. The percentage of retained austenite was then calculated using the American Society of Testing and Materials standard for determining the amount of retained austenite for randomly oriented samples and was plotted as a function of magnetic field intensity, magnetic field apply time, and magnetic field wait time after quenching to determine what relationships exist with the amount of retained austenite present. In the SAE 52100 steel samples, stronger field strengths resulted in lower percentages of retained austenite for fixed apply times. The results were inconclusive when applying a fixed magnetic field strength for varying amounts of time. When applying a magnetic field after waiting a specific amount of time after quenching, the analyses indicate that shorter wait times result in less retained austenite. The SAE 1045 results were inconclusive. The samples showed no retained austenite regardless of magnetic field strength, indicating that tempering removed the retained austenite. It is apparent

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

    SciTech Connect

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

    2006-06-15

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

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

    NASA Astrophysics Data System (ADS)

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

    2004-08-01

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

  11. A Feasibility Study on Low Temperature Thermochemical Treatments of Austenitic Stainless Steel in Fluidized Bed Furnace

    NASA Astrophysics Data System (ADS)

    Haruman, Esa; Sun, Yong; Triwiyanto, Askar; Manurung, Yupiter H. P.; Adesta, Erry Y.

    2011-04-01

    In this work, the feasibility of using an industrial fluidized bed furnace to perform low temperature thermochemical treatments of austenitic stainless steels has been studied, with the aim to produce expanded austenite layers with combined wear and corrosion resistance, similar to those achievable by plasma and gaseous processes. Several low temperature thermochemical treatments were studied, including nitriding, carburizing, combined nitridingcarburizing (hybrid treatment), and sequential carburizing and nitriding. The results demonstrate that it is feasible to produce expanded austenite layers on the investigated austenitic stainless steel by the fluidized bed heat treatment technique, thus widening the application window for the novel low temperature processes. The results also demonstrate that the fluidized bed furnace is the most effective for performing the hybrid treatment, which involves the simultaneous incorporation of nitrogen and carbon together into the surface region of the component in nitrogen and carbon containing atmospheres. Such hybrid treatment produces a thicker and harder layer than the other three processes investigated.

  12. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments.

    PubMed

    Sun, C; Zheng, S; Wei, C C; Wu, Y; Shao, L; Yang, Y; Hartwig, K T; Maloy, S A; Zinkle, S J; Allen, T R; Wang, H; Zhang, X

    2015-01-01

    Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304 L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500 °C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M(23)C(6) precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments. PMID:25588326

  13. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments

    SciTech Connect

    Sun, C.; Zheng, S.; Wei, C. C.; Wu, Y.; Shao, L.; Yang, Y.; Hartwig, K. T.; Maloy, S. A.; Zinkle, S. J.; Allen, T. R.; Wang, H.; Zhang, X.

    2015-01-15

    Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M₂₃C₆ precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments.

  14. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments

    DOE PAGESBeta

    Sun, C.; Zheng, S.; Wei, C. C.; Wu, Y.; Shao, L.; Yang, Y.; Hartwig, K. T.; Maloy, S. A.; Zinkle, S. J.; Allen, T. R.; et al

    2015-01-15

    Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size ofmore » ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M₂₃C₆ precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments.« less

  15. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments

    PubMed Central

    Sun, C.; Zheng, S.; Wei, C. C.; Wu, Y.; Shao, L.; Yang, Y.; Hartwig, K. T.; Maloy, S. A.; Zinkle, S. J.; Allen, T. R.; Wang, H.; Zhang, X.

    2015-01-01

    Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M23C6 precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments. PMID:25588326

  16. Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments

    NASA Astrophysics Data System (ADS)

    Sun, C.; Zheng, S.; Wei, C. C.; Wu, Y.; Shao, L.; Yang, Y.; Hartwig, K. T.; Maloy, S. A.; Zinkle, S. J.; Allen, T. R.; Wang, H.; Zhang, X.

    2015-01-01

    Nuclear energy provides more than 10% of electrical power internationally, and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M23C6 precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments.

  17. Mechanical properties of steels with a microstructure of bainite/martensite and austenite islands

    NASA Astrophysics Data System (ADS)

    Syammach, Sami M.

    Advanced high strength steels (AHSS) are continually being developed in order to reduce weight and improve safety for automotive applications. There is need for economic steels with improved strength and ductility combinations. These demands have led to research and development of third generation AHSS. Third generation AHSS include steel grades with a bainitic and tempered martensitic matrix with retained austenite islands. These steels may provide improved mechanical properties compared to first generation AHSS and should be more economical than second generation AHSS. There is a need to investigate these newer types of steels to determine their strength and formability properties. Understanding these bainitic and tempered martensitic steels is important because they likely can be produced using currently available production systems. If viable, these steels could be a positive step in the evolution of AHSS. The present work investigates the effect of the microstructure on the mechanical properties of steels with a microstructure of bainite, martensite, and retained austenite, so called TRIP aided bainitic ferrite (TBF) steels. The first step in this project was creating the desired microstructure. To create a microstructure of bainite, martensite, and austenite an interrupted austempering heat treatment was used. Varying the heat treatment times and temperatures produced microstructures of varying amounts of bainite, martensite, and austenite. Mechanical properties such as strength, ductility, strain hardening, and hole-expansion ratios were then evaluated for each heat treatment. Correlations between mechanical properties and microstructure were then evaluated. It was found that samples after each of the heat treatments exhibited strengths between 1050 MPa and 1350 MPa with total elongations varying from 8 pct to 16 pct. By increasing the bainite and austenite volume fraction the strength of the steel was found to decrease, but the ductility increased. Larger

  18. Intergranular stress distributions in polycrystalline aggregates of irradiated stainless steel

    NASA Astrophysics Data System (ADS)

    Hure, J.; El Shawish, S.; Cizelj, L.; Tanguy, B.

    2016-08-01

    In order to predict InterGranular Stress Corrosion Cracking (IGSCC) of post-irradiated austenitic stainless steel in Light Water Reactor (LWR) environment, reliable predictions of intergranular stresses are required. Finite elements simulations have been performed on realistic polycrystalline aggregate with recently proposed physically-based crystal plasticity constitutive equations validated for neutron-irradiated austenitic stainless steel. Intergranular normal stress probability density functions are found with respect to plastic strain and irradiation level, for uniaxial loading conditions. In addition, plastic slip activity jumps at grain boundaries are also presented. Intergranular normal stress distributions describe, from a statistical point of view, the potential increase of intergranular stress with respect to the macroscopic stress due to grain-grain interactions. The distributions are shown to be well described by a master curve once rescaled by the macroscopic stress, in the range of irradiation level and strain considered in this study. The upper tail of this master curve is shown to be insensitive to free surface effect, which is relevant for IGSCC predictions, and also relatively insensitive to small perturbations in crystallographic texture, but sensitive to grain shapes.

  19. Precipitation and stability of reversed austenite in 9Ni steel

    NASA Astrophysics Data System (ADS)

    Yang, Yue-Hui; Cai, Qing-Wu; Tang, Di; Wu, Hui-Bin

    2010-10-01

    A new method was used to analyze the factors affecting the precipitation of reversed austenite during tempering. The samples were kept at various tempering temperatures for 10 min and their length changes were recorded. Then, the precipitation of reversed austenite which led to the length reduction was shown by thermal expansion curves. The results show that the effects of process parameters on the precipitation of reversed austenite can be determined more accurately by this method than by X-ray diffraction. When the quenching and tempering process is adopted, both the lower quenching temperature and higher tempering temperature can promote the precipitation of reversed austenite during tempering; and when the quenching, lamellarizing, and tempering process is used, intercritical quenching is considered beneficial to the precipitation of reversed austenite in the subsequent tempering because of Ni segregation during holding at the intercritical temperature.

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

    DOEpatents

    Anton, Donald L.; Lemkey, Franklin D.

    1988-01-01

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

  1. Recrystallization Behavior of a Heavily Deformed Austenitic Stainless Steel During Iterative Type Annealing

    NASA Astrophysics Data System (ADS)

    Ravi Kumar, B.; Sharma, Sailaja

    2014-09-01

    The study describes evolution of the recrystallization microstructure in an austenitic stainless steel during iterative or repetitive type annealing process. The starting heavily cold deformed microstructure consisted of a dual phase structure i.e., strain-induced martensite (SIM) (43 pct in volume) and heavily deformed large grained retained austenite. Recrystallization behavior was compared with Johnson Mehl Avrami and Kolmogorov model. Early annealing iterations led to reversion of SIM to reversed austenite. The microstructure changes observed in the retained austenite and in the reverted austenite were mapped by electron backscatter diffraction technique and transmission electron microscope. The reversed austenite was characterized by a fine polygonal substructure consisting of low-angle grain boundaries. With an increasing number of annealing repetitions, these boundaries were gradually replaced by high-angle grain boundaries and recrystallized into ultrafine-grained microstructure. On the other hand, recrystallization of retained austenite grains was sluggish in nature. Progress of recrystallization in these grains was found to take place by a gradual evolution of subgrains and their subsequent transformation into fine grains. The observed recrystallization characteristics suggest continuous recrystallization type process. The analysis provided basic insight into the recrystallization mechanisms that enable the processing of ultrafine-grained fcc steels by iterative type annealing. Tensile properties of the processed material showed a good combination of strength and ductility.

  2. Recrystallization Behavior of a Heavily Deformed Austenitic Stainless Steel During Iterative Type Annealing

    NASA Astrophysics Data System (ADS)

    Ravi Kumar, B.; Sharma, Sailaja

    2014-12-01

    The study describes evolution of the recrystallization microstructure in an austenitic stainless steel during iterative or repetitive type annealing process. The starting heavily cold deformed microstructure consisted of a dual phase structure i.e., strain-induced martensite (SIM) (43 pct in volume) and heavily deformed large grained retained austenite. Recrystallization behavior was compared with Johnson Mehl Avrami and Kolmogorov model. Early annealing iterations led to reversion of SIM to reversed austenite. The microstructure changes observed in the retained austenite and in the reverted austenite were mapped by electron backscatter diffraction technique and transmission electron microscope. The reversed austenite was characterized by a fine polygonal substructure consisting of low-angle grain boundaries. With an increasing number of annealing repetitions, these boundaries were gradually replaced by high-angle grain boundaries and recrystallized into ultrafine-grained microstructure. On the other hand, recrystallization of retained austenite grains was sluggish in nature. Progress of recrystallization in these grains was found to take place by a gradual evolution of subgrains and their subsequent transformation into fine grains. The observed recrystallization characteristics suggest continuous recrystallization type process. The analysis provided basic insight into the recrystallization mechanisms that enable the processing of ultrafine-grained fcc steels by iterative type annealing. Tensile properties of the processed material showed a good combination of strength and ductility.

  3. Gas porosity evolution and ion-implanted helium behavior in reactor ferritic/martensitic and austenitic steels

    NASA Astrophysics Data System (ADS)

    Chernov, I. I.; Kalin, B. A.; Staltsov, M. S.; Oo, Kyi Zin; Binyukova, S. Yu.; Staltsova, O. S.; Polyansky, A. A.; Ageev, V. S.; Nikitina, A. A.

    2015-04-01

    The peculiarities of gas porosity formation and helium retention and release in reactor ferritic/martensitic EP-450 and EP-450-ODS and austenitic ChS-68 steels are investigated by transmission electron microscopy and helium thermal desorption spectrometry (HTDS). The samples were irradiated by 40 keV He+ ions up to a fluence of 5 · 1020 m-2 at 293 and 923 K. An nonuniform distribution of helium bubbles and high-level gas swelling in ferritic/martensitic steels were found at high-temperature helium implantation. The same irradiation conditions result in formation of uniformly distributed helium bubbles and low-level swelling in ChS-68 steel. Temperature range of helium release from EP-450-ODS steel was considerably wider in comparison to HTDS-spectra of the EP-450 steel. A considerable quantity of helium is released from ODS steel in the high-temperature range after the main peak of the HTDS-spectrum.

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

    SciTech Connect

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

    1995-04-01

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

  5. Grain refinement of a nickel and manganese free austenitic stainless steel produced by pressurized solution nitriding

    SciTech Connect

    Mohammadzadeh, Roghayeh Akbari, Alireza

    2014-07-01

    Prolonged exposure at high temperatures during solution nitriding induces grain coarsening which deteriorates the mechanical properties of high nitrogen austenitic stainless steels. In this study, grain refinement of nickel and manganese free Fe–22.75Cr–2.42Mo–1.17N high nitrogen austenitic stainless steel plates was investigated via a two-stage heat treatment procedure. Initially, the coarse-grained austenitic stainless steel samples were subjected to an isothermal heating at 700 °C to be decomposed into the ferrite + Cr{sub 2}N eutectoid structure and then re-austenitized at 1200 °C followed by water quenching. Microstructure and hardness of samples were characterized using X-ray diffraction, optical and scanning electron microscopy, and micro-hardness testing. The results showed that the as-solution-nitrided steel decomposes non-uniformly to the colonies of ferrite and Cr{sub 2}N nitrides with strip like morphology after isothermal heat treatment at 700 °C. Additionally, the complete dissolution of the Cr{sub 2}N precipitates located in the sample edges during re-austenitizing requires longer times than 1 h. In order to avoid this problem an intermediate nitrogen homogenizing heat treatment cycle at 1200 °C for 10 h was applied before grain refinement process. As a result, the initial austenite was uniformly decomposed during the first stage, and a fine grained austenitic structure with average grain size of about 20 μm was successfully obtained by re-austenitizing for 10 min. - Highlights: • Successful grain refinement of Fe–22.75Cr–2.42Mo–1.17N steel by heat treatment • Using the γ → α + Cr{sub 2}N reaction for grain refinement of a Ni and Mn free HNASS • Obtaining a single phase austenitic structure with average grain size of ∼ 20 μm • Incomplete dissolution of Cr{sub 2}N during re-austenitizing at 1200 °C for long times • Reducing re-austenitizing time by homogenizing treatment before grain refinement.

  6. TEM studies of plasma nitrided austenitic stainless steel.

    PubMed

    Stróz, D; Psoda, M

    2010-03-01

    Cross-sectional transmission electron microscopy and X-ray phase analysis were used to study the structure of a layer formed during nitriding the AISI 316L stainless steel at temperature 440 degrees C. It was found that the applied treatment led to the formation of 6-microm-thick layer of the S-phase. There is no evidence of CrN precipitation. The X-ray diffraction experiments proved that the occurred austenite lattice expansion - due to nitrogen atoms - depended on the crystallographic direction. The cross-sectional transmission electron microscopy studies showed that the layer consisted of a single cubic phase that contained a lot of defects such as dislocations, stacking faults, slip bands and twins. The high-resolution electron microscopy observations were applied to study the defect formation due to the nitriding process. It was shown that the presence of great number of stacking faults leads to formation of nanotwins. Weak, forbidden {100} reflections were still another characteristic feature of the S-phase. These were not detected in the X-ray spectra of the phase. Basing on the high-resolution electron microscopy studies it can be suggested that the short-range ordering of the nitrogen atoms in the octahedral sites inside the f.c.c. matrix lattice takes place and gives rise to appearance of these spots. It is suggested that the cubic lattice undergoes not only expansion but also slight rombohedral distortion that explains differences in the lattice expansion for different crystallographic directions. PMID:20500370

  7. Nanoindentation on ion irradiated steels

    NASA Astrophysics Data System (ADS)

    Hosemann, P.; Vieh, C.; Greco, R. R.; Kabra, S.; Valdez, J. A.; Cappiello, M. J.; Maloy, S. A.

    2009-06-01

    Radiation induced mechanical property changes can cause major difficulties in designing systems operating in a radiation environment. Investigating these mechanical property changes in an irradiation environment is a costly and time consuming activity. Ion beam accelerator experiments have the advantage of allowing relatively fast and inexpensive materials irradiations without activating the sample but do in general not allow large beam penetration depth into the sample. In this study, the ferritic/martensitic steel HT-9 was processed and heat treated to produce one specimen with a large grained ferritic microstructure and further heat treated to form a second specimen with a fine tempered martensitic lath structure and exposed to an ion beam and tested after irradiation using nanoindentation to investigate the irradiation induced changes in mechanical properties. It is shown that the HT-9 in the ferritic heat treatment is more susceptible to irradiation hardening than HT-9 after the tempered martensitic heat treatment. Also at an irradiation temperature above 550 °C no detectable hardness increase due to irradiation was detected. The results are also compared to data from the literature gained from the fast flux test facility.

  8. Investigation of fatigue behavior of two austenitic stainless steels

    NASA Astrophysics Data System (ADS)

    Kalnaus, Sergiy

    2009-12-01

    Fatigue of two stainless steels, AISI 304L and AL6-XN, was systematically investigated. While AISI 304L is well known in industry and has been used in engineering applications over the years, AL6-XN is a relatively new alloy and fatigue properties of this material have not been fully investigated by researchers. Both materials belong to one group of austenitic stainless steels. Tension-compression, torsion, and axial-torsion fatigue experiments were conducted on the two alloys to experimentally investigate the cyclic plasticity behavior and the fatigue behavior. Both materials are found to display significant non-proportional hardening. While AISI 304L exhibits cyclic hardening, the AL6-XN alloy displays overall softening under applied cyclic load. Under tension-compression, the cracking plane is perpendicular to the axial loading direction regardless of the loading amplitude for both alloys. The strain-life curves under fully reversed tension-compression and pure torsion for AISI 304L steel are smooth as expected for most metallic materials and can be described by a three-parameter power equation. However, the shear strain-life curve under pure torsion loading for AL6-XN alloy displays a distinct plateau in the fatigue life range approximately from 20,000 to 60,000 loading cycles. The shear strain amplitude corresponding to the plateau is approximately 1.0%. When the shear strain amplitude is above 1.0% under pure shear, the material displays shear cracking. When the shear strain amplitude is below 1.0%, the material displays tensile cracking. A transition from shear cracking to tensile cracking is associated with the plateau in the shear strain-life curve. Three different multiaxial fatigue criteria were evaluated based upon the experimental results on the material for the capability of the criteria to predict fatigue life and the cracking direction. Despite the difference in theory, all the three multiaxial criteria can reasonably correlate the experiments in

  9. Compatibility of martensitic/austenitic steel welds with liquid lead bismuth eutectic environment

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Almazouzi, A.

    2009-04-01

    The high-chromium ferritic/martensitic steel T91 and the austenitic stainless steel 316L are to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both tungsten inert gas (TIG) and electron beam (EB) T91/316L welds have been examined by means of metallography, scanning electron microscopy (SEM-EDX), Vickers hardness measurements and tensile testing both in inert gas and in LBE. Although the T91/316L TIG weld has very good mechanical properties when tested in air, its properties decline sharply when tested in LBE. This degradation in mechanical properties is attributed to the liquid metal embrittlement of the 309 buttering used in TIG welding of T91/316L welds. In contrast to mixed T91/316L TIG welding, the mixed T91/316L EB weld was performed without buttering. The mechanical behaviour of the T91/316L EB weld was very good in air after post weld heat treatment but deteriorated when tested in LBE.

  10. Ferrite and austenite phase identification in duplex stainless steel using SPM techniques

    NASA Astrophysics Data System (ADS)

    Guo, L. Q.; Lin, M. C.; Qiao, L. J.; Volinsky, Alex A.

    2013-12-01

    It can be challenging to properly identify the phases in electro-polished duplex stainless steel using optical microscopy or other characterization techniques. This letter describes magnetic force microscopy to properly identify the phases in electropolished duplex stainless steel. The results are also confirmed with the current sensing atomic force and scanning Kelvin probe force microscopy. The difference in topography heights between the ferrite and austenite phases is attributed to the different etching rates during electropolishing, although these phases have different mechanical properties. The current in the austenite is much higher compared with the ferrite, thus current sensing atomic force microscopy can also be used to properly identify the phases.

  11. Studies on Stress Corrosion Cracking of Super 304H Austenitic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Prabha, B.; Sundaramoorthy, P.; Suresh, S.; Manimozhi, S.; Ravishankar, B.

    2009-12-01

    Stress corrosion cracking (SCC) is a common mode of failure encountered in boiler components especially in austenitic stainless steel tubes at high temperature and in chloride-rich water environment. Recently, a new type of austenitic stainless steels called Super304H stainless steel, containing 3% copper is being adopted for super critical boiler applications. The SCC behavior of this Super 304H stainless steel has not been widely reported in the literature. Many researchers have studied the SCC behavior of steels as per various standards. Among them, the ASTM standard G36 has been widely used for evaluation of SCC behavior of stainless steels. In this present work, the SCC behavior of austenitic Fe-Cr-Mn-Cu-N stainless steel, subjected to chloride environments at varying strain conditions as per ASTM standard G36 has been studied. The environments employed boiling solution of 45 wt.% of MgCl2 at 155 °C, for various strain conditions. The study reveals that the crack width increases with increase in strain level in Super 304H stainless steels.

  12. Melt expulsion during ultrasonic vibration-assisted laser surface processing of austenitic stainless steel.

    PubMed

    Alavi, S Habib; Harimkar, Sandip P

    2015-05-01

    Simultaneous application of ultrasonic vibrations during conventional materials processing (casting, welding) and material removal processes (machining) has recently been gaining widespread attention due to improvement in metallurgical quality and efficient material removal, respectively. In this paper, ultrasonic vibration-assisted laser surface melting of austenitic stainless steel (AISI 316) is reported. While the application of ultrasonic vibrations during laser processing delays the laser interaction with material due to enhancement of surface convection, it resulted in expulsion of melt from the irradiated region (forming craters) and transition from columnar to equiaxed dendritic grain structure in the resolidified melt films. Systematic investigations on the effect of ultrasonic vibrations (with vibrations frequency of 20 kHz and power output in the range of 20-40%) on the development of microstructure during laser surface melting (with laser power of 900 W and irradiation time in the range of 0.30-0.45 s) are reported. The results indicate that the proposed ultrasonic vibration-assisted laser processing can be designed for efficient material removal (laser machining) and improved equiaxed microstructure (laser surface modifications) during materials processing. PMID:25670412

  13. The composition of precipitated austenite in 5.5ni steel

    NASA Astrophysics Data System (ADS)

    Kim, J. I.; Morris, J. W.

    1981-11-01

    Both scanning transmission electron microscopic (STEM) and chemical extraction techniques were used to analyze the chemical content of precipitated austenite in 5.5Ni steel as a function of heat treatment. Austenite was introduced into the steel by tempering at 600 °C for 2 h (QT2) or 100 h (QT100), by tempering for 1 h at 670 °C (QL), or by double tempering 1 h at 670 °C + 1 h at 600 °C (QLT). The two methods of chemical analysis employed for the analysis of this austenite differ quantitatively in the measured austenite composition but are in qualitative agreement; they show an austenite enrichment in Ni, Cr, Mn, and Mo which is most pronounced (and nearly equivalent) for the QT100 and QLT treatments. The similar enrichment of the QT100 and QLT material is interpreted in light of the sequence of reactions leading to the QLT structure. A good correlation is found between the apparent solute enrichment of the precipitated austenite and its thermal stability on cooling to 77 K.

  14. On the measurement of austenite in supermartensitic stainless steel by X-ray diffraction

    SciTech Connect

    Tolchard, Julian Richard; Sømme, Astri; Solberg, Jan Ketil; Solheim, Karl Gunnar

    2015-01-15

    Sections of a 13Cr supermartensitic stainless steel were investigated to determine the optimum sample preparation for measurement of the austenite content by X-ray diffraction. The surface of several samples was mechanically ground or polished using media of grit sizes in the range 1–120 μm. The strained surface layer was afterwards removed stepwise by electropolishing, and the austenite content measured at each step. It was found that any level of mechanical grinding or polishing results in a reduction of the measured austenite fraction relative to the true bulk value, and that coarser grinding media impart greater damage and greater reduction in the measured austenite content. The results thus highlight the importance of the electropolishing step in preparation of such samples, but suggest that the American Society for Testing and Materials standard E975-03 substantially overestimates the amount of material which needs to be removed to recover the true “bulk” content. - Highlights: • Quantitative Rietveld analysis of austenite/martensite ratio in supermartensitic stainless steels • Critical evaluation of sample preparation for residual austenite measurements by X-ray diffraction • Highlighting of the importance of electropolishing as a final preparation step.

  15. Influence of reverted austenite on the texture and magnetic properties of 350 maraging steel

    NASA Astrophysics Data System (ADS)

    Abreu, Hamilton F. G.; Silva, Jean J.; Silva, Manoel R.; Gomes da Silva, Marcelo J.

    2015-11-01

    The aging temperature to improve magnetic properties in Maraging-350 steel (Mar-350) is limited by the onset of austenite reversion. The traditional process of cooling after aging is to remove the piece from the oven and then to air cool it. The purpose of this research was to characterize the reverted austenite and to investigate the effect of cooling below the martensite start temperature (Ms) on the magnetic properties. The Mar350 samples aged at temperatures above 550 °C, and subsequently cooled in liquid nitrogen presented less austenite than samples cooled in air, resulting in higher magnetization saturation and a lower coercive force. A combination of optical microscopy (OM), X-ray diffraction (XRD) and electron backscatter diffraction (EBSD) techniques were used to characterize the presence of reverted austenite. The crystallographic texture of both martensite and reverted austenite were analyzed. The texture of the reverted austenite coincides with the texture of the parent austenite indicating that a phenomenon of texture memory is present.

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

    SciTech Connect

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

    2007-01-01

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

  17. Characterization of the Carbon and Retained Austenite Distributions in Martensitic Medium Carbon, High Silicon Steel

    NASA Astrophysics Data System (ADS)

    Sherman, Donald H.; Cross, Steven M.; Kim, Sangho; Grandjean, Fernande; Long, Gary J.; Miller, Michael K.

    2007-08-01

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

  18. Microstructural effects on the stability of retained austenite in Transformation Induced Plasticity steels

    NASA Astrophysics Data System (ADS)

    Mark, Alison Fiona Lockie

    Transformation Induced Plasticity (TRIP) steels have both high strength and high ductility. Retained austenite in the microstructure, upon straining, transforms to martensite and this absorbs energy and improves the work hardening of the steel, giving improved elongation. The transformation can be either stress-assisted or strain-induced and the initiation and the mechanism depend on the composition of, the size and shape of, and the phases surrounding, the austenite grains. It is important to understand the relationship between these variables and the properties of the TRIP steel. The aim of this work was to determine how the microstructure of the TRIP steel affects the transformation. Four experimental microstructures were developed, containing austenite grains with different sizes, shapes, and surrounding phases. The Fine microstructure had thin elongated austenite laths between fine bainitic ferrite laths, the Coarse microstructure had elongated austenite grains between coarser bainitic ferrite laths, the Equiaxed microstructure had equiaxed austenite grains in a matrix of equiaxed ferrite and the Acicular microstructure had elongated austenite grains surrounded by recovered ferrite laths. Tensile tests were performed and detailed characterization, using neutron diffraction, was done of samples with the four microstructures. The variation in the amount of austenite during deformation was measured. The tensile tests revealed that the microstructures had different mechanical properties and different transformation behaviours. Fine had the lowest elongation and the highest strength. Acicular and Equiaxed had good elongation but lower strength. Coarse had intermediate strength and Equiaxed had sustained work hardening. The transformation in Fine and Coarse was minimal. Coarse had some slow, steady transformation, but Fine may have had none. The transformation in Equiaxed was larger. It started quickly and then slowed at higher strains. The austenite in Acicular

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

    SciTech Connect

    Huang, F.H.

    1992-02-01

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

  20. A role of {delta}-ferrite in edge-crack formation during hot-rolling of austenitic stainless steels

    SciTech Connect

    Czerwinski, F.; Brodtka, A.; Cho, J.Y.; Szpunar, J.A.; Zielinska-Lipiec, A.; Sunwoo, J.H.

    1997-10-15

    Austenitic stainless steels are substantially harder during hot-rolling than either ferritic or mild steels. The objective of this study is to verify the possible correlation between the edge-crack formation during hot-rolling and the presence of {delta} ferrite in austenitic stainless steel. Hot-rolled plates of austenitic stainless steels, examined at room temperatures, contain up to 9% of {delta} ferrite in austenitic matrix. The distribution of ferrite in steel plate is inhomogeneous: the highest ferrite content is located in the vicinity of the plate edge. Moreover, the content of {delta} ferrite changes irregularly across the plate thickness. The results obtained from analysis of several plates suggest a correlation between the maximum content of {delta} ferrite in steel microstructure and the length of the edge-crack formed during hot-rolling: the higher the volume fraction of ferrite, the longer the edge-crack.

  1. Micro-mechanical investigation for effects of helium on grain boundary fracture of austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Miura, Terumitsu; Fujii, Katsuhiko; Fukuya, Koji

    2015-02-01

    Effects of helium (He) on grain boundary (GB) fracture of austenitic stainless steel were investigated by micro-tensile tests. Micro-bicrystal tensile specimens were fabricated for non-coincidence site lattice boundaries of He ion-irradiated 316 stainless steel by focused ion beam (FIB) micro-processing. Micro-tensile tests were conducted in a vacuum at room temperature in the FIB system. Specimens containing more than 2 at.% He fractured at GBs. The criteria for brittle fracture occurrence on GBs were: (1) He concentrations higher than 2 at.%; (2) formation of He bubbles on the GBs with less than a 5 nm spacing; and (3) matrix hardening to more than 4.6 GPa (nano-indentation hardness). The fracture stress of GB brittle fracture was lower for a specimen with higher He concentration while the size and areal density of the GB He bubbles were the same. The specimens that contained 10 at.% He and had been annealed at 923 K after irradiation fractured at the GB nominally in a brittle manner; however the inter-bubble matrix at the GB experienced ductile fracture. The annealing caused He bubbles to grow but decreased the areal density so that the spacing of the GB He bubbles widened and the hardness decreased, therefore the fracture mode changed from brittle to ductile. The findings revealed that He promotes GB fracture by weakening the GB strength and hardening the matrix due to the formation of He bubbles both on GBs and in the matrix. In addition, the findings suggested that GB segregated He atoms may have a role in GB fracture.

  2. Review of environmental effects on fatigue crack growth of austenitic stainless steels

    SciTech Connect

    Shack, W.J.; Kassner, T.F.

    1994-05-01

    Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry.

  3. Solidification behavior and microstructural analysis of austenitic stainless steel laser welds

    SciTech Connect

    David, S.A.; Vitek, J.M.

    1981-01-01

    Solidification behavior of austenitic stainless steel laser welds has been investigated with a high-power laser system. The welds were made at speeds ranging from 13 to 60 mm/s. The welds sowed a wide variety of microstructural features. The ferrite content in the 13-mm/s weld varied from less than 1% at the root of the weld to about 10% at the crown. The duplex structure at the crown of the weld was much finer than the one observed in conventional weld metal. However, the welds made at 25 and 60 mm/s contained an austenitic structure with less than 1% ferrite throughout the weld. Microstructural analysis of these welds used optical microscopy, transmission electron microscopy, and analytical electron microscopy. The austenitic stainless steel welds were free of any cracking, and the results are explained in terms of the rapid solidification conditions during laser welding.

  4. Effect of austenitizing temperature on microstructure in 16Mn steel treated by cerium

    NASA Astrophysics Data System (ADS)

    Wen, Bin; Song, Bo; Pan, Ning; Hu, Qing-Yun; Mao, Jing-Hong

    2011-12-01

    The change of inclusions and microstructure of 16Mn steel treated by Ce were observed, and the effect of austenitizing temperature on the microstructure was also examined. The results show that the inclusions are transformed from Si-Mn-Al composite oxide and MnS into AlCeO3, Ce2O2S, and MnS composite inclusions after being treated by Ce. Plenty of intragranular ferrites are formed in 16Mn steel containing ˜0.017wt% Ce. A large amount of intragranular acicular ferrites are formed after being austenitized for 20 min at 1473 K. The prior austenite grain size fit for the formation of intragranular acicular ferrites is about 120 μm.

  5. Estimation of fatigue strain-life curves for austenitic stainless steels in light water reactor environments.

    SciTech Connect

    Chopra, O. K.; Smith, J. L.

    1998-02-12

    The ASME Boiler and Pressure Vessel Code design fatigue curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue lives of austenitic stainless steels (SSs) in light water reactor (LWR) environments. Unlike those of carbon and low-alloy steels, environmental effects on fatigue lives of SSs are more pronounced in low-dissolved-oxygen (low-DO) water than in high-DO water, This paper summarizes available fatigue strain vs. life data on the effects of various material and loading variables such as steel type, DO level, strain range, and strain rate on the fatigue lives of wrought and cast austenitic SSs. Statistical models for estimating the fatigue lives of these steels in LWR environments have been updated with a larger data base. The significance of the effect of environment on the current Code design curve has been evaluated.

  6. The effect of hydrogen on strain hardening and fracture mechanism of high-nitrogen austenitic steel

    NASA Astrophysics Data System (ADS)

    Maier, G. G.; Astafurova, E. G.; Melnikov, E. V.; Moskvina, V. A.; Vojtsik, V. F.; Galchenko, N. K.; Zakharov, G. N.

    2016-07-01

    High-nitrogen austenitic steels are perspective materials for an electron-beam welding and for producing of wear-resistant coatings, which can be used for application in aggressive atmospheres. The tensile behavior and fracture mechanism of high-nitrogen austenitic steel Fe-20Cr-22Mn-1.5V-0.2C-0.6N (in wt.%) after electrochemical hydrogen charging for 2, 10 and 40 hours have been investigated. Hydrogenation of steel provides a loss of yield strength, uniform elongation and tensile strength. The degradation of tensile properties becomes stronger with increase in charging duration - it occurs more intensive in specimens hydrogenated for 40 hours as compared to ones charged for 2-10 hours. Fracture analysis reveals a hydrogen-induced formation of brittle surface layers up to 6 μm thick after 40 hours of saturation. Hydrogenation changes fracture mode of steel from mixed intergranular-transgranular to mainly transgranular one.

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

    DOEpatents

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

    2012-11-27

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

  8. Acoustic Emission Technique for Characterizing Deformation and Fatigue Crack Growth in Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Mukhopadhyay, C. K.; Jayakumar, T.

    2003-03-01

    Acoustic emission (AE) during tensile deformation and fatigue crack growth (FCG) of austenitic stainless steels has been studied. In AISI type 316 stainless steel (SS), AE has been used to detect micro plastic yielding occurring during macroscopic plastic deformation. In AISI type 304 SS, relation of AE with stress intensity factor and plastic zone size has been studied. In AISI type 316 SS, fatigue crack growth has been characterised using acoustic emission.

  9. Delta ferrite-containing austenitic stainless steel resistant to the formation of undesirable phases upon aging

    SciTech Connect

    Leitnaker, J.M.

    1981-05-05

    Austenitic stainless steel alloys containing delta ferrite, such as are used as weld deposits, are protected against the transformation of delta ferrite to sigma phase during aging by the presence of carbon plus nitrogen in a weight percent 015-0.030 times the volume percent ferrite present in alloy. The formation of chi phase upon aging is controlled by controlling the mo content.

  10. Determination of Proper Austenitization Temperatures for Hot Stamping of AISI 4140 Steel

    NASA Astrophysics Data System (ADS)

    Samadian, Pedram; Parsa, Mohammad Habibi; Shakeri, Amid

    2014-04-01

    High strength steels are desirable materials for use in automobile bodies in order to reduce vehicle weight and increase the safety of car passengers, but steel grades with high strength commonly show poor formability. Recently, steels with controlled microstructures and compositions are used to gain adequate strength after hot stamping while maintaining good formability during processing. In this study, microstructure evolutions and changes in mechanical properties of AISI 4140 steel sheets resulting from the hot stamping process at different austenitization temperatures were investigated. To determine the proper austenitization temperatures, the results were compared with those of the cold-worked and cold-worked plus quench-tempered specimens. Comparisons showed that the austenitization temperatures of 1000 and 1100 °C are proper for hot stamping of 3-mm-thick AISI 4140 steel sheets due to the resultant martensitic microstructure which led to the yield and ultimate tensile strength of 1.3 and 2.1 GPa, respectively. Such conditions resulted in more favorable simultaneous strength and elongation than those of hot-stamped conventional boron steels.