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Sample records for austenitic alloys irradiated

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

    Science.gov (United States)

    Gan, J.; Was, G. S.

    2001-08-01

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

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

    Science.gov (United States)

    Gan, Jian

    1999-11-01

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

  3. An assessment of Fe-Cr-Mn austenitic alloys for fusion service using fast reactor irradiation

    International Nuclear Information System (INIS)

    A series of model Fe-Cr-Mn alloys and various solute-modified high manganese alloys have been irradiated in the Fast Flux Test Facility in order to provide an early assessment of the consequences of substituting manganese for nickel in austenitic stainless steel. The purpose of this substitution is to reduce the level of long term radioactivation of this alloy, a candidate structural material for use in fusion energy devices. Simple Fe-Mn and Fe-Cr-Mn alloys were found to exhibit much of the same behavior observed in Fe-Ni and Fe-Ni-Cr alloys. In particular, they tend to swell at /approximately/1%/dpa after an incubation period that is dependent on irradiation temperature, alloy composition and cold working. The phase stability, both in and out of reactor, is altered substantially by the substitution of manganese, however. It nonetheless appears that appropriate levels of solute modification can be used to improve both the swelling resistance and phase stability. An alloy with a base composition of Fe-20Mn-15Cr appears to offer the best promise for further research. 24 refs., 9 figs., 4 tabs

  4. Neutron irradiation creep experiments on austenitic stainless steel alloys

    International Nuclear Information System (INIS)

    Results of measurements of the neutron induced creep elongation on AMCR-steels (Mn-base), on 316 CE-reference steels, and on US 316 jand US PCA steels are reported. It was found that the stationary creep rate is not very sensitive to variations of the irradiation temperature between 300 and 400 degC and that the stress-exponent of plastically deformed and of annealed materials is n ≅ and n ≅ 1.59, respectively. A small primary creep stage is found in annealed materials. Deformed materials show a negative creep elongation at the beginning of the irradiation, which increases for decreasing stresses and decreases for increasing irradiation temperatures. (author). 7 refs.; 7 figs.; 1 tab

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

    International Nuclear Information System (INIS)

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

  6. Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe–21Cr–32Ni)

    International Nuclear Information System (INIS)

    An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe–21Cr–32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and γ′-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly γ′ precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements

  7. Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe-21Cr-32Ni)

    International Nuclear Information System (INIS)

    An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 deg C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and y'-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly y' precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

  8. Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe-21Cr-32Ni)

    Energy Technology Data Exchange (ETDEWEB)

    L. Tan; J. T. Busby; H. J. M. Chichester; K. Sridharan; T. R. Allen

    2013-06-01

    An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 degrees C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and y'-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly y' precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

  9. Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe–21Cr–32Ni)

    Energy Technology Data Exchange (ETDEWEB)

    Tan, L., E-mail: tanl@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Busby, J.T. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Chichester, H.J.M. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Sridharan, K.; Allen, T.R. [University of Wisconsin, Madison, WI 53706 (United States)

    2013-06-15

    An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe–21Cr–32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 °C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and γ′-Ni{sub 3}(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly γ′ precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  11. The Primary Origin of Dose Rate Effects on Microstructural Evolution of Austenitic Alloys During Neutron Irradiation

    International Nuclear Information System (INIS)

    The effect of dose rate on neutron-induced microstructural evolution was experimentally estimated. Solution-annealed austenitic model alloys were irradiated at approximately 400 degrees C with fast neutrons at seven different dose rates that vary more than two orders difference in magnitude, and two different doses were achieved at each dose rate. Both cavity nucleation and growth were found to be enhanced at lower dose rate. The net vacancy flux is calculated from the growth rate of cavities that had already nucleated during the first cycle of irradiation and grown during the second cycle. The net vacancy flux was found to be proportional to (dpa/sec) exp (1/2) up to 28.8 dpa and 8.4 x 10 exp (-7) dpa/sec. This implies that mutual recombination dominates point defect annihilation, in this experiment even though point defect sinks such as cavities and dislocations were well developed. Thus, mutual recombination is thought to be the primary origin of the effect of dose rate on microstructural evolution

  12. Precipitation at grain boundaries in irradiated austenitic Fe-Cr-Mn alloys

    International Nuclear Information System (INIS)

    In previous work, the phase stability of Fe-Cr-Mn alloys during irradiation was investigated in a study that included simple binaries, simple ternaries and commercially produced alloys. These low activation alloys are being considered for fusion reactor service in the first wall and in other structural applications subject to high neutron doses. In addition to phase instabilities observed within the grains, grain boundaries were susceptible to varying levels of precipitation dependent upon alloy composition, displacement dose and irradiation temperature. This paper describes the grain boundary microstructures that developed in these Fe-Cr-Mn alloys during irradiation

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-16

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

  16. Effect of Ge, Sn, Sb on the resistance to swelling of austenitic alloys irradiated by 1 MeV electrons

    International Nuclear Information System (INIS)

    The effect of new solute elements namely Ge, Sn and Sb on the void swelling resistance of austenitic alloys irradiated with 1 MeV electrons has been studied. Except for tin in Ti-modified 316, all solute improve the swelling resistance of base alloys. Tin addition shifts the swelling peak of 316 S.S. to high temperature. In fact, these solute additions have the same qualitative effect on the swelling components: they enhance the void density and decrease strongly void growth rate. This effect is opposite to the one of usual swelling inhibitors such as Si or Ti which decrease the void density. We have explained this influence on the void nucleation and void growth by introducing a strong interaction between vacancies and solute atoms in a void growth model

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  20. The influence of titanium additions on the swelling of austenitic steels and nickel alloys irradiated with electrons

    International Nuclear Information System (INIS)

    It is shown that the addition of titanium is beneficial to the swelling behaviour of austenitic steels. The magnitude of the observed effects depends greatly on the nature and concentration of the other minor elements in the austenite matrix. (author)

  1. Investigation of joining techniques for advanced austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  2. Tensile and fracture toughness properties of copper alloys and their HIP joints with austenitic stainless steel in unirradiated and neutron irradiated condition

    Energy Technology Data Exchange (ETDEWEB)

    Taehtinen, S.; Pyykkoenen, M. [VTT Manufacturing Technology, Espoo (Finland); Singh, B.N.; Toft, P. [Risoe National Lab., Roskilde (Denmark). Materials Research Dept.

    1998-03-01

    The tensile strength and ductility of unirradiated CuAl25 IG0 and CuCrZr alloys decreased continuously with increasing temperature up to 350 deg C. Fracture toughness of unirradiated CuAl25 IG0 alloy decreased continuously with increasing temperature from 20 deg C to 350 deg C whereas the fracture toughness of unirradiated CuCrZr alloy remained almost constant at temperatures up to 100 deg C, was decreased significantly at 200 deg C and slightly increased at 350 deg C. Fracture toughness of HIP joints were lower than that of corresponding copper alloy and fracture path in HIP joint specimen was always within copper alloy side of the joint. Neutron irradiation to a dose level of 0.3 dpa resulted in hardening and reduction in uniform elongation to about 2-4% at 200 deg C in both copper alloys. At higher temperatures softening was observed and uniform elongation increased to about 5% and 16% for CuAl25 IG0 and CuCrZr alloys, respectively. Fracture toughness of CuAl25 IG0 alloy reduced markedly due to neutron irradiation in the temperature range from 20 deg C to 350 deg C. The fracture toughness of the irradiated CuCrZr alloy also decreased in the range from 20 deg C to 350 deg C, although it remained almost unaffected at temperatures below 200 deg C and decreased significantly at 350 deg C when compared with that of unirradiated CuCrZr alloy. (orig.)

  3. Tensile and fracture toughness properties of copper alloys and their HIP joints with austenitic stainless steel in unirradiated and neutron irradiated condition

    International Nuclear Information System (INIS)

    The tensile strength and ductility of unirradiated CuAl25 IG0 and CuCrZr alloys decreased continuously with increasing temperature up to 350 deg C. Fracture toughness of unirradiated CuAl25 IG0 alloy decreased continuously with increasing temperature from 20 deg C to 350 deg C whereas the fracture toughness of unirradiated CuCrZr alloy remained almost constant at temperatures up to 100 deg C, was decreased significantly at 200 deg C and slightly increased at 350 deg C. Fracture toughness of HIP joints were lower than that of corresponding copper alloy and fracture path in HIP joint specimen was always within copper alloy side of the joint. Neutron irradiation to a dose level of 0.3 dpa resulted in hardening and reduction in uniform elongation to about 2-4% at 200 deg C in both copper alloys. At higher temperatures softening was observed and uniform elongation increased to about 5% and 16% for CuAl25 IG0 and CuCrZr alloys, respectively. Fracture toughness of CuAl25 IG0 alloy reduced markedly due to neutron irradiation in the temperature range from 20 deg C to 350 deg C. The fracture toughness of the irradiated CuCrZr alloy also decreased in the range from 20 deg C to 350 deg C, although it remained almost unaffected at temperatures below 200 deg C and decreased significantly at 350 deg C when compared with that of unirradiated CuCrZr alloy. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

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

  9. The influence of helium on mechanical properties of model austenitic alloys, determined using 59Ni isotopic tailoring and fast reactor irradiation

    International Nuclear Information System (INIS)

    The objective of this effort is to study the separate and synergistic effects of helium and other important variables on the evolution of microstructure and macroscopic properties during irradiation of structural metals. The alloys employed in this study were nominally Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr-45Ni (wt %) in both the cold worked and annealed conditions. Tensile testing and microscopy continue on specimens removed from the first, second and third discharges of the 59Ni isotopic doping experiment. The results to date indicate that helium/dpa ratios typical of fusion reactors (4 to 19 appm/dpa) do not lead to significant changes in the yield strength of model Fe-Cr-Ni alloys. Measurements of helium generated in undoped specimens from the second and third discharges show that the helium/dpa ratio increases during irradiation in FFTF due to the production of 59Ni. In specimens doped with 59Ni prior to irradiation, the helium/dpa ratio can increase, decrease or remain the same during the second irradiation interval. This behavior occurs because the cross sections for the production and burnout of 59Ni are very sensitive to core location and the nature of neighboring components. 14 refs., 5 figs., 3 tabs

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

    Science.gov (United States)

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

    2013-07-01

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

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

    Science.gov (United States)

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

    2001-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-21

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  15. Study of irradiation damage structures in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

    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. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-01

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

  18. Characterization and understanding of ion irradiation effect on the microstructure of austenitic stainless steels

    International Nuclear Information System (INIS)

    Austenitic stainless steels are widely used in nuclear industry for internal structures. These structures are located close to the fuel assemblies, inside the pressure vessel. The exposure of these elements to high irradiation doses (the accumulated dose, after 40 years of operation, can reach 80 dpa), at temperature close to 350 C, modifies the macroscopic behavior of the steel: hardening, swelling, creep and corrosion are observed. Moreover, in-service inspections of some of the reactor internal structures have revealed the cracking of some baffle bolts. This cracking has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). In order to understand this complex phenomenon, a first step is to identify the microstructural changes occurring during irradiation, and to understand the mechanisms at the origin of this evolution. In this framework, a large part of the European project 'PERFORM 60' is dedicated to the study of the irradiation damage in austenitic stainless steels. The objective of this PhD work is to bring comprehensive data on the irradiation effects on microstructure. To reach this goal, two model alloys (FeNiCr and FeNiCrSi) and an industrial austenitic stainless steel (316 steel) are studied using Atom Probe Tomography (APT), Transmission Electron Microscope (TEM) and Positron Annihilation Spectroscopy (PAS). They are irradiated by Ni ions in CSNSM (Orsay) at two temperatures (200 and 450 C) and three doses (0.5, 1 and 5 dpa). TEM observations have shown the appearance of dislocation loops, cavities and staking fault tetrahedra. The dislocation loops in 316 steel were preferentially situated in the vicinity of dislocations, while they were randomly distributed in the FeNiCr alloy. APT study has shown the redistribution of Ni and Si under irradiation in FeNiCrSi model alloy and 316 steel, leading to the appearance of (a) Cottrell clouds along dislocation lines, dislocation loops and other non-identified crystalline defects and (b

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.

    1996-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

  5. Radiation damage simulation studies of selected austenitic and ferritic alloys

    International Nuclear Information System (INIS)

    Results are presented of a study of the radiation damage structure of selected alloys following ion bombardment to simulate fusion-reactor exposures up to 40 dpa (approx. 3 MW-yr m-2) at temperatures from 4750C-6500C. Gas concentrations appropriate to fusion-reactor conditions were simulated using a mixed gas beam of 4 MeV He + 2 MeV H2. A beam of 46 MeV Ni ions was used in sequence with the gas beam to provide gas-to-damage ratios of 13 appm He/dpa and 52 appm H/dpa at a nickel-ion damage rate of approx. 1 dpa/hr. The materials investigated comprised three austenitic stainless steels (316L, modified 316-Ti and 316-Nb), a ferritic alloy (1.4914) and a commercial low-activation alloy containing Mn (TENELON). The results reveal that ferritic steels have good radiation damage resistance and are far superior to austenitic steels in respect of void-induced swelling. (author)

  6. Defect microstructures and deformation mechanisms in irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Microstructural evolution and deformation behavior of austenitic stainless steels are evaluated for neutron, heavy-ion and proton irradiated materials. Radiation hardening in austenitic stainless steels is shown to result from the evolution of small interstitial dislocation loops during light-water-reactor (LWR) irradiation. Available data on stainless steels irradiated under LWR conditions have been analyzed and microstructural characteristics assessed for the critical fluence range (0.5 too 10 dpa) where irradiation-assisted stress corrosion cracking susceptibility is observed. Heavy-ion and proton irradiations are used to produce similar defect microstructures enabling the investigation of hardening and deformation mechanisms. Scanning electron, atomic force and transmission electron microscopies are employed to examine tensile test strain rate and temperature effects on deformation characteristics. Dislocation loop microstructures are found to promote inhomogeneous planar deformation within the matrix and regularly spaced steps at the surface during plastic deformation. Twinning is the dominant deformation mechanism at rapid strain rates and at low temperatures, while dislocation channeling is favored at slower strain rates and at higher temperatures. Both mechanisms produce highly localized deformation and large surface slip steps. Channeling, in particular, is capable of creating extensive dislocation pileups and high stresses at internal grain boundaries which may promote intergranular cracking

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

    International Nuclear Information System (INIS)

    Microstructural evolution in solution-annealed Japanese-PCA (JPCA-SA) and four other austenitic stainless steels, irradiated at 400 deg. C to 17.3 dpa in the ORR and the high flux isotope reactor (HFIR) spectrally tailored experiment, were investigated by transmission electron microscopy (TEM). The mean He/dpa ratio throughout the irradiation fell between 12 and 16 appm He/dpa , which is close to the He/dpa values expected for fusion. In all the specimens, a bi-modal size distribution of cavities was observed and the number densities were about 1.0x1022 m-3. There was no significant difference between the number densities in the different alloys, although the root mean cubes of the cavity radius are quite different for each alloy. Precipitates of the MC type were also observed in the matrix and on grain boundaries in all alloys except a high-purity (HP) ternary alloy. The JPCA-SA (including 0.06% carbon and 0.027% phosphorus) and standard type 316 steel (including 0.06% carbon and 0.028% phosphorus) showed quite low-swelling values of about 0.016 and 0.015%, respectively, while a HP ternary austenitic alloy showed the highest swelling value of 2.9%. This suggests that the existence of impurities affects the cavity growth in austenitic stainless steels even at 400 deg. C

  8. High temperature irradiation creep in austenitic steels

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-01

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

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

    Science.gov (United States)

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

    2010-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

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

    Science.gov (United States)

    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 ||TA (tensile axis) and ||TA were twin free and grain with ||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

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

    Institute of Scientific and Technical Information of China (English)

    Selva Büyükakkas; H Aktas; S Akturk

    2007-01-01

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

  16. Electron irradiation induced solute segregation near grain boundaries in austenitic stainless steel

    International Nuclear Information System (INIS)

    Radiation-induced solute segregation near internal defect sinks such as high angle grain boundaries was investigated, through the interaction between point defects and solute atom in austenitic stainless steel and its model alloys. Electron irradiation was performed in a high voltage electron microscope (H V E M) at a dose rate of about 2 multiple 10-3 d p a.S-1 at a temperature range of 350-600 degree C. Solute concentration profile near grain boundaries was measured by E D X in S T E M mode. Strong enrichment and depletion of solutes were observed on grain boundaries during irradiation and segregation rate went through a maximum at 450 degree C. These facts indicate that grain boundaries act as preferential sinks for radiation-induced point defects

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

    Directory of Open Access Journals (Sweden)

    Rouxel Baptiste

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  19. Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

    International Nuclear Information System (INIS)

    Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x1024n/m2 (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

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

    Institute of Scientific and Technical Information of China (English)

    Jie HUANG; Zhou XU; Xin XING

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Microstructural evolution in austenitic stainless steels subjected to displacement damage at high temperature is strongly influenced by the interactions between helium atoms and second phase particles. Cavity nucleation occurs by the trapping of helium at partially coherent particle-matrix interfaces. The recent precipitate point defect collector theory describes the more rapid growth of precipitate-attached cavities compared to matrix cavities where the precipitate-matrix interface collects point defects to augment the normal point deflect flux to the cavitry. Data are presented which support these ideas. It is shown that during nickel ion irradiation of a titanium-modified stainless steel at 6750C the rate of injection of helium has a strong effect on the total swelling and also on the nature and distribution of precipitate phases

  2. Development of highest strength nitrogen alloyed austenitic steels

    International Nuclear Information System (INIS)

    This work deals with different possibilities to increase the strength of austenitic stainless steels. It may be interesting to the steel producer and to the steel user, because it shows ways to increase the strength of such steels. It may also be of interest to the metallurgist, because it analyzes the achieved results on the basis of physical metallurgy. It is shown that the increase of the nitrogen content in solid solution has a positive effect on the different hardening mechanisms. The investigation, therefore, focus on nitrogen alloyed steels. Grain boundary hardening, the increase of strength with decreasing strain size, is strongly pronounced in nitrogen alloyed steels. By means of an ultra rapid recrystallization heat treatment it is possible to produce a grain size as small as 2-3 microns. The yield strength reaches an unusually high value of 1030 MPa with an elongation of 48% and an area reduction of 63%. This completely recrystallized steel is free of any precipitation. From the Hall-Petch relation the yield strength may only be extrapolated down to a grain size of about 10 microns. The slope of the Hall-Petch plot, ky, decreases with decreasing grain size. This behavior may be explained by the different distribution of dislocation glide in fine and in coarse grained material. Strain hardening, the increase of strength due to deformation, was thoroughly investigated as a second hardening mechanism. The most important results for room temperature deformation are presented. Deformation at an increased temperature of about 300-400 oC was investigated as an alternative to room temperature deformation. The potentially useful results are also presented. Strain aging may further increase the strength of cold worked nitrogen alloys steels. A heat treatment between 300 and 500oC for a few minutes may lead to an increase in strength of more than 300 MPa. (author) 73 figs., refs

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

    Institute of Scientific and Technical Information of China (English)

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

    2004-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

    The dynamics of deformation localization and dislocation channel formation were investigated in situ in a neutron-irradiated AISI 304 austenitic stainless steel and a model 304-based austenitic alloy by combining several analytical techniques including optic microscopy and laser confocal microscopy, scanning electron microscopy, electron backscatter diffraction, and transmission electron microscopy (TEM). Channel formation was observed at ∼70% of the polycrystalline yield stress of the irradiated materials (σ{sub 0.2}). It was shown that triple junction points do not always serve as a source of dislocation channels; at stress levels below the σ{sub 0.2}, channels often formed near the middle of the grain boundary. For a single grain, the role of elastic stiffness value (Young’s modulus) in channel formation was analyzed; it was shown that in the irradiated 304 steels the initial channels appeared in “soft” grains with a high Schmid factor located near “stiff” grains with high elastic stiffness. The spatial organization of channels in a single grain was analyzed; it was shown that secondary channels operating in the same slip plane as primary channels often appeared at the middle or at one-third of the way between primary channels. The twinning nature of dislocation channels was analyzed for grains of different orientation using TEM. In the AISI 304 steel, channels in grains oriented close to 〈0 0 1〉||TA (tensile axis) and 〈1 0 1〉||TA were twin free and grain with 〈1 1 1〉||TA and grains oriented close to a Schmid factor maximum contained deformation twins.

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

    International Nuclear Information System (INIS)

    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

  8. Regularities of structure formation during hot deformation of austenite in alloy steels

    International Nuclear Information System (INIS)

    Regularities of substructure formation during hot working of austenite in 110Kh6 and 40Kh8G8 alloy steels, structural peculiarities and relations between structure development and a hot deformation curve were investigated. The possibility of structure formation modeling is also evaluated for deformation under commercial procedure conditions. Hot deformation during high temperatue thermomechanical treatment was carried out by rolling and compression. It is found that in alloy steel austenite during hot deformation up to 7-10% the processes of intensive strain hardening develop which result in formation of substructure with high density of dislocations either distributed uniformly or forming a cellular type substructure. Strain softening processes (dynamic polygonization) arise with a deformation degree increase. The relationship found between a hot deformation curve and structural changes during hot working of alloy steel austenite provides the option for conditions of high temperature thermomechanical treatment of commerical alloy steels softening according to a dynamic polygonization mechanism

  9. Oxidization and stress corrosion cracking initiation of austenitic alloys in supercritical water

    International Nuclear Information System (INIS)

    This study determined the stress corrosion cracking behaviour of austenitic alloys in pure supercritical water. Austenitic stainless steels 310S, 316L, and Inconel 625 were tested as static capsule samples at 500oC for up to 5000 h. After that period, crack initiations were readily observed in all samples, signifying susceptibility to stress corrosion cracking. The microcracks in 316L stainless steel and Inconel 625 were almost intergranular, whereas transgranular microcrack initiation was observed in 310S stainless steel. (author)

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

    International Nuclear Information System (INIS)

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

  11. Lean-alloyed austenitic stainless steel with high resistance against hydrogen environment embrittlement

    International Nuclear Information System (INIS)

    Highlights: · Hydrogen environment embrittlement of austenitic steel. · Novel alloying concept for austenitic stainless steel with improved HEE resistance. · Influence of austenite stability and strain-induced α-martensite on HEE. · Cost efficiency by reduced amounts of nickel and molybdenum. · Influence of silicon on HEE. - Abstract: To address the upcoming austenitic stainless steel market for automotive applications involving hydrogen technology, a novel lean - alloyed material was developed and characterized. It comprises lower contents of nickel and molybdenum compared to existing steels for high - pressure hydrogen uses, for instance 1.4435 (AISI 316L). Alloying with manganese and carbon ensures a sufficient stability of the austenite at 8 wt.% of nickel while silicon is added to improve resistance against embrittlement by dissolved hydrogen. Investigations were performed by tensile testing in air and 400 bar hydrogen at 25 deg. C, respectively. In comparison to a standard 1.4307 (AISI 304L) material, a significant improvement of ductility was found. The materials concept is presented in general and discussed with regard to austenite stability and microstructure.

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

    Directory of Open Access Journals (Sweden)

    Z. Brytan

    2010-05-01

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

  13. The mechanical stability of retained austenite in low-alloyed TRIP steel under shear loading

    International Nuclear Information System (INIS)

    The microstructure evolution during shear loading of a low-alloyed TRIP steel with different amounts of the metastable austenite phase and its equivalent DP grade has been studied by in-situ high-energy X-ray diffraction. A detailed powder diffraction analysis has been performed to probe the austenite-to-martensite transformation by characterizing simultaneously the evolution of the austenite phase fraction and its carbon concentration, the load partitioning between the austenite and the ferritic matrix and the texture evolution of the constituent phases. Our results show that for shear deformation the TRIP effect extends over a significantly wider deformation range than for simple uniaxial loading. A clear increase in average carbon content during the mechanically-induced transformation indicates that austenite grains with a low carbon concentration are least stable during shear loading. The observed texture evolution indicates that under shear loading the orientation dependence of the austenite stability is relatively weak, while it has previously been found that under tensile load the {110}〈001〉 component transforms preferentially. The mechanical stability of retained austenite in TRIP steel is found to be a complex interplay between the interstitial carbon concentration in the austenite, the grain orientation and the load partitioning

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

    Science.gov (United States)

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

    2016-07-01

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

  15. Evaluation of the austenitic alloys 304L, 316L, and alloy 825 under Tuff repository conditions

    International Nuclear Information System (INIS)

    Austenitic alloys 304L and 316L and stainless steel 825 were investigated as candidate materials for containers for waste disposal in the relatively benign conditions of the Yucca Mountain site. In this vault there will be very little water, and what there is will contain small amounts of chlorides, nitrates, sulphates and carbonates. The radiation fields will be 104 rad/h initially, but will decay to low levels by the end of the containment period. The initial temperature will be around 250 C, and it will remain above the boiling point of water for the containment period (approximately 300 years). There will be no lithostatic or hydrostatic pressure. Type 304L stainless steel is a base case material used in comparisons with other candidates. Type 316L stainless steel possesses enhanced resistance to sensitization and localized corrosion; alloy 825 is stabilized to have a much better resistance to sensitization and localized corrosion and performs better in chloride environments

  16. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-15

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

  17. Influence of manganese, carbon and nitrogen on high-temperature strength of Fe-Cr-Mn austenitic alloys

    International Nuclear Information System (INIS)

    High Mn-Cr-Fe base alloys are candidates for the first wall material of fusion reactors because of rapid decay of radioactivity of the alloys after neutron irradiation compared with that of Ni-Cr-Fe base alloys. Their high temperature properties, however, are not clearly understood at present. In this paper, a study has been made of the effects of Mn, C and N content on the high-temperature tensile strength and creep properties of a 12% CR-Fe base alloy. Mn tends to decrease tensile strength and proof stress at intermediate temperatures. At higher temperatures in the austenite range, however, tensile properties scarcely depend on Mn content. C and N additions improve the tensile properties markedly. The combined addition of 0.2%C and 0.2%N to a 12%Cr-15%Mn-Fe base alloy makes the strength at 873K as high as that of a modified type 316 stainless steel. Combined alloying with C and N also improves the creep strength. Cold working is very useful in increasing the creep strength because of the finely dispersed precipitates in the matrix during creep. From these results, Fe-12%Cr-15%Mn-15%Mn-0.2%c-0.2%N is recommended as one of the most suitable alloys in this system for high temperature usage. (author)

  18. Fracture toughness of austenitic stainless steels after BWR irradiation

    International Nuclear Information System (INIS)

    Austenitic stainless steels used for the core internal materials in BWRs are hardened by exposure to fast neutrons, and the fracture toughness is reduced by neutron fluence. It is important for integrity estimation of aged core internals to understand the reduction behavior of fracture toughness by neutron irradiation. In this study, core shroud materials (Type 304 SS) with 38 mm thickness and top guide materials (Type 316 SS) with 9.1 mm thickness, actually used for over twenty years in BWRs, were transported to PIE facility. Neutron fluence of type 304 SS was 1-6*1024 (n/m2, E > 1 MeV) and that of type 316 SS was 3-17*1024 (n/m2, E > 1 MeV). Fracture toughness tests for base metal were performed at 288 C in air using the CT specimen with real thickness of core shroud and top guide to obtain valid fracture toughness. And fracture toughness tests for heat affected zone (HAZ) were conducted using 0.7 TCT by the restriction of weld line direction. JIC and JQ of irradiated type 304 and type 316 base metals decreased with neutron fluence. JIC values of type 304 SS base metal and HAZ were obtained over 140 kJ/m2 at 5*1024 n/m2. JIC values of type 316 SS base metal were obtained over 240 kJ/m2 at 1.7*1025 n/m2. JIC values of type 304 HAZ were similar to those of base metal. JQ values of type 316 SS base metal had a higher value than JIC of type 304 SS base metal at similar neutron fluence. The difference between type 304 SS and type 316 SS is considered to be caused by the orientation effect of microstructure in CT specimen. From the SEM observation of crack surfaces, a linear relationship between JIC and the critical stretched zone width (SZWC) was found to exist in irradiated stainless steel materials. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Non-metallic inclusions in high manganese austenitic alloys

    Directory of Open Access Journals (Sweden)

    A. Grajcar

    2011-07-01

    Full Text Available Purpose: The aim of the paper is to identify the type, fraction and chemical composition of non-metallic inclusions modified by rare-earth elements in an advanced group of high-manganese austenitic C-Mn-Si-Al-type steels with Nb and Ti microadditions.Design/methodology/approach: The heats of 3 high-Mn steels of a various content of Si, Al and Ti were melted in a vacuum induction furnace and a modification of non-metallic inclusions was carried out by the mischmetal in the amount of 0.87 g or 1.74 g per 1 kg of steel. Evaluation of the metallurgical purity of steels with non-metallic inclusions was done basing on determination their fraction, type, size and morphology. Stereological parameters of the inclusions were assessed by the use of automatic image analyzer cooperating with light microscope. EDS method was used to assess the chemical composition of non-metallic inclusions.Findings: It was found that the steels are characterized by high metallurgical purity connected to low concentrations of phosphorus and gases at a slightly higher sulphur content, introduced to a melt together with electrolytic manganese. The steels contain fine sulfide inclusions with a mean size from 21 to 25 µm2 in a majority and their fraction equals from 0.047 to 0.09%, depending on sulphur content. MnS, carbonitrides of the (Ti,Nb(C,N type and complex carbosulfides containing Mn, Ti and Nb were identified in steels. The beneficial influence in decreasing a fraction of non-metallic inclusions and their susceptibility to elongate in a rolling direction has a higher addition of mischmetal and titanium microaddition. A modification of the chemical composition of non-metallic inclusions by Ce, La and Nd proceeds in an external zone of inclusions.Research limitations/implications: Further investigations relating the type and morphology of non-metallic inclusions to mechanical properties of sheets at various sections according to the rolling direction are needed

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

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

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

    2011-01-01

    Martensitic transformation in a model Fe-Ni-0.6wt%C alloy was investigated at sub-zero Celsius temperature. The influence of the thermal path in determining the conditions leading to the formation of martensite was studied. In the investigation, samples were austenitized and quenched, whereafter......-situ synchrotron X-ray diffraction by evaluating austenite and martensite Bragg reflections. Also, the state of internal strain in austenite was determined....

  10. The influence of nitrogen alloying on the pitting and crevice corrosion of austenitic and duplex stainless steels

    International Nuclear Information System (INIS)

    The effect of nitrogen alloying on the pitting corrosion resistance of duplex and austenitic stainless steels has been examined. In order to avoid alteration of the phase ratio as a result of nitrogen alloying of the duplex steels, a simultaneous decrease has been made in the nickel content. Austenitic alloys of compositions corresponding to the austenite phase of the duplex steels have been investigated and compared to the behaviour of austenitic steels in which the nitrogen content or the nickel content alone has been varied. Nitrogen has a beneficial effect on pitting and crevice corrosion resistance in all cases but the duplex stainless steel exhibit a lower resistance to pitting and a higher resistance to crevice corrosion than predicted from the austenite nitrogen content. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    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.

  13. Advances in development of refractory austenitic steels and nickel alloys for power engineering

    International Nuclear Information System (INIS)

    An evaluation is presented of the current state of knowledge of the properties and technologies of refractory austenitic steels and Ni alloys, this mainly of materials used in the temperature range of 600 to 1100 degC where the main causes of damage are creep, fatigue and high temperature corrosion. Attention is mainly devoted to the results of applied research. The problems of concrete applications in nuclear engineering were dealt with in the paper ''Assessment of long-term refractory properties of selected types of austenitic steels''. (J.B.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-22

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

  15. Plastic localization phenomena in a Mn-alloyed austenitic steel

    OpenAIRE

    Matteis, Paolo; Firrao, Donato; Scavino, Giorgio; Russo Spena, Pasquale

    2010-01-01

    A 0.5 wt pct C, 22 wt pct Mn austenitic steel, recently proposed for fabricating automotive body structures by cold sheet forming, exhibits plastic localizations (PLs) during uniaxial tensile tests, yet showing a favorable overall strength and ductility. No localization happens during biaxial Erichsen cupping tests. Full-thickness tensile and Erichsen specimens, cut from as-produced steel sheets, were polished and tested at different strain rates. During the tensile tests, the PL phenomena co...

  16. Stress-corrosion and fatigue cracking behaviour of nitrogen-alloyed austenitic and ferritic-austenitic chrome-nickel-(molybdenum)steels

    International Nuclear Information System (INIS)

    Under unfavorable heat-exchanger conditions simulated with 3 % sodium chloride solutions of different rhoH-values and redox potentials there excists a close connection between the stress-corrosion and fatigue behavior and the results of electrochemical measurements for nitrogen-alloyed austenitic and ferritic-austenitic high-alloy stells. Elevated contents of chromium and molybdenum have a positive effect. With free corrosion the materials no. 1.4311 and 1.4406 as well as partly also 1.4439 and the corresponding weldings are not suited if there is a hazard of stress corrosion. For 1.4439 and 1.4462 and their weldings of the same type a lower-bound stress may be given for the hazard of stress corrosion. It is within the order of magnitude of the garanteed elevated temperature yield strength and is therefore distinctly higher for the ferritic-austenitic steel 1.4462 than for nitrogenous austenitic steels. For the nitrogenous austenitic steels mentioned the dynamic loading capacity with and without the action of corrosive media is marked by lower than for the steel 1.4462 with ferritic-austenitic structure and its welding of the same type. As the notch sensitivity for both groups of material can be taken as equal the steel 1.4462 appears suitable for being used under unfavorable heat-exchanger conditions if high resistance against stress-corrosion and fatigue cracking is taken into account. (orig.)

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

    Science.gov (United States)

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

    2016-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Stress corrosion cracking and oxidation of austenitic stainless steel 316 L and model alloy in supercritical water reactor

    International Nuclear Information System (INIS)

    In this work, an austenitic stainless steel type 316 L was tested in deaerated supercritical water at 400 deg. C and 500 deg. C and 25 MPa to determine how variations in water conditions influence its stress corrosion cracking behaviour and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Moreover, the influence of plastic deformation in the resistance of the material to SCC was also studied at both temperatures. In addition to this, previous oxidation experiments at 400 deg. C and 500 deg. C and at 25 MPa were taken into account to gain some insight in this kind of processes. Furthermore, a cold worked model alloy based on the stainless steel 316 L with some variations in the chemical composition in order to simulate the composition of the grain boundary after irradiation was tested at 400 deg. C and 25 MPa in deaerated supercritical water. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-25

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2004-08-01

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

  5. Effects of alloy and solution chemistry on the fracture of passive films on austenitic stainless steel

    International Nuclear Information System (INIS)

    The Taguchi analysis method was used to simultaneously study the effects of alloy chemistry, pH, and halide ion concentrations on the fracture of electrochemically grown passive films using a nanoindentation technique. Three austenitic stainless steels, 304L, 316L, and 904L were potentiostatically polarized in hydrochloric acid solutions. The fracture load was dominated primarily by alloy chemistry. Passive films mechanically weaken as the atomic iron concentration increases in the film. Prolonged anodic ageing time increases the fracture load of passive films

  6. Effects of alloy and solution chemistry on the fracture of passive films on austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Alamr, A. [School of Mechanical and Materials Engineering, Washington State University, P.O. Box 642920, Pullman, WA 99164-2920 (United States)]. E-mail: alamrz@wsu.edu; Bahr, D.F. [School of Mechanical and Materials Engineering, Washington State University, P.O. Box 642920, Pullman, WA 99164-2920 (United States)]. E-mail: bahr@mail.wsu.edu; Jacroux, Michael [Department of Statistics, Washington State University, Pullman, WA 99164-3144 (United States) ]. E-mail: jacroux@wsu.edu

    2006-04-15

    The Taguchi analysis method was used to simultaneously study the effects of alloy chemistry, pH, and halide ion concentrations on the fracture of electrochemically grown passive films using a nanoindentation technique. Three austenitic stainless steels, 304L, 316L, and 904L were potentiostatically polarized in hydrochloric acid solutions. The fracture load was dominated primarily by alloy chemistry. Passive films mechanically weaken as the atomic iron concentration increases in the film. Prolonged anodic ageing time increases the fracture load of passive films.

  7. Features of structure-phase changes in high-manganese type EP-838 austenitic stainless steel irradiated with heavy ions and neutrons

    International Nuclear Information System (INIS)

    Production of ecologically acceptable fusion energy is closely connected with development of low-activated materials. Microstructure and irradiation behaviour for Fe-Cr-Mn alloys, a lower activation class of alloys relative to commercial Fe-Cr-Ni-Mo steels, have not been sufficiently investigated, especially phase stability and void swelling characteristics. This paper presents results of an investigation on the microstructural development in austenitic high-managenese alloy type EP-838 (solution-annealed (SA) and 30% cold-worked (CW)) during ion (Cr2+, E=3 MeV) and fast-neutron irradiation, as a function of dose and temperature. It was found that the swelling rate of alloy type EP-838 was lower than that of steel type Cr16Ni15Mo3Nb, but was still high after a dose of 100 dpa in the temperature region 550-650 C. The influence of cold work on swelling is weak. Differences in defect structure in these steels are connected with the difference in behaviour of point defects in nickel and manganese austenite. (orig.)

  8. The irradiation effects on zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Negut, Gh. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania)]. E-mail: joenegut@yahoo.com; Ancuta, M. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania); Radu, V. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania); Ionescu, S. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania); Stefan, V. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania); Uta, O. [Institute for Nuclear Research Pitesti, Campului Str., 1, Mioveni P.O. Box 78, 0300 Pitesti (Romania); Prisecaru, I. [University Politehnica Bucharest, Bucharest (Romania); Danila, N. [University Politehnica Bucharest, Bucharest (Romania)

    2007-05-31

    Pressure tube samples were irradiated under helium atmosphere in the TRIGA Steady State Research and Material Test Reactor of the Romanian Institute for Nuclear Research (INR). These samples are made of the Zr-2.5%Nb alloy used as structural material for the CANDU Romanian power reactors. After irradiation, mechanical tests were performed in the Post Irradiation Examination Laboratory (PIEL) to study the influence of irradiation on zirconium alloys mechanical behaviour. The tensile test results were used for structural integrity assessment. Results of the tests are presented. The paper presents, also, pressure tube structural integrity assessment.

  9. The irradiation effects on zirconium alloys

    International Nuclear Information System (INIS)

    Pressure tube samples were irradiated under helium atmosphere in the TRIGA Steady State Research and Material Test Reactor of the Romanian Institute for Nuclear Research (INR). These samples are made of the Zr-2.5%Nb alloy used as structural material for the CANDU Romanian power reactors. After irradiation, mechanical tests were performed in the Post Irradiation Examination Laboratory (PIEL) to study the influence of irradiation on zirconium alloys mechanical behaviour. The tensile test results were used for structural integrity assessment. Results of the tests are presented. The paper presents, also, pressure tube structural integrity assessment

  10. Hydrogen embrittlement for austenitic alloys: behaviour of microstructure and segregation of sulphur and phosphorus impurities

    International Nuclear Information System (INIS)

    The mechanical properties of austenitic alloys can be highly modified in the presence of a hydrogenation. The purpose of this study is to specify this influence on two alloys (800 and 600) for which the chemical composition on the one hand, and the microstructure on the other hand have significant differences. The hydrogenation was done before tensile testing under potentiostatic cathodic polarization at 300 deg c for times varying between 5 and 48 h. A high embrittlement from the hydrogen was shown in the case of the alloy 600 in the quenched annealed state with precipitates at the grain boundaries. It is lower in the quenched state without precipitates at the grain boundaries. On the other hand, the alloy 800 is not embrittled by the hydrogen, neither in the quenched state nor in the annealed state, even in the presence of precipitates. The influence of a phosphorus segregation on the grain boundaries can explain the differences observed. (authors)

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

    Science.gov (United States)

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

    2016-07-01

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

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

    Science.gov (United States)

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

    2014-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

  15. Atom probe tomography of the austenite-ferrite interphase boundary composition in a model alloy Fe-C-Mn

    Energy Technology Data Exchange (ETDEWEB)

    Thuillier, O. [Groupe de Physique des Materiaux, UMR CNRS 6634, Institut des Materiaux de Rouen, Universite de Rouen, 76 801 Saint Etienne du Rouvray Cedex (France)]. E-mail: olivier.thuillier@etu.univ-rouen.fr; Danoix, F. [Groupe de Physique des Materiaux, UMR CNRS 6634, Institut des Materiaux de Rouen, Universite de Rouen, 76 801 Saint Etienne du Rouvray Cedex (France); Goune, M. [Arcelor Research, Voie Romaine B.P. 320, 57214 Maizieres-Les-Metz (France); Blavette, D. [Groupe de Physique des Materiaux, UMR CNRS 6634, Institut des Materiaux de Rouen, Universite de Rouen, 76 801 Saint Etienne du Rouvray Cedex (France)

    2006-12-15

    A tomographic atom p analysis has been developed to study the interfacial conditions during isothermal austenite transformation to ferrite at 700 deg. C in an Fe-C-Mn model alloy. The interfacial conditions lead to different alloying element profiles across the interface, and a comparison is made between this experimental result and the DICTRA software predictions under the various conditions.

  16. Atom probe tomography of the austenite-ferrite interphase boundary composition in a model alloy Fe-C-Mn

    International Nuclear Information System (INIS)

    A tomographic atom probe analysis has been developed to study the interfacial conditions during isothermal austenite transformation to ferrite at 700 deg. C in an Fe-C-Mn model alloy. The interfacial conditions lead to different alloying element profiles across the interface, and a comparison is made between this experimental result and the DICTRA software predictions under the various conditions

  17. Carburization of austenitic alloys by gaseous impurities in helium

    International Nuclear Information System (INIS)

    The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H2, CO, CH4, H2O and CO2 was studied. Corrosion tests were conducted in a temperature range from 649 to 10000C (1200 to 18320F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 μatm H2, 450 μatm CO, 50 μatm CH4 and 50 μatm H2O for Environment A; 200 μatm H2, 100 μatm CO, 20 μatm CH4, 50 μatm H2O and 5 μatm CO2 for Environment B; 500 μatm H2, 50 μatm CO, 50 μatm CH4 and 2O for Environment C; and 500 μatm H2, 50 μatm CO, 50 μatm CH4 and 1.5 μatm H2O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys

  18. Precipitate evolution in low-nickel austenitic stainless steels during neutron irradiation at very low dose rates

    Energy Technology Data Exchange (ETDEWEB)

    Isobe, Y.; Sagisaka, M. [Nuclear Fuel Industries, Osaka (Japan); Garner, F. [Pacific Northwest National Laboratory, Richland WA, AK (United States); Okita, T. [Tokyo Univ. (Japan)

    2007-07-01

    Full text of publication follows: Not all components of a fusion reactor will be subjected to high atomic displacement rates. Some components outside the plasma containment may experience relatively low displacement rates but data generated under long-term irradiation at low dpa rates is hard to obtain. In another study the neutron-induced microstructural evolution in response to long term irradiation at very low dose rates was studied for a Russian low-nickel austenitic stainless steel that is analogous to AISI 304. The irradiated samples were obtained from an out-of-core anti-crush support column for the BN-600 fast reactor with doses ranging from 1.5 to 22 dpa generated at 3x10{sup -9} to 4x10{sup -8} dpa/s. The irradiation temperatures were in a very narrow range of 370-375 deg. C. Microstructural observation showed that in addition to voids and dislocations, an unexpectedly high density of small carbide precipitates was formed that are not usually observed at higher dpa rates in this temperature range. These results required us to ask if such unexpected precipitation was anomalous or was a general feature of low-flux, long-term irradiation. It is shown in this paper that a similar behavior was observed in a western stainless steel, namely AISI 304 stainless steel, irradiated at similar temperatures and dpa rates in the EBR-II fast reactor, indicating that irradiation at low dpa rates for many years leads to a different precipitate microstructure and therefore different associated changes in matrix composition than are generated at higher dpa rates. One consequence of this precipitation is a reduced lattice parameter of the alloy matrix, leading to densification that increases in strength with increasing temperature and dose. A. non-destructive method to evaluate these precipitates is under development and is also discussed in this paper. (authors)

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

    Institute of Scientific and Technical Information of China (English)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2004-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-12-01

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

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

    Science.gov (United States)

    Danilchenko, Vitalij

    2016-12-01

    The X-ray and magnetometry methods were used to study α-γ transformation mechanisms on heating quenched Fe-22.7 wt.% Ni-0.58 wt.% С alloy. Variation of heating rate within 0.03-80 K/min allowed one to switch from diffusive to non-diffusive mechanism of the α-γ transformation. Heating up primary austenitic single crystal specimen at a rate of less than 1.0-0.5 K/min has led to formation of aggregate of grains with different orientation and chemical composition in the reverted austenite. Significant fraction of these grains was determined to have sizes within nanoscale range. PMID:26860715

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

    International Nuclear Information System (INIS)

    Highlights: • We model the precipitation kinetics in irradiated 316 austenitic stainless steels. • Radiation-induced phases are predicted to form at over 10 dpa segregation conditions. • The Si content is the most critical for the formation of radiation-induced phases. - Abstract: 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  5. Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F. [Pacific Northwest National Laboratory, P.o. Box 999, Richland WA, AK 99352 (United States)

    2007-07-01

    Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to {approx} 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

  6. On the recovery of neutron irradiation defects of some metals and alloys

    International Nuclear Information System (INIS)

    This work deals with the recovery of mechanical properties of neutron irradiated material to the pre-irradiating values. Rate of migration of defects responsible for radiation hardening and those inducing radiation embrittlement is analyzed. Role of crystalline structure is also studied. Materials of FCC crystal structure used in these investigations are pure Cu, Cu-5 at. % , Al, Cu-5 at. % Si, some Ni base binary alloys and some austenitic stainless steels mainly of AISI types 304 and 316. Among materials of BCC crystalline structure Fe-6 wt % Cr alloy is used. Alloys with CPH structure used in the present investigations are Zr-l wt. % Nb and Mg - 4.8 wt % Li alloys. History of material is studied such as cold worked state and annealed condition. Character of alloying elements and their amounts were of interest in this study. The result showed that the higher the percentage radiation hardening, the slower is the migration of radiation defects. Irradiated pure metals recovered at a higher temperature than alloys. Cold work accelerated the migration of radiation defects. The amount of alloying elements had little effect on the recovery temperatures. Character of solute alloying elements (substitutional or interstitial) revealed sensitive effect on the migration of radiation defects. Rate of migration of defects causing hardening can be different from those causing embrittlement. (author)

  7. Simulation of austenite formation kinetics during fast heating of iron and its alloys

    International Nuclear Information System (INIS)

    The austenitization of iron-based alloys was studied by working out a phase transformation model that makes possible the calculation of the kinetic curves of the α → γ transformation at various heating rates. The modelling algorithm was based on the discrete description of the transformation process. The effect of the nucleation parameters upon the kinetics of the isothermal transformation was studied. It was found that at a constant nucleation probability, the appearance of a predetermined proportion of pre-formed transformation nuclei has little effect upon the rate of transformation and the position of the isothermal plateau. And conversely, the increase in the nucleation probability lowers substantially the initial transformation point

  8. The relationship of irradiation creep, the local stress state and void swelling in PWR austenitic internals

    International Nuclear Information System (INIS)

    'Full text:' Swelling-induced distortion of PWR austenitic internals has been raised as a potential issue for plant life extension. Although void swelling generates volumetric strains, it is irradiation creep that responds to the local stress state to distribute the linear strains. The situation is complicated in that swelling gradients or constraints against swelling generate stresses that drive irradiation creep which attempts to lower the stresses, but swelling rates are accelerated by the stress. Finally, the creep rate accelerates in proportion to the swelling rate, leading to a complex relationship between irradiation creep, swelling, swelling gradients arising from gradients in temperature and dpa rate, and the distribution of internal stresses. An analysis of swelling and creep data for annealed 304 and cold-worked 316 is presented to show that irradiation creep is accelerated by swelling, and that swelling is accelerated by stress. The relationship between swelling and creep is independent of the steel identity and thermo-mechanical condition. (author)

  9. Irradiation creep of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Science.gov (United States)

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

    2013-06-01

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

  11. Growth of creep life of type-347H austenitic stainless steel by micro-alloying elements

    International Nuclear Information System (INIS)

    Research highlights: → B, Ce and N can improve the creep life significantly at high temperature. → The precipitate of B element at the grain boundaries can improve the creep life. → The removing O through Ce provided the steel with longer creep life. → N increased the creep life by stabilizing austenite and solid solution strengthening. - Abstract: The creep life of type-347H austenitic stainless steel modified with B, Ce and N was measured, and microstructures were analyzed by optical microscope, X-ray diffraction, scanning electron microscope and transmission electron microscope equipped with energy dispersive spectroscopy. The results indicate that B, Ce and N can improve the creep life significantly at high temperature. The growth of creep life was mainly due to the precipitate of B in the elemental form at the grain boundaries and the removing O through Ce. N addition made for solid solution strengthening and effectively suppressed the precipitate of δ-ferrite at high temperature. The micro-alloying elements have a beneficial effect on creep life of type-347H austenitic stainless steel at high temperature.

  12. Study of interactions between liquid lead-lithium alloy and austenitic and martensitic steels

    International Nuclear Information System (INIS)

    In the framework of Fusion Technology, the behaviour of structural materials in presence of liquid alloy Pb17Li is investigated. First, the diffusion coefficients of Fe and Cr have been determined at 500 deg C. Then mass transfer experiments in Pb17Li have been conducted in an anisothermal container with pure metals (Fe, Cr, Ni), Fe-Cr steels and austenitic steels. These experiments showed a very high loss of Nickel, which is an accordance with its high solubility, and Cr showed mass-losses one order of magnitude higher than for pure iron, as the diffusion coefficient of Cr is three orders of magnitude higher than for pure Fe. The corrosion rate of binary Fe-Cr and pure Fe are identical. In austenitic steels, the gamma lattice allows a higher mass-transfer of Cr than the alpha lattice, the presence of Cr slows downs the dissolution of Ni, and the porosity of corrosion layers results of losses of Cr and Ni. Finally, a review of our results and those of other laboratories allowed an identification of the corrosion limiting step. In the case of 1.4914 martensitic steel it is the diffusion of Fe in Pb17Li, while in the case of 316L austenitic steel it is the diffusion of Cr in Pb17Li

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

    Science.gov (United States)

    Chen, Hao; Zhu, Benqiang; Militzer, Matthias

    2016-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Kim, Jin Weon; Byun, Thak Sang

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2012-01-01

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

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

    Science.gov (United States)

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

    2015-11-01

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

  19. High corrosion resistance of austenitic stainless steel alloyed with nitrogen in an acid solution

    Energy Technology Data Exchange (ETDEWEB)

    Metikos-Hukovic, M., E-mail: mmetik@fkit.h [Department of Electrochemistry, Faculty of Chemical Engineering and Technology, University of Zagreb, Savska 16, P.O. Box 177, 100000 Zagreb (Croatia); Babic, R. [Department of Electrochemistry, Faculty of Chemical Engineering and Technology, University of Zagreb, Savska 16, P.O. Box 177, 100000 Zagreb (Croatia); Grubac, Z. [Department of General and Inorganic Chemistry, Faculty of Chemistry and Technology, University of Split, 21000 Split (Croatia); Petrovic, Z. [Department of Electrochemistry, Faculty of Chemical Engineering and Technology, University of Zagreb, Savska 16, P.O. Box 177, 100000 Zagreb (Croatia); Lajci, N. [Faculty of Mine and Metallurgy, University of Prishtina, 10000 Prishtina, Kosovo (Country Unknown)

    2011-06-15

    Highlights: {yields} ASS alloyed with nitrogen treated at 1150 {sup o}C exhibits microstructure homogeneity. {yields} Passivation peak of ASS corresponds to oxidation of metal and absorbed hydrogen. {yields} Transfer phenomena and conductivity depend on the film formation potential. {yields} Electronic structure of the passive film and its corrosion resistance correlate well. {yields} Passive film on ASS with nitrogen is low disordered and high corrosion resistant. - Abstract: Passivity of austenitic stainless steel containing nitrogen (ASS N25) was investigated in comparison with AISI 316L in deareated acid solution, pH 0.4. A peculiar nature of the passivation peak in a potentiodynamic curve and the kinetic parameters of formation and growth of the oxide film have been discussed. The electronic-semiconducting properties of the passive films have been correlated with their corrosion resistance. Alloying austenitic stainless steel with nitrogen increases its microstructure homogeneity and decreases the concentration of charge carriers, which beneficially affects the protecting and electronic properties of the passive oxide film.

  20. High corrosion resistance of austenitic stainless steel alloyed with nitrogen in an acid solution

    International Nuclear Information System (INIS)

    Highlights: → ASS alloyed with nitrogen treated at 1150 oC exhibits microstructure homogeneity. → Passivation peak of ASS corresponds to oxidation of metal and absorbed hydrogen. → Transfer phenomena and conductivity depend on the film formation potential. → Electronic structure of the passive film and its corrosion resistance correlate well. → Passive film on ASS with nitrogen is low disordered and high corrosion resistant. - Abstract: Passivity of austenitic stainless steel containing nitrogen (ASS N25) was investigated in comparison with AISI 316L in deareated acid solution, pH 0.4. A peculiar nature of the passivation peak in a potentiodynamic curve and the kinetic parameters of formation and growth of the oxide film have been discussed. The electronic-semiconducting properties of the passive films have been correlated with their corrosion resistance. Alloying austenitic stainless steel with nitrogen increases its microstructure homogeneity and decreases the concentration of charge carriers, which beneficially affects the protecting and electronic properties of the passive oxide film.

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

    Science.gov (United States)

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

    2016-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

  3. Effect of heat treatment on the structure and creep resistance of austenitic Fe–Ni alloy

    Directory of Open Access Journals (Sweden)

    K.J. Ducki

    2011-01-01

    Full Text Available Purpose: The paper addresses the problem of determining the dependence between the initial heat treatment of an austenitic Fe–Ni alloy and its structure, and its creep resistance. Specimens of Fe–Ni alloy were subjected to tests after two variants of heat treatment, i.e. solution heat treatment followed by typical single-stage ageing, and solution heat treatment followed by novel two-stage ageing.Design/methodology/approach: For the investigated Fe–Ni alloy after solution heat treatment in the conditions: 980°C/2h/water, two variants of specimen ageing were applied for a comparison: single-stage ageing (715°C/16h/air and two-stage ageing (720°C/8h + cooling in the furnace up to the temperature of 650°C + 650°C/8h/air. The thermally treated specimens were then subjected to a static tensile test at room and elevated temperatures, and to a creep test in a temperature range of 650-750°C, at stresses from 70 to 340 MPa.Findings: It was found that both, at the room and elevated temperatures, the specimens of Fe–Ni alloy after 2-stage ageing were distinguished by higher strength properties (Y.S, T.S with a little lower plastic properties (EL., R.A. As regards extrapolated results of creep tests, it was found that at a longer exposure time of ca. 10.000 h, specimens after single-stage ageing were characterized with higher creep resistance. Lower creep resistance of the Fe–Ni alloy after two-stage ageing can be explained by increased brittleness of the material in boundary areas.Practical implications: The obtained test results may be used to optimise heat treatment and forecast the operation conditions of products made out of Fe–Ni alloy at an elevated temperature.Originality/value: The study shows a significant effect of the applied ageing variants on mechanical properties and creep resistance of the tested austenitic Fe–Ni alloy.

  4. Role of alloyed molybdenum in austenitic stainless steels in the inhibition of pitting in neutral halide solutions

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, K.; Sawada, Y.

    1976-09-01

    In the passive region of austenitic stainless steels alloyed with Mo, the formation of MoO/sub 4//sup 2 -/ ions can be expected in neutral halide solutions by the transpassive dissolution of Mo. It has been shown that MoO/sub 4//sup 2 -/ ions added to neutral NaCl solutions act as an effective inhibitor against pitting of austenitic stainless steels with and without Mo. The interaction between alloyed Mo in the steels and added MoO/sub 4//sup 2 -/ ions in the solutions is appreciable. It is likely that the inhibition of pit growth by the adsorption of MoO/sub 4//sup 2 -/ ions which are thought to result from the dissolution of the steels at the initial stage of pitting leads to increased pitting resistance of austenitic stainless steels containing Mo.

  5. Effect of triple ion beam irradiation on mechanical properties of high chromium austenitic stainless steel

    International Nuclear Information System (INIS)

    A high-chromium austenitic stainless steel has been developed for an advanced fuel cladding tube considering waterside corrosion and irradiation embrittlement. The candidate material was irradiated in triple ion (Ni, He, H) beam modes at 573 K up to 50 dpa to simulate irradiation damage by neutron and transmutation product. The change in hardness of the very shallow surface layer of the irradiated specimen was estimated from the slope of load/depth-depth curve which is in direct proportion to the apparent hardness of the specimen. Besides, the Swift's power low constitutive equation (σ=A(ε0 + ε)n, A: strength coefficient, ε0: equivalent strain by cold rolling, n: strain hardening exponent) of the damaged parts was derived from the indentation test combined with an inverse analysis using a finite element method (FEM). For comparison, Type304 stainless steel was investigated as well. Though both Type304SS and candidate material were also hardened by ion irradiation, the increase in apparent hardness of the candidate material was smaller than that of Type304SS. The yield stress and uniform elongation were estimated from the calculated constitutive equation by FEM inverse analysis. The irradiation hardening of the candidate material by irradiation can be expected to be lower than that of Type304SS. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

    S. Buyukakkas; H. Aktas; S. Akturk

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

  10. The low temperature magnetic properties of austenitic Fe-Cr-Ni alloys

    International Nuclear Information System (INIS)

    The compositional dependence of the Neel temperature has been studied, from data derived by different techniques and by various authors, for 30 austenitic stainless steels or special Fe-Cr-Ni alloys whose compositions fall near the range of the AISI 300 series. A linear relationship enables the predicted Neel temperature, Tsub(N) to be evaluated with an rms deviation of 3.5 K on the basis of the wt% of alloy constituents. The effect of alloying elements in lowering the calculated value, T*sub(N), increases in the order Ci, Ni, Mo, and Si, while Mn is unique in raising T*sub(N). By comparing this equation for T*sub(N) with previous equations to predict the onset of a martensitic phase change at a temperature Msub(s), it is concluded that isotherms for Msub(s) in ternary Fe-Cr-Ni alloys should also be parallel to the tangent to the boundary between fcc and bcc phases calculated from thermodynamic data. This conclusion is discussed with reference to results obtained by other workers. The significance of the results is discussed in terms of the application of stainless steel in magnetic environments at low temperatures. (author)

  11. Modeling of cavity swelling-induced embrittlement in irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    During long-time neutron irradiation occurred in Pressurized Water Reactors (PWRs), significant changes of the mechanical behavior of materials used in reactor core internals (made of 300 series austenitic stainless steels) are observed, including irradiation induced hardening and softening, loss of ductility and toughness. So far, much effect has been made to identify radiation effects on material microstructure evolution (dislocations, Frank loops, cavities, segregation, etc.). The irradiation-induced cavity swelling, considered as a potential factor limiting the reactor lifetime, could change the mechanical properties of materials (plasticity, toughness, etc.), even lead to a structure distortion because of the dimensional modifications between different components. The principal aim of the present PhD work is to study qualitatively the influence of cavity swelling on the mechanical behaviors of irradiated materials. A micromechanical constitutive model based on dislocation and irradiation defect (Frank loops) density evolution has been developed and implemented into ZeBuLoN and Cast3M finite element codes to adapt the large deformation framework. 3D FE analysis is performed to compute the mechanical properties of a polycrystalline aggregate. Furthermore, homogenization technique is applied to develop a Gurson-type model. Unit cell simulations are used to study the mechanical behavior of porous single crystals, by accounting for various effects of stress triaxiality, of void volume fraction and of crystallographic orientation, in order to study void effect on the irradiated material plasticity and roughness at polycrystalline scale. (author)

  12. High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

    International Nuclear Information System (INIS)

    The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties

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

    Energy Technology Data Exchange (ETDEWEB)

    Huang, H.F.; Li, J.J.; Li, D.H.; Liu, R.D.; Lei, G.H.; Huang, Q.; Yan, L., E-mail: yanlong@sinap.ac.cn

    2014-11-15

    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 Xe{sup 26+} ions. TEM results reveal the presence of dislocation loops with a number density of approximately 3 × 10{sup 22} m{sup −3} and sizes between 3 to 10 nm due to the collapse of vacancy rich cores inside displacement cascades. Peak broadening observed in XRD diffraction patters reveal systematic changes to lattice parameters due to irradiation. The calculated indentation values in irradiated 316 SS were found to be much higher in comparison to the unirradiated specimen, indicating the dose dependent effect of irradiation on hardness. The relationship between irradiation induced microstructural evolution and the changes to the mechanical properties of CW 316 SS are discussed in the context of fluence and irradiation temperature.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  15. Bend-fatigue properties of JPCA and Alloy800H specimens irradiated in a spallation environment

    International Nuclear Information System (INIS)

    To evaluate the lifetime of the beam window of an accelerator-driven transmutation system (ADS) and spallation neutron source, post irradiation examination (PIE) of the STIP (SINQ target irradiation program, SINQ; Swiss spallation neutron source) specimens has been carried out. The specimens tested in this study were made from the austenitic steel JPCA (Japan Primary Candidate Alloy) and high-Ni steel Alloy800H. The specimens were irradiated at SINQ Target 4 (STIP-II) with high-energy protons and spallation neutrons. The irradiation conditions were as follows: the proton energy was 580 MeV, irradiation temperatures ranged from 120 to 350 °C and displacement damage levels ranged from 7.0 to 19.3 dpa. Bend-fatigue tests were performed in air at room temperature under deflection control mode. The wave form of the control signal was a sine curve with a frequency of 26 Hz. Fracture surface observation after the tests was done by SEM. The results on the irradiated JPCA in this study are identical to with the result of STIP-I specimens (−11 dpa). Namely, the numbers of cycles to failure (Nf) were not changed by irradiation. Dpa dependence of Nf was not clearly seen in the irradiation conditions. In spallation environment, He atom production ratio is very high and most of He atoms are retained in the materials. In spite of large number of He atoms, all JPCA specimens show transgranular fracture surface

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  19. Preirradiation microstructrual development designed to minimize properties degradation during irradiation in austentic alloys

    International Nuclear Information System (INIS)

    The first-generation Prime Candidate Alloy (PCA) for the austenitic stainless steel class of alloys for application as a Magnet Fusion Energy (MFE) first-wall material is a 14 Cr-16 Ni-0.25 Ti modification of Type 316 stainless steel. A key parameter for material performance is wall lifetime. The ability of the material to resist swelling and resist embrittlement during irradiation is important to longer wall lifetimes. The microstructure that evolves during irradiation is primarily responsible for both the swelling and embrittlement responses, and helium plays a central role in this microstructural evolution. This paper indicated how preirradiation microstructures that employ control of MC precipitation and dislocation density are designed and produced for fusion application of PCA

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

    Directory of Open Access Journals (Sweden)

    Serna, A.

    2003-12-01

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

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  3. Effect of Grain Size on Void Formation during High-Energy Electron Irradiation of Austenitic Stainless Steel

    DEFF Research Database (Denmark)

    Singh, Bachu Narain

    1974-01-01

    Thin foils of an ‘ experimental ’ austenitic stainless steel, with and without dispersions of aluminium oxide particles, are irradiated with 1 MeV electrons in a High Voltage Electron Microscope at 600°C. Evidence of grain size dependent void nucleation, void concentration, and void volume swelling...

  4. Examination of Austenitic Steels Microstructure Change Induced Neutron Irradiation by Using of Neutron Scattering Methods

    International Nuclear Information System (INIS)

    Two cladding pin austenitic steels ChS68 and EK164 irradiated in the BN-600 reactor up to maximum dose ∼80 dpa were examined by methods: neutron diffraction; small angle neutron scattering; X ray and electron microscopy. The formation and growth of both dislocation loops and radiation voids, as well as formation of segregation profiles, second phase precipitates, and chemical composition changes in a crystalline matrix were revealed in the steels after irradiation. All these resulted in growth of micro deformations in the crystalline matrix and changes of texture of the cladding pin steels, which were fixed by neutron diffraction and X ray methods. Texture decreasing more clearly occurred after low irradiation temperature (370°C) then after higher temperature irradiation (520°C). Observing of small angle neutron scattering increasing with dose growth is probably caused in He-v voids nucleus and segregations formed along dislocations and grain boundaries. At the present time the nature of scattering centers were not uniquely established. It is connected with an insufficiently adequate calculation model of small angle neutron scattering. Therefore solving of this problem requires additional development of calculation models. On the whole the examination shows that neutron scattering methods are very effective for research radiation induced structure changes. (author)

  5. Effect of prior austenite carbon partitioning on martensite hardening variation in a low alloy ferrite–martensite dual phase steel

    Energy Technology Data Exchange (ETDEWEB)

    Ghasemi Banadkouki, S.S.; Fereiduni, E., E-mail: e.fereiduni@yahoo.com

    2014-12-01

    The aim of this research work is to investigate in detail the carbon partitioning within prior austenite developed during austenite to ferrite phase transformation, and consequently its relation to the martensite hardening variation in a low alloy ferrite–martensite dual phase (DP) steel. For this purpose, a wide variety of ferrite–martensite DP samples with different volume fractions of ferrite and martensite have been developed using step quenching heat treatment processes at 600 °C for various holding times after being austenitized at 860 °C for 60 min. Both spot and line scan energy dispersive X-ray spectroscopy for carbon analyses have been used in conjunction with nanoindentation tests to follow the variation of carbon partitioning within prior austenite areas and consequently the associated martensite hardening response in the DP specimens. Experimental results showed that the martensite hardening behavior was quite variable in the ferrite–martensite DP samples and even within a specific martensite area within a specific DP microstructure. A higher level and also a more scattered nanohardness were observed for martensite in the DP samples treated at 600 °C for longer holding times. These results were rationalized due to the variation of carbon partitioning within the prior austenite area developed during isothermal holding in the ferrite–austenite DP region. Longer isothermal holding times were associated with more carbon redistribution within prior austenite as a consequence of more ferrite formation, which resulted in the formation of harder martensite with a more scattered hardening response. Furthermore, compared to the central locations of martensite area, those nearer to the ferrite–martensite interfaces contained higher carbon concentration and consequently higher hardening responses.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-04-01

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

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

    Science.gov (United States)

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

    2016-06-01

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

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

    Science.gov (United States)

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

    2016-04-01

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

  9. Nanomechanical insights into the deformation behavior of austenitic alloys with different stacking fault energies and austenitic stability

    Energy Technology Data Exchange (ETDEWEB)

    Misra, R.D.K., E-mail: dmisra@louisiana.edu [Center for Structural and Functional Materials, University of Louisiana at Lafayette, P.O. Box 44130, Lafayette, LA 70504 (United States); Zhang, Z.; Jia, Z.; Surya, P.K.C. Venkat [Center for Structural and Functional Materials, University of Louisiana at Lafayette, P.O. Box 44130, Lafayette, LA 70504 (United States); Somani, M.C.; Karjalainen, L.P. [Department of Mechanical Engineering, University of Oulu, P.O. Box 4200, 90014 Oulu (Finland)

    2011-08-25

    Highlights: {yields} Deformation mechanisms of Type 316L, 301LN, and TWIP steels were elucidated. {yields} Nanoindentation and electron microscopy was used to explain deformation behavior. {yields} Multiple pop-ins depend on the stability and stacking fault energy of the steels. {yields} Strain-induced martensite formation and twinning involve variant selection. - Abstract: Nanoscale experiments and electron microscopy were combined to probe the deformation behavior in near defect-free volume of three austenitic steels (Type 316L, 301LN, and TWIP steel) with different stacking fault energies and austenite stability. In all the three steels, the occurrence of first pop-in is related to nucleation of dislocations in the small defect-free volume. But the second and subsequent pop-ins describe the load-displacement response resulting from the multiplication, motion and pile-up of dislocations and twinning in stable 316L stainless steel, phase transition such as strain-induced austenite-to-martensite phase transformation in metastable 301LN steel, and twinning in TWIP steel. Pop-ins associated with deformation twinning occur at a lower displacement in TWIP steel as compared to 316L steel, consistent with the lower stacking fault energy of TWIP steel. Both strain-induced martensite formation and twinning involve variant selection.

  10. Nanomechanical insights into the deformation behavior of austenitic alloys with different stacking fault energies and austenitic stability

    International Nuclear Information System (INIS)

    Highlights: → Deformation mechanisms of Type 316L, 301LN, and TWIP steels were elucidated. → Nanoindentation and electron microscopy was used to explain deformation behavior. → Multiple pop-ins depend on the stability and stacking fault energy of the steels. → Strain-induced martensite formation and twinning involve variant selection. - Abstract: Nanoscale experiments and electron microscopy were combined to probe the deformation behavior in near defect-free volume of three austenitic steels (Type 316L, 301LN, and TWIP steel) with different stacking fault energies and austenite stability. In all the three steels, the occurrence of first pop-in is related to nucleation of dislocations in the small defect-free volume. But the second and subsequent pop-ins describe the load-displacement response resulting from the multiplication, motion and pile-up of dislocations and twinning in stable 316L stainless steel, phase transition such as strain-induced austenite-to-martensite phase transformation in metastable 301LN steel, and twinning in TWIP steel. Pop-ins associated with deformation twinning occur at a lower displacement in TWIP steel as compared to 316L steel, consistent with the lower stacking fault energy of TWIP steel. Both strain-induced martensite formation and twinning involve variant selection.

  11. Neutron irradiation effect on thermomechanical properties of shape memory alloys

    International Nuclear Information System (INIS)

    Alloys of Ti-Ni, Ti-Ni-Pd, Fe-Mn-Si, Mn-Cu-Cr, Mn-Cu, Cu-Al-Mn, Cu-Al-Ni systems are investigated after irradiation in IVV-2M reactor at various temperatures with neutron fluence of 1019 - 1020 cm-2. The degradation of shape memory effect in titanium nickelide base alloys is revealed after irradiation. Mn-Cu and Mn-Cu-Cr alloys show the best results. Trends in shape memory alloy behaviour depending on irradiation temperature are found. A consideration is given to the possibility of using these alloys for components of power reactor control and protection systems

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  13. Improved swelling resistance for PCA austenitic stainless steel under HFIR irradiation through microstructural control

    International Nuclear Information System (INIS)

    Six microstructural variants of Prime Candidate Alloy (PCA) were evaluated for swelling resistance during HFIR irradiation, together with several heats of type 316 stainless steel (316). Swelling was negligible in all the steels at 3000C after approx. 44 dpa. At 500 to 6000C 25%-cold-worked PCA showed better void swelling resistance than type 316 at approx. 44 dpa. There was less swelling variability among alloys at 4000C, but again 25%-cold-worked PCA was the best. Microstructurally, swelling resistance correlated with development of fine, stable bubbles whereas high swelling was due to coarser distributions of bubbles becoming unstable and converting to voids (bias-driven cavities)

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

    Directory of Open Access Journals (Sweden)

    E. Rożniata

    2011-12-01

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

  15. A proposal to alloy design for low activation high manganese austenitic stainless steel - role of carbon and nitrogen

    International Nuclear Information System (INIS)

    The role of carbon and nitrogen in high Mn-Cr-Fe base alloy has been investigated in order to propose a favorable starting composition for a low activation austenitic stainless steel. The base composition of Fe-12% Cr-15% Mn was selected by the results of our previous study, because of prevention of δ ferrite formation and retardation of σ phase. The combined addition of carbon and nitrogen is very beneficial in making a stable γ phase, preventing σ phase formation and increasing high-temperature strength at around 875 K. Cold work of 20% is also very useful in increasing creep rupture strength because of finely dispersed precipitation of carbide during creep. From the consideration of these results, an alloy system of Fe-12% Cr-15% Mn-0.2% C-0.2% N has been designed as one of the preferable primary low activation austenitic stainless steel. (orig.)

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

    International Nuclear Information System (INIS)

    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). In addition, several phases devear preheating unit anium southward as well as to deeper dened al half-lives with experimental ones, over a range of 24 orders of magnitude was obtained. This is a strong argument that the alpha decay could be considered a fission process with very high mass asymmetry and charge density asymmetry

  17. Growth of austenite from as-quenched martensite during intercritical annealing in an Fe–0.1C–3Mn–1.5Si alloy

    International Nuclear Information System (INIS)

    The growth of austenite from as-quenched martensite during intercritical annealing was studied in a quaternary Fe–0.1C–3Mn–1.5Si alloy. Fine austenite grains either grew from interlath-retained austenite films or were newly nucleated at lath and martensite packet boundaries. Both types grew to a size comparable to the width of the martensite lath. It was found both metallographically and by dilatometry that the austenite grew to an amount in excess of the volume fraction at final equilibrium. Simulation by DICTRA, which assumed local equilibrium at the α/γ boundary, confirmed that the development of austenite is composed of three stages: initial negligible-partitioning growth controlled by rapid carbon diffusion in ferrite, which is gradually replaced by carbon diffusion in austenite; intermediate slow growth, controlled by diffusion of Mn and/or Si in ferrite; and a final stage controlled by diffusion of substitutional elements in austenite for final equilibration, which may result in the shrinkage of austenite. The formation of austenite in excess of the equilibrium amount is considered to occur due to very slow substitutional diffusion in the growing austenite compared to the boundary migration.

  18. Modelling irradiation creep of zirconium alloys

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

  1. Swelling and in-pile creep of neutron irradiated 15Cr15NiTi austenitic steels in the temperature range of 400 to 600 deg. C

    International Nuclear Information System (INIS)

    A pressurized tube experiment was carried out in the Prototype Fast Reactor (PFR) ad Dounreay in order to determine swelling, stress-induced swelling and in-pile creep of different austenitic steels. The tubes were made out of different heats of the commercial German austenitic steel DIN 1.4970 and a number of model plain Fe-15Cr-15Ni stainless steels. Special attention was paid on the influence of minor alloying elements like Si, Ti, degree of Ti/C relation and others. The maximum doses achieved are 106 dpaNRT at 420 deg. C, 81 dpaNRT 500 deg. C and 61 dpaNRT at 600 deg. C. The hoop stresses of the pressurized tubes were 0, 60 and 120 MPa at all irradiation temperatures. The length and diameter changes of the pressurized capsules have been determined at up to four intermediate stages and after irradiation. Post irradiation examinations by immersion density measurements and transmission electron microscopy (TEM) are partially done. All alloys exhibited the highest swelling values at 420 deg. C and nearly no swelling at 600 deg.C. The measurements show the large effect of the minor alloying elements upon swelling and in-pile creep. The maximum swelling suppression is achieved for DIN 1.4970 through a high Si-content and an under stoichiometric Ti/C relation (under stabilization). This yields linear swelling of 1.9% after 106 dpaNRT at 420 deg. C. The formerly observed inter correlation between swelling and in-pile creep is confirmed up to 106 dpaNRT. It can be described by an equation consisting of a SIPA term (stress induced preferential absorption) and an inter correlation term similar to the I-creep proposed by Gittus. The estimates of the stress-induced swelling using the Soderberg theorem and the length measurements are compared with the immersion density measurements and results by TEM. The immersion density measurements agree rather good with length measurements. The stress-induced linear swelling can reach values of 0.8% at 100 dpaNRT and 120 MPa hoop stress

  2. The Effects of CO2 Pressure on Corrosion and Carburization Behaviors of Chromia-forming Austenitic Alloys

    International Nuclear Information System (INIS)

    By applying S-CO2 cycle to SFR, the inherent safety could be improved by alleviating the concern of explosive reaction between high temperature steam and liquid sodium as well as increased thermal efficiency at 500-550 .deg. C compared to helium Brayton cycle. Meanwhile, from the material point of view, a compatibility such as corrosion and carburization of candidate materials in S-CO2 environment should be evaluated to assure the long-term integrity of IHX. It has been previously reported that Ni-base alloys and high-Cr Fe-base austenitic alloys showed a good corrosion resistance by the formation of thin chromia layer while carburization behaviors of those materials were not properly investigated. Corrosion and carburization behaviors of three chromia-forming austenitic alloys (Ni-base alloys and Alloy 800HT) were evaluated in S-CO2 (200 bar) and CO2 (1 bar) environment at 550.650 .deg. C for 1000 h. For all test materials, a good corrosion resistance was exhibited by the formation of thin chromia (Cr2O3) with small amount of minor oxides such as Mn1.5Cr1.5O4, Al2O3, and TiO2

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

    International Nuclear Information System (INIS)

    High- and commercial-purity specimens of Type 304 SS from BWR absorber rod tubes, irradiated during service to fluence levels of 6 x 1020 to 2 x 1021 n·cm-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

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

    International Nuclear Information System (INIS)

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

  5. Prediction of Irradiation Damage by Artificial Neural Network for Austenitic Stainless Steels

    International Nuclear Information System (INIS)

    The internal structures of pressurized water reactors (PWR) located close to the reactor core are used to support the fuel assemblies, to maintain the alignment between assemblies and the control bars and to canalize the primary water. In general these internal structures consist of baffle plates in solution annealed (SA) 304 stainless steel and baffle bolts in cold worked (CW) 316 stainless steel. These components undergo a large neutron flux at temperatures between 280 and 380 .deg. C. Well-controlled irradiation-assisted stress corrosion cracking (IASCC) data from properly irradiated, and properly characterized, materials are sorely lacking due to the experimental difficulties and financial limitations related to working with highly activated materials. In this work, we tried to apply the artificial neural network (ANN) approach, predicted the susceptibility to an IASCC for an austenitic stainless steel SA 304 and CW 316. G.S. Was and J.-P. Massoud experimental data are used. Because there is fewer experimental data, we need to prediction for radiation damage under the internal structure of PWR. Besides, we compared experimental data with prediction data by the artificial neural network

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

    Science.gov (United States)

    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.

  7. Self-irradiation study of plutonium alloys

    International Nuclear Information System (INIS)

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

  8. High-energy X-ray diffraction study on the temperature-dependent mechanical stability of retained austenite in low-alloyed TRIP steels

    International Nuclear Information System (INIS)

    The stability of the retained austenite has been studied in situ in low-alloyed transformation-induced-plasticity (TRIP) steels using high-energy X-ray diffraction during tensile tests at variable temperatures down to 153 K. A detailed powder diffraction analysis has been performed to probe the austenite-to-martensite transformation by characterizing the evolution of the phase fraction, load partitioning and texture of the constituent phases simultaneously. Our results show that at lower temperatures the mechanically induced austenite transformation is significantly enhanced and extends over a wider deformation range, resulting in a higher elongation at fracture. Low carbon content grains transform first, leading to an initial increase in average carbon concentration of the remaining austenite. Later the carbon content saturates while the austenite still continues to transform. In the elastic regime the probed {h k l} planes develop different strains reflecting the elastic anisotropy of the constituent phases. The observed texture evolution indicates that the austenite grains oriented with the {2 0 0} plane along the loading direction are transformed preferentially as they show the highest resolved shear stress. For increasing degrees of plastic deformation the combined preferential transformation and grain rotation results in the standard deformation texture for austenite with the {1 1 1} component along the loading direction. The mechanical stability of retained austenite in TRIP steel is found to be a complex interplay between carbon concentration in the austenite, grain orientation, load partitioning and temperature.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Grandjean, Y.

    1996-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  12. Radiation induced segregation at grain boundary in an austenitic stainless steel under ion irradiation

    International Nuclear Information System (INIS)

    Chromium depletion near grain boundaries of austenitic stainless steel during irradiation was investigated. Specimens were kept at 1,473 K for 30 min, and were quenched into the water. Irradiations were done using 400 keV He+ ions at 573, 673 and 773 K up to 10 dpa with a dose rate of 2.4x10-4 dpa/s. After irradiation, the Cr concentration profile near the grain boundary was measured using an analytical electron microscope with a 1 nm beam diameter. At 573 K, Cr depletion is small, and its concentration at the grain boundary decreases to 15.5 mass% at 3 dpa from the initial concentration of 18.5 mass%. At 673 and 773 K, Cr concentration at the grain boundary rapidly decreases between 0 and 0.2 dpa, and then gradually approaches a constant value, 7.0 mass% at 673 K and 5.0 mass% at 773 K. Two stages are found in radiation induced segregation (RIS) behavior, one stage in which Cr depletion and Ni enrichment balance and another in which Fe depletion and Ni enrichment balance. These experimental results were compared with the calculations based on the vacancy-induced inverse Kirkendall effect. Predicted Cr segregation at 673 and 773 K above 3 dpa agrees with the experimental results. But Cr depletions at low doses which were obtained in the experiments are much faster than calculated. At 573 K in the experiments, depletion is smaller than calculated up to 10 dpa. (author)

  13. Mechanism of hydrogen embrittlement in a gamma-prime phase strengthened Fe–Ni based austenitic alloy

    International Nuclear Information System (INIS)

    The mechanism of hydrogen embrittlement (HE) in a γ′-Ni3(Al,Ti) phase strengthened Fe–Ni based austenitic alloy has been investigated in detail. Hot hydrogen charging experiment and tensile test reveal that the alloy with coherent γ′ phase exhibits a much higher decrease in reduction of area (RA) than that of the alloy in the solution-treated state. However, three-dimensional atom probe (3DAP) experiment shows that segregation of hydrogen atoms is not found at the coherent interface between the γ′ phase and the matrix, which indicates that the interface is not a strong hydrogen trap. Furthermore, high-resolution transmission electron microscopy (TEM) observation indicates that the interface coherency is maintained during the deformation, even tensile to fracture. It is found that macroscale slip band rupture and intergranular fracture are promoted by serious dislocation planar slip, which become the predominant features in the tensile-to-fracture sample after hydrogen charging. This phenomenon has been interpreted as a result of combined effects of the γ′ phase and hydrogen in the precipitation-strengthened Fe–Ni based austenitic alloy.

  14. Swelling under 1 MeV electron irradiation of some martensito-austenitic and ferrito-martensitic steels

    International Nuclear Information System (INIS)

    Swelling results by 1 MeV electron irradiation of two steels, one martensito-austenitic named climax and one ferrito-martensitic EM 12 capable of being used in a fuel assembly. The climax type steel is excluded from a hot use due to a phase change which carries along a prohibitive overswelling; it is shown also that EM 12 sweels between 400 and 5000C

  15. Fracture toughness of irradiated stainless steel alloys

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-08-01

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

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

    CERN Document Server

    Subbotina, A V

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    The microstructures of a series of silicon-modified AISI 316 alloys irradiated to fast neutron fluences of about 2-3 and 10 x 1022 n/cm2 (E > 0.1 MeV at temperatures ranging from 4000C to 6000C have been examined. The irradiation of AISI 316 leads to an extensive repartition of several elements, particularly nickel and silicon, between the matrix and various precipitate phases. The segregation of nickel at void and grain boundary surfaces at the expense of other faster-diffusing elements is a clear indication that one of the mechanisms driving the microchemical evolution is the Inverse Kirkendall effect. There is evidence that at one sink this mechanism is in competition with the solute drag process associated with interstitial gradients

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

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, EH

    2001-08-01

    Austenitic 316LN alloy was ion-irradiated using the unique Triple Ion Beam Facility (TIF) at ORNL to investigate radiation damage effects relevant to spallation neutron sources. The TIF was used to simulate significant features of GeV proton irradiation effects in spallation neutron source target materials by producing displacement damage while simultaneously injecting helium and hydrogen at appropriately high gas/dpa ratios. Irradiations were carried out at 80, 200, and 350 C using 3.5 MeV Fe{sup 2}, 360 keV He{sup +}, and 180 keV H{sup +} to accumulate 50 dpa by Fe, 10,000 appm of He, and 50,000 appm of H. Irradiations were also carried out at 200 C in single and dual ion beam modes. The specific ion energies were chosen to maximize the damage and the gas accumulation at a depth of {approx} 1 {micro}m. Variations in microstructure and hardness of irradiated specimens were studied using transmission electron microscopy (TEM) and a nanoindentation technique, respectively. TEM investigation yielded varying damage defect microstructures, comprising black dots, faulted and unfaulted loops, and a high number density of fine bubbles (typically less than 1 nm in diameter). With increasing temperature, faulted loops had a tendency to unfault, and bubble microstructure changed from a bimodal size distribution to a unimodal distribution. Triple ion irradiations at the three temperatures resulted in similar increases in hardness of approximately a factor of two. Individually, Fe and He ions resulted in a similar magnitude of hardness increase, whereas H ions showed only a very small effect. The present study has yielded microstructural information relevant to spallation neutron source conditions and indicates that the most important concern may be radiation induced hardening and associated ductility loss.

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

    International Nuclear Information System (INIS)

    Austenitic 316LN alloy was ion-irradiated using the unique Triple Ion Beam Facility (TIF) at ORNL to investigate radiation damage effects relevant to spallation neutron sources. The TIF was used to simulate significant features of GeV proton irradiation effects in spallation neutron source target materials by producing displacement damage while simultaneously injecting helium and hydrogen at appropriately high gas/dpa ratios. Irradiations were carried out at 80, 200, and 350 C using 3.5 MeV Fe2, 360 keV He+, and 180 keV H+ to accumulate 50 dpa by Fe, 10,000 appm of He, and 50,000 appm of H. Irradiations were also carried out at 200 C in single and dual ion beam modes. The specific ion energies were chosen to maximize the damage and the gas accumulation at a depth of ∼ 1 microm. Variations in microstructure and hardness of irradiated specimens were studied using transmission electron microscopy (TEM) and a nanoindentation technique, respectively. TEM investigation yielded varying damage defect microstructures, comprising black dots, faulted and unfaulted loops, and a high number density of fine bubbles (typically less than 1 nm in diameter). With increasing temperature, faulted loops had a tendency to unfault, and bubble microstructure changed from a bimodal size distribution to a unimodal distribution. Triple ion irradiations at the three temperatures resulted in similar increases in hardness of approximately a factor of two. Individually, Fe and He ions resulted in a similar magnitude of hardness increase, whereas H ions showed only a very small effect. The present study has yielded microstructural information relevant to spallation neutron source conditions and indicates that the most important concern may be radiation induced hardening and associated ductility loss

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

  5. Fracture behavior of neutron-irradiated high-manganese austenitic steels

    International Nuclear Information System (INIS)

    The instrumented Charpy impact test was applied to study the fracture behavior of high-manganese austenitic steels before and after neutron irradiations. Quarter-size specimens of a commercial high-manganese steel (18% Mn-5% Ni-16% Cr), three reference steels (21% Mn-1% Ni-9% Cr, 20% Mn-1% Ni-11% Cr, 15% Mn-1% Ni-14% Cr) and two model steels (17% Mn-4.5% Si-6.5% Cr, 22% Mn-4.5% Si-6.5% Cr-0.2% N) were used for the impact tests at temperatures between 77 and 523 K. The load-deflection curves showed typical features corresponding to characteristics of the fracture properties. The temperature dependences of fracture energy and failure deflection obtained from the curves clearly demonstrate only small effects up to 2x1023 n/m2 (E > 0.1 MeV) and brittleness at room temperature in 17% Mn-Si-Cr steel at 1.6x1025 n/m2 (E > 0.1 MeV), while ductility still remains in 22% Mn-Si-Cr steel. (orig.)

  6. Dependence of Steady-state swelling rate of 0.1C-16Cr-15Ni-3Mo-1Mn austenitic steel on Dpa rate and irradiation temperature

    Energy Technology Data Exchange (ETDEWEB)

    Kozlov, A.V.; Portnykh, I.A. [FSUE, Institute of Nuclear Materials, Zarechney, Sverdlovsk (Russian Federation)

    2007-07-01

    Full text of publication follows: In order to predict void swelling under high dose neutron irradiation for fission and fusion spectra it is needed to describe both the onset (transient duration) of steady-state swelling as well as the steady-state rate. One current conception is that the steady-state swelling rate of austenitic alloys is {approx}1 %/dpa, independent of dpa rate and temperature over a rather broad range of typical irradiation conditions. The objective of the paper is to explore the behavior of the steady-state swelling rate with temperature and damage rate, using {approx}1800 swelling measurements made on 0.1C-16Cr-15Ni-3Mo-1Mn austenitic steel used as fuel cladding at 370-600 deg. C in the BN-600 fast reactor. These data were analyzed over 10 deg. C increments and the steady-state rate was determined for swelling levels above 10% for each increment. The steady-state swelling rate was found to be relatively invariant at somewhat less than 1%/dpa over most of the temperature range but to climb strongly to levels in excess of 2% /dpa at the highest temperatures. A quantitative model of point defect migration in austenitic steels to sinks generated under irradiation was developed using statistical thermodynamics of solid bodies. Vacancy and interstitial fluxes to voids, dislocations and precipitates developed in this steel were calculated using earlier experimentally derived descriptions of temperature dependencies of void integral surfaces for this steel. The model was calibrated to fit the derived temperature dependence at 10{sup -6} dpa/s characteristic of BN-600. The model was then used to explore the dependence of the steady-state swelling rate on dpa rate over a very wide range. Over most of the reactor-relevant temperature range the predicted steady-state rate is independent of dpa rate. At the highest temperatures with higher steady-state swelling rates, the swelling rate is shown to be independent of dpa rate over a much smaller range of dpa

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

    Institute of Scientific and Technical Information of China (English)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 x 1025 neutrons.m-2, E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 x 1023 m-3) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen

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

    OpenAIRE

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

    2009-01-01

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

  11. Effects of alloys elements, impurities and microstructural factors in austenitic stainless steel to utilize in fuel rod of nuclear reactors

    International Nuclear Information System (INIS)

    Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 1022n/cm2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 6500C. (author)

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

    Directory of Open Access Journals (Sweden)

    Białobrzeska B.

    2015-09-01

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

  13. Change of relative Gibbs energy of martensite and austenite alloys of Fe-Ni system in the pre-martensite temperature range

    International Nuclear Information System (INIS)

    Chemical potentials of the components of quenched Fe-Ni alloys (28.7-32.7 at. % Ni) with martensite and austenite structures have been found with the Touch Instant Electromotive Force method. Differences between Gibbs energies of martensite and austenite phases have been calculated in the temperature range of 253-315 K which characterize the relative thermodynamic stability of these metastable phases. By means of interpolation the temperatures were determined when Gibbs energies of alloys with both types of structures are the same. Non-chemical contribution into Gibbs energy of martensite transformation has been evaluated

  14. Effect of ion irradiation on the microstructure of an iron--nickel--chromium alloy

    International Nuclear Information System (INIS)

    Void and disloation structures in an Fe-25Ni-15Cr alloy were studied following irradiation with 2.8 MeV 58Ni+ ions at temperatures between 600 and 7500C (1112 and 13820F) to maximum damage levels up to 80 displacements per atom (dpa). Void formation was observed at all the temperatures investigated, with the maximum swelling between 650 and 7000C (1202 and 12920F). The swelling versus dose relations exhibited an incubation dose followed by swelling at a rate that increased with increasing damage level. These data were consistent with previous swelling results for austenitic alloys irradiated with charged particles, which indicate that the swelling should become linear with irradiation dose at higher damage levels. Tangled dislocation networks were observed to form at low doses and to be fairly stable up to the highest damage levels examined. With the assumption of the observed stable dislocation networks, the dose dependence of swelling could be explained by a general form of the chemical rate theory for swelling due to void growth

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  16. Neutron irradiation creep at 100 C on 316L, AMCR, and welded 316L stainless steel alloys

    International Nuclear Information System (INIS)

    The authors performed measurements on the elongation of many different austenitic stainless steel alloys irradiating at 100 C in a low flux channel of the High Flux Reactor at Petten varying the applied stress between zero and 300 Mpa. They irradiated in total 98 samples in two irradiation rigs. Of these samples only 26 samples could be tested up to a dose of 2.1 dpa, and 13 samples up to a dose of 0.21 dpa. The steels tested are listed in Table 1. In the second irradiation rig four TIG-welded samples and one EB-welded sample were irradiated. They found that the length of the samples increased up to an irradiation dose of 0.11 dpa and then either decreased or increased slightly depending on the magnitude of the applied stress. They attributed the increase in length to the volume change due to the formation of carbides and to the accommodation of carbides to the applied stress. The decrease of the length with irradiation time is attributed to the formation of brittle α-ferrite. The amount of α-ferrite formed increases with decreasing irradiation temperature and increases with decreasing applied stress. Eight samples broke during irradiation in 8 columns or stems in two rigs before the first elongation test at 0.11 or 0.21 dpa could be performed. Irradiation of 343 samples of the same materials in the last fifteen years at temperatures between 300 and 500 C did not cause fracture

  17. Neutron irradiation creep at 100 C on 316L, AMCR, and welded 316L stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hausen, H.; Schuele, W. [Johann Wolfgang Goethe-Univ. Frankfurt (Germany). Inst. fuer Angewandte Physik

    1999-10-01

    The authors performed measurements on the elongation of many different austenitic stainless steel alloys irradiating at 100 C in a low flux channel of the High Flux Reactor at Petten varying the applied stress between zero and 300 Mpa. They irradiated in total 98 samples in two irradiation rigs. Of these samples only 26 samples could be tested up to a dose of 2.1 dpa, and 13 samples up to a dose of 0.21 dpa. The steels tested are listed in Table 1. In the second irradiation rig four TIG-welded samples and one EB-welded sample were irradiated. They found that the length of the samples increased up to an irradiation dose of 0.11 dpa and then either decreased or increased slightly depending on the magnitude of the applied stress. They attributed the increase in length to the volume change due to the formation of carbides and to the accommodation of carbides to the applied stress. The decrease of the length with irradiation time is attributed to the formation of brittle {alpha}-ferrite. The amount of {alpha}-ferrite formed increases with decreasing irradiation temperature and increases with decreasing applied stress. Eight samples broke during irradiation in 8 columns or stems in two rigs before the first elongation test at 0.11 or 0.21 dpa could be performed. Irradiation of 343 samples of the same materials in the last fifteen years at temperatures between 300 and 500 C did not cause fracture.

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

    Science.gov (United States)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  1. Effective interactions approach to phase stability in alloys under irradiation

    International Nuclear Information System (INIS)

    Phase stability in alloys under irradiation is studied considering effective thermodynamic potentials. A simple kinetic model of a binary alloy with phase separation is investigated. Time evolution in the alloy results from two competing dynamics: thermal diffusion, and irradiation induced ballistic exchanges The dynamical (steady state) phase diagram is evaluated exactly performing Kinetic Monte Carlo simulations. The solution is then compared to two theoretical frameworks: the effective quasi-interactions model as proposed by Vaks and Kamishenko, and the effective free energy model as proposed by Martin. New developments of these models are proposed to allow for quantitative comparisons. Both theoretical frameworks yield fairly good approximations to the dynamical phase diagram

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    International Nuclear Information System (INIS)

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

  4. The formation of radiation-induced segregation at twin bands in ion-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Radiation-induced segregation (RIS) at twins was investigated using transmission electron microscopy (TEM) for ion-irradiated austenitic stainless steel. Significant RIS was found to occur at twin boundaries. TEM analysis indicates that interfacial dislocations at partially coherent twin boundaries are potential sites for strong RIS phenomenon. The RIS causes the formation of thin bands having a higher Ni and lower Cr concentration in twin bands with a width less than 15 nm. In wider twin bands, strong RIS occurs only at the outer twin boundaries, but not inside the band. The possible mechanism for the formation of the RIS thin band is discussed

  5. Water radiolysis effect on IASCC growth behavior in BWR water conditions in highly irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    For study of water radiolysis effect caused by gamma-rays from radioactive material on irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth tests in highly irradiated austenitic stainless steel are performed in simulated BWR water conditions (at ∼563 K). The compact tension (CT) specimens made of 316L stainless steels are irradiated with neutrons up to ∼12 dpa in the Japan Materials Testing Reactor (JMTR). Post-irradiation annealing (PIA) at 973 K for 1 hour is applied to one of the specimens, which shows the recovery of material properties corresponding to the unirradiated ones but the radioactivity of highly irradiated material as it is. The gamma-ray absorbed dose rate in water is calculated near the crack tip of the CT specimen, and the stable concentrations of H2O2, O2 and H2 in water near the crack tip are estimated by radiolysis calculation for some feed water conditions of normal water chemistry (NWC), deaerated water and hydrogen water chemistry (HWC). The preliminary results of the crack growth rate (CGR) for the highly irradiated specimens and the annealed specimen are presented, and the relationship between the CGRs and the water chemistry such as the concentrations of radiolytic species and the electrochemical corrosion potential (ECP) is discussed. (author)

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

    Science.gov (United States)

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

    2016-12-01

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

  7. Current research into the self-irradiation effects of δ-phase Pu-Ga alloys

    International Nuclear Information System (INIS)

    Delta-phase Pu-Ga alloys (δ-Pu) radiogenically age unlike most face-centered cubic (FCC) metals such as austenitic steels, Al, Cu, and other structural alloys when exposed to irradiation. If it did, it would exhibit void swelling at 20-40 years of age or 2-4 dpa. Why doesn't δ-Pu void swell? Will aging δ-Pu start to void swell sometime in the future? Currently, we don't know the answers to these questions. Pu is a notoriously unstable metallic element with multiple phases and the properties of technological Pu alloys are changing with time due to Pu α-decay. The most influential drivers for radiogenic change in δ-Pu are He ingrowth, radiation-induced lattice defect accumulation and phase instability of δ-Pu. Each of these drivers induces changes that depend on temperature, Ga concentration, radioactive α-decay dose, and α-decay rate and have profound implications for materials chemistry, physics and engineering. How do these radiogenic products interact with each other with an inherently unstable lattice? When lattice damage recovery processes do occur, is it complete or do residual effects remain? Because each driver has a different dependence on these controllable variables, the appropriate set of theory, computation and experiments can provide a unique opportunity to determine accurately the strength of each driver in determining the response of Pu to radiogenic processes

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-10-01

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

  9. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    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.

  11. Precipitation and irradiation damage in proton-irradiated palladium-chromium alloys

    International Nuclear Information System (INIS)

    Irradiation damage of Pd-Cr alloys containing 15, 20 and 25 at.% Cr was studied over the temperature range 100 to 5500C, primarily in samples irradiated to a dose of 0.7 d.p.a. The solubility limit in this range of temperatures varies from 22 to 38% Cr, and precipitation of a phase having the L12 crystal structure was observed in unirradiated samples of the 25% Cr alloy aged at temperatures as low as 1000C. Octahedrally shaped voids, with faces parallel to {1 1 1}, were found only in the 25% Cr alloy irradiated from 350 to 5500C, but not in the other alloys at any temperature. The undersized chromium atoms migrate to point-defect sinks during irradiation, resulting in solute segregation and, eventually, precipitation under certain conditions. The precipitation of the L12 phase, was irradiation-induced at dislocation loops, voids and grain boundaries in the more concentrated undersaturated alloys. This precipitation was observed in the mid-range and surface regions of the samples containing 20 and 25% Cr, but not in those containing 15% Cr. Comparisons were made with Pd-Fe and Ni-Si alloys, which had also been proton-irradiated, and the similarities and differences noted and discussed. (author)

  12. The mechanism of irradiation hardening of iron model alloys

    International Nuclear Information System (INIS)

    The mechanism of irradiation hardening of iron-model alloys is reviewed. Irradiation hardening consists of several components which are caused by several sorts of damage structures, namely hardeners. The contribution of each hardener, that is, cooper clusters and matrix defects such as microvoids and interstitial dislocation loops (I-loops) to irradiation hardening is quantitatively evaluated based on the microstructure-hardening correlation studies. Neutron and electron irradiation hardening and post-irradiation annealing behavior in iron model alloys have been investigated. Neutron irradiations were performed in the Japan Material Testing Reactor (JMTR) up to the maximum neutron fluence of 7 x 1019 n/cm2 (>1MeV) with different neutron fluxes ranging from 1.5 x 1011 to 1013 n/cm2, with using a multi-division temperature-control irradiation rig that enabled to remove a part of the specimens during reactor operation under controlling irradiation temperature. Electron irradiations were performed up to electron doses between 2.1 x 1019 e-/cm2 (0.9 mdpa) and 5.1 x 1020 e-/cm2 (22 mdpa) at around 290degC. Tensile tests were carried out at a crosshead speed of 0.2 mm/min at room temperature. Micro-Vickers hardness tests, positron annihilation spectroscopy (PAS) measurements and electrical resistivity measurements using a conventional four probes method were carried out at room temperature. Post-irradiation annealing behavior of the hardening and microstructure changes were also investigated by these experimental methods. The amount of irradiation hardening of iron-copper alloys is significantly larger than those of pure Fe, indicating that copper clusters are an important factor of irradiation hardening. The contribution of copper clusters to the hardening is rather large but almost similar to the hardening caused by I-loops in copper alloys. Microvoids are not a high potential hardener. Another high potential hardener is manganese. The iron-manganese model alloy suffers

  13. Contribution to the high temperature deformation behaviour of the austenitic steel 1.4981 after neutron irradiation

    International Nuclear Information System (INIS)

    The austenitic steel 1.4981 thermomechanically treated to provide a range of microstructures has been irradiated at temperatures of 400-5000C with neutron fluxes of 1,55 x 1023 to 3 x 1025 m-2. The degree of high temperature helium embrittlement caused by (n, α) nuclear reactions was determined in post-irradiation tensile and stress rupture tests at temperatures in the range 400-8500C. Under the irradiation conditions used, the production of helium within the grains was so great that the helium concentration at grain boundaries which leads to ductility loss was controlled by the diffusion of helium. At test temperatures below 7000C, the difference in the duration of tensile and of stress rupture tests has a significant effect. The test results show that the microstructural variations of the steel 1.4981 tested under the same conditions exhibited similar relative ductility loss. However, the absolute value of the rupture elongation after irradiation was proportional to the ductility of the material in the unirradiated condition, which was in turn dependent on the thermomechanical treatment. Specimens which had been embrittled by helium were found to contain numerous intercrystalline microcracks, which were orientated normal to the stress direction and eventually caused rupture. The elongation of the specimens resulted mainly from the widening of the microcracks; the grains themselves were not elongated to the same extent as found in unirradiated specimens. This explains why for the irradiated specimens little or no secondary creep was observed. (orig.)

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

    Science.gov (United States)

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

    2016-05-01

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

  15. Cyclotron irradiation of Cu3Au alloys at low temperatures

    International Nuclear Information System (INIS)

    Ordered and disordered Cu3Au alloys were irradiated at liquid nitrogen and helium temperatures with protons and α-particles from a cyclotron. An increase of resistivity by irradiation and its recovery by isochronal annealing were measured. The results were compared with those of copper irradiated and annealed under the same conditions. Defect introduction rates were carefully determined, based on which a resistivity due to Frenkel pairs in the alloy and a number of replacements in a cascade were estimated. The recovery of the order in the alloy occurred in the stage III and in the 4000K stage but not in the stage I and II, which suggests that the migrating defects in the former stages are vacancy families and those in the latter stages are interstitial families. It was shown that in the stage I annealing of the ordered alloy, the similar processes proceed as in the pure copper, while in the disordered alloy the long range migration of defects is prevented. It was also shown that the defects survived the stage I annealing in the disordered alloy disappear in the stage II

  16. Effects of self-irradiation in plutonium alloys

    Science.gov (United States)

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

    2016-04-01

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

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

    Science.gov (United States)

    Timofeeva, L. F.

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ermi, A.M. [Westinghouse Hanford Company, Richland, WA (United States); Gelles, D.S. [Pacific Northwest Laboratory, Richland, WA (United States)

    1995-04-01

    The objective of this effort was to irradiate a series of alloys over the temperature range 300 to 600{degrees}C to approximately 10 dpa in the High Flux Isotope Reactor (HFIR). The alloys covered a wide range of materials and treatments. The Japanese specimen matrix consisted of ferritic steels, vanadium alloys, copper alloys, molybdenum alloys, and titanium-aluminum compounds. The US specimen matrix consisted of vanadium alloys, 316 stainless steels, and isotopically tailored ferritic and austenitic alloys.

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

    Institute of Scientific and Technical Information of China (English)

    Y. Sun

    2004-01-01

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

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

    Science.gov (United States)

    Maksimkin, O. P.

    2016-04-01

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

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

    International Nuclear Information System (INIS)

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

  2. Preliminary evaluation of irradiation characteristics of new K alloys irradiated in HANARO

    International Nuclear Information System (INIS)

    Korean Atomic Energy Research Institute (KAERI) is trying various tests to develop zirconium based new alloys for nuclear fuel, which has better performance than that of Zircaloy-4 alloy. To evaluate the in-pile performance of newly developed K alloys preliminarily, KAERI had prepared the test specimens of K alloys, irradiated them upto the fluence of 8.63-9.27 x 1019n/cm2 at 320 ± 7 .deg. C in HANARO, and performed the hardness and tensile tests in IMEF. After the irradiation the hardness of K alloys increased from 24% to 37%, the yield strength from 17% to 37%, the ultimate tensile strength from 12% to 21% with the decease of maximum elongation from 6 to 39%

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

    International Nuclear Information System (INIS)

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

  4. Irradiation-induced precipitation in Ni--Si alloys

    International Nuclear Information System (INIS)

    The microstructures of Ni+ ion-irradiated Ni--Si solid-solution alloys, containing 2, 4, 6 and 8 at. percent Si were investigated as a function of dose, dose-rate, and temperature. Results of transmission electron microscopy and other data show the precipitation of γ' (Ni3Si) in all samples irradiated at 5000C. Characteristics of the precipitates are described and a mechanism for their formation is suggested. (U.S.)

  5. Comparison of the irradiated tensile properties of a high-manganese austenitic steel and type 16 stainless steel

    International Nuclear Information System (INIS)

    The USSR steel EP-838 is a high-manganese (13.5%), low-nickel (4.2%) steel that also has lower chromium and molybdenum than type 316 stainless steel. Tensile specimens of 20%-cold-worked EP-838 and type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at the coolant temperature (approx. 500C). A displacement damage level of 5.2 dpa was reached for the EP-838 and up to 9.5 dpa for the type 316 stainless steel. Tensile tests at room temperature and 3000C on the two steels indicated that the irradiation led to increased strength and decreased ductility compared to the unirradiated steels. Although the 0.2% yield stress of the type 316 stainless steel in the unirradiated condition was greater than that for the EP-838, after irradiation there was essentially no difference between the strength or ductility of the two steels. The results indicate that the replacement of the majority of the nickel by manganese and a reduction of chromium and molybdenum in an austenitic stainless steel of composition near that for type 316 stainless steel has little effect on the irradiated and unirradiated tensile properties at low temperatures

  6. Investigations into radiation swelling and mechanical properties of irradiated austenitic chromium-manganese steel

    International Nuclear Information System (INIS)

    Presented are the results of investigations of material with fast-dropping radioactivity-austenitic Cr-Mn steel (04Kh12G14N4YuM), damage dose dependence of swelling and mechanical properties of steel. It is shown that 04Kh12G14N4YuM steel has high radiation swelling resistance up to a damage dose of 60 dpa at temperature ranging from 340 to 530 deg C. Mechanical properties of steel are as good as those of austenitic Cr-Ni steels up to 30 dpa and 40 deg C. Ductility does not decrease lower than 2-5 %. Dose and temperature increase leads to instability of γ-solid solution, that is confirmed by the results of TEM and X-ray examinations. 4 refs.; 6 figs.; 2 tabs

  7. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  8. Neutron and deuteron irradiation creep in stainless steel alloys

    International Nuclear Information System (INIS)

    Irradiation rigs have been developed for HFR in Petten in which 49 specimens at a time can be irradiated in uniaxial tension. The creep elongations of AMCR type steels and of 20% cold-worked US 316 stainless steels were measured at intervals of irradiation for tensile stresses varying between 100 and 130 MPa and for temperatures between 350 and 4200C. Deuteron irradiation creep of 0.13 mm thick and 20% cold-worked 316 stainless steel alloys was investigated for an irradiation temperature of 3000C by means of a torsional creep facility installed at the Ispra Cyclotron. It was found that the creep rate increases linearly with the displacement rate in the range between 10-6 to 10-5 dpa.s-1. The stress dependence of the irradiation creep rate is quadratic for stresses between 50 and 100 MPa and linear for stresses ranging between 100 and 150 MPa

  9. Neutron and deuteron irradiation creep in stainless steel alloys

    International Nuclear Information System (INIS)

    Irradiation rigs have been developed for HFR in Petten in which 49 specimens at a time can be irradiated in uniaxial tension. The creep elongations of AMCR type steels and of 20% cold-worked US 316 stainless steels were measured at intervals of irradiation for tensile stresses varying between 100 and 130 MPa and or temperatures between 350 and 4200C. Deuteron irradiation creep of 0.13 mm thick and 20% cold-worked 316 stainless steel alloys was investigated for an irradiation temperature of 3000C by means of a torsional creep facility installed at the Ispra Cyclotron. It was found that the creep rate increases linearly with the displacement rate in the range between 10-6 to 10-5 dpa.s-1. The stress dependence of the irradiation creep rate is quadratic for stresses between 50 and 100 MPa and linear for stresses ranging between 100 and 150 MPa. (author)

  10. C-O relations of the extremely low carbon austenitic stainless steels and nickel base high alloys in vacuum induction melting

    International Nuclear Information System (INIS)

    It is well known that in vacuum-melted austenitic stainless steel and nickel base alloy, the impurities of minute amounts affect adversely the corrosion resistance and high temperature strength. Therefore the materials of high quality, such as those in extremely low carbon range below 0.01%, are required in nuclear and chemical plants. In this study, austenitic stainless steel such as SUS 308, 309 and 316 and nickel base alloy such as Ni-20 Cr-2.6 Nb and Ni-20 Mo-3W were melted in a 200 kg vacuum induction furnace, and the behaviors of C and O during the refining were investigated, also the thermodynamical analysis was performed. For comparison, pure iron was studied at the same time. The amounts of C and O were reduced from the beginning of melting through intensive boiling period, and when quiescent period was reached, the equilibrium relation of C and O was able to be applied also to the case of austenitic stainless steel. In case of the nickel base alloy, it was presumed that the relation of C and O in quiescent period of molten alloy was near the equilibrium state. The partial pressure of CO in the stainless steel was low as compared with the pure iron, because the effect of refractory material to the oxygen potential of molten steel is different according to the steel composition. (auth.)

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

    Science.gov (United States)

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

    2001-07-01

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

  12. Irradiated behavior at high burnup for HiFi alloy

    International Nuclear Information System (INIS)

    Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased that to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film. (author)

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

    International Nuclear Information System (INIS)

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

  14. Consideration of the growth mode in isochronal austenite-ferrite transformation of ultra-low-carbon Fe-C alloy

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hao; Liu, Yongchang; Li, Yanli; Zhang, Lifang [Tianjin University, Tianjin Key Lab of Advanced Jointing Technology, School of Materials Science and Engineering, Tianjin (China); Yan, Zesheng [Tianjin Pipe (Group) Cooperation, Tianjin (China)

    2010-01-15

    The three cooling rates of 10, 100, 200 K/min dilatometry experiments are used to investigate the kinetics of the isochronal austenite ({gamma}) to ferrite ({alpha}) transformation of Fe-0.0036wt.%C alloy. ''Normal transformation'' and ''abnormal transformation'' have both been observed for transformations at different cooling rates. In accordance with the thermodynamic characteristics of the {gamma}{yields}{alpha} transformation investigated here and previous kinetic considerations, a JMAK-like approach for the kinetics of isochronal phase transformations was developed that incorporates three overlapping processes: site saturation nucleation, alternate growth modes (from interface-controlled to diffusion-controlled to interface-controlled growth), as well as impingement for random distribution nuclei. The JMAK-like approach has been employed to fit the experimental results, and the fitting results show that for the {gamma}{yields}{alpha} transformation of the Fe-C alloy at all applied cooling rates, the growth mode evolves in the corresponding order: from interface-controlled to diffusion-controlled growth; from interface-controlled to diffusion-controlled to interface-controlled growth; and interface-controlled growth. (orig.)

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

    International Nuclear Information System (INIS)

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

  16. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-09-01

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

  17. Neutron irradiation creep in stainless steel alloys

    Science.gov (United States)

    Schüle, Wolfgang; Hausen, Hermann

    1994-09-01

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

  18. Neutron irradiation creep in stainless steel alloys

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-08-01

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

  1. Neutron irradiation damage of a stress relieved TZM alloy

    International Nuclear Information System (INIS)

    The objective of this work is to study defect microstructures and irradiation hardening in a stress relieved TZM alloy after irradiation in the Fast Flux Test Facility (FFTF) using the Materials Open Test Assembly (MOTA). Disk specimens of the molybdenum alloy TZM that had been stress relieved at 1199 K (929 C) for 0.9 ks (15 min.) were irradiated in the FFTF/MOTA 1F at 679, 793 and 873 K (406, 520, and 600 C) to a fast fluence of ∼9.6 x 1022 n/cm2. Microstructures were observed in a transmission electron microscope (TEM). Dislocation structures consisted of isolated loops, aggregated loops (rafts) and elongated dislocations. The size of the loops increased with the irradiation temperature. Void swelling was about 1 and 2% at 793 and 873 K (520 and 600 C), respectively. A void lattice was developed in the body centered cubic (bcc) structure with a spacing of 26 - 28 nm. The fine grain size (0.5 - 2 μm) was retained following high temperature irradiation, indicating that the stress relief heat treatment may extend the material's resistance to radiation damage up to high fluence levels. Microhardness measurements indicated that irradiation hardening increased with irradiation temperature. The relationship between the microstructure and the observed hardening was determined

  2. Evolution of microstructure in zirconium alloys during irradiation

    CERN Document Server

    Griffiths, M; Winegar, J E

    1997-01-01

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

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

    Science.gov (United States)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

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

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

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

  6. Radio-induced brittleness of austenitic stainless steels at high temperatures

    International Nuclear Information System (INIS)

    In a first part, the author recalls some metallurgical characteristics and properties of iron (atomic properties, crystalline structure, transformation), of iron carbon systems and steels (ferrite, austenite, cementite, martensite, bainite, phase diagrams of iron chromium alloy and iron nickel alloy), aspects regarding the influence of addition elements in the case of stainless steels (mutual interaction of carbon, chromium and nickel in their iron alloys, indication of the various stainless steels, i.e. martensitic, ferritic, austenitic, austenitic-ferritic, and non ferrous), and presents and discusses various mechanical tests (tensile tests, torsion tests, resilience tests, hardness tests, creep tests). In a second part, he discusses the effects of irradiation on austenitic stainless steels: irradiation and deformation under low temperature, irradiation at intermediate temperature, irradiation at high temperature. The third part addresses mechanisms of intergranular fracture in different temperature ranges (400-600, 700-750, and about 800 C). The author then discusses the effect of Helium on the embrittlement of austenitic steels, and finally evokes the perspective of development of a damage model

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

    International Nuclear Information System (INIS)

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

  8. Development of irradiation capsules in liquid metal environment in Joyo and their application to irradiation creep measurement of vanadium alloys

    International Nuclear Information System (INIS)

    In order to perform irradiation experiments in a liquid metal environment in a nuclear reactor, an irradiation technique with sodium bonding irradiation capsules was developed and a series of neutron irradiation experiments with sodium bonding irradiation capsules were performed in Joyo. The design and fabrication of sodium bonding capsules, sodium filling into capsules, capsule loading to Joyo, irradiation experiments, dismantling for irradiated capsules, removing the irradiated specimens from sodium-filled capsules, and sodium cleaning of the irradiated specimens were established through this study. Using the Joyo irradiation with the sodium bonding capsules where irradiation temperature was distributed uniformly, the irradiation creep experiment for highly purified V-4Cr-4Ti alloys, NIFS-Heat, was carried out and the knowledge about the irradiation creep behavior of the alloys was obtained. (author)

  9. Aspects of ion irradiations to study localized deformation in austenitic stainless steels

    International Nuclear Information System (INIS)

    The effect of irradiation depth on localized deformation in strained SUS304 and SUS316 was assessed by irradiation to various doses using different energies of Fe++ and protons. The average height and spacing of slip steps formed on the surfaces after 2% plastic strain in 300 °C argon were characterized using AFM and SEM. The step height and spacing were nearly unchanged for irradiation depth greater than 1/3 of the average grain size. 2.8 MeV Fe++ irradiation, with an irradiation depth of <1 μm, resulted in slightly larger step spacing compared to the unirradiated samples but it was much smaller compared to that in proton-irradiated samples with much greater irradiation depths. Step height and spacing also appeared to be affected by the grain size as they were nearly double in magnitude for 30 μm compared to 14 μm grain sizes in SUS304

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Irradiation creep and growth of guide thimble alloys

    International Nuclear Information System (INIS)

    The comprehension and the control of the fuel assembly distortion phenomena necessitate a better knowledge of the irradiation creep and growth of guide thimbles. Therefore, experiments on guide thimble samples have been conducted under neutron irradiation in the OSIRIS reactor facilities to characterize both phenomena. The samples were irradiated at temperatures between 570 and 588 K in a flux between 1 and 2 1018 n.m-2.s-1 in a NaK environment for lengths of time up to 3500 h. Axial creep tests were performed on tubular samples for stresses between 70 and 100 MPa, and the growth was determined from unstressed samples. The following alloys, all in the fully recrystallized metallurgical state have been studied: Zircaloy-4, ZrNb2.5O0.12 and ZrNbSnFe. These alloys show large differences in their growth behavior, as expected from the literature. However, although significant differences between these alloys are observed in out-of-pile axial creep tests, these differences become negligible under these irradiating conditions. (author)

  14. Effects of minor alloying additions on the strength and swelling behavior of an austenitic stainless steel

    International Nuclear Information System (INIS)

    A set of 32 alloys consisting of various additions of the elements Mo, W, Al, Ti, Nb, C and Si to an Fe-7.5 Cr-20 Ni alloy were made in order to investigate the effects of these solute additions on alloy swelling and strength. Both single and multiple additions were examined. The influence of various solute elements on the swelling behavior in the range 500 to 7300C was investigated using 4 MeV Ni ion bombardment to a dose 170 dpa. It was found that on an atomic percent basis, the elements may be arranged in order of decreasing effectiveness in reducing peak temperature swelling as follows: Ti, C, Nb, Si, and Mo. Small amounts of aluminum enhance swelling. Additions of Si, Ti, or Nb truncate the high temperature swelling regime of the ternary alloy. Mo, W, and C do not have a strong effect on the temperature dependence of swelling. The results may be interpreted in terms of the effect of point defect trapping on void growth rates, and it is suggested that the changes in peak temperature are the result of small changes in the free vacancy formation energy. A method for treating certain multiple additions is proposed. The effect of these alloying additions on short time high temperature strength properties was estimated using hot hardness measurements over the temperature range 22 to 8500C. On an atom percent basis Nb and Ti were most effective in conferring solid solution strengthening and Si the least effective. In the regime 22 to approximately 6500C, the hardness data was found to fit an equation of the form: H = H0 + b/T; where H is the hardness, T is the temperature, and H0 and b are constants for a given alloy. An empirical method was devised to estimate the hot hardness of alloys containing more than one solute addition

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Hydrogen desorption from austenitic steel specimens irradiated by high temperature pulse plasma fluxes

    International Nuclear Information System (INIS)

    Hydrogen desorption from the 12Kh18N10T austenitic steel specimens electrolytically saturated by hydrogen by interaction with deuterium plasma (energy content in the pulse 40-60 kJ) is studied. Formation of anomalously great blisters with the cap of ∼ 1 μm thick containing hydrocarbons in particular methane is identified. Three peaks were observed on the curve of the hydrogen thermodesorption from the specimens containing blisters. It is shown that the low-temperature peak (∼ 500 K) corresponds to the desorption of hydrogen over saturated from the blisters wherein it was present in the molecular form. The peak of the hydrogen desorption at ∼ 700 K corresponds to the hydrogen desorption from a solid solution. The high-temperature peak corresponds to the hydrogen desorption from the blisters, containing hydrocarbons by their dissociation. The mechanism of anomalous blisters formation is proposed

  18. Irradiated behavior for BWR advanced Zr alloy (HiFi alloy)

    International Nuclear Information System (INIS)

    Irradiation tests of BWR advanced Zr alloys (HiFi alloy) were carried out in a Japanese commercial reactor and the irradiation properties of the material were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. As a result of TEM observation, the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate in high fluence. Zr-Fe-Cr SPPs did not completely disappear for even 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, the electrochemical technique was applied for oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen fraction was estimated. From the relation, it was thought that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film. In addition, a DHC test was carried out to estimate the extent of hydrogen embrittlement using Zry-2 tube. The hydrogen redistribution and concentration of hydride at the crack tip were observed under a tensile load of 100-200 N in elastic deformation region, which contributed to propagation of the crack length. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Asahi, Hitoshi; Yagi, Akira; Ueno, Masakatsu

    1998-05-01

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

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

    Directory of Open Access Journals (Sweden)

    Vicente Braz Trindade

    2005-12-01

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

  2. Positron lifetime study in dilute electron irradiated lead based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Moya, G. [Lab. de Physique des Materiaux, 13 Marseille (France); Li, X.H. [D.R.F.M., S.P.2.M., M.P., C.E.N.G., 38 Grenoble (France); Menai, A. [Lab. de Physique des Materiaux, 13 Marseille (France); Kherraz, M. [Lab. de Physique des Materiaux, 13 Marseille (France); Amenzou, H. [Lab. de Physique des Materiaux, 13 Marseille (France); Bernardini, J. [Lab. de Metallurgie, 13 Marseille (France); Moser, P. [D.R.F.M., S.P.2.M., M.P., C.E.N.G., 38 Grenoble (France)

    1995-06-01

    The recovery of defects in two dilute solute-lead based alloys (Pb-Au, Pb-Cd) has been followed by positron lifetime measurements after a 3 MeV electron irradiation at 20 K. Two distinct isochronal annealing stages, the first centred at about 150 K and the other around 275 K, are to be observed as exactly the same in both the pure Pb and dilute alloys but the vacancy clustering over the second stage seen in lead and Pb-Au is completely suppressed in the Pb-Cd alloy. The results are discussed in terms of a high interaction between the cadmium atoms and vacancies in agreement with a probable presence of atomic excitons. (orig.)

  3. Positron lifetime study in dilute electron irradiated lead based alloys

    International Nuclear Information System (INIS)

    The recovery of defects in two dilute solute-lead based alloys (Pb-Au, Pb-Cd) has been followed by positron lifetime measurements after a 3 MeV electron irradiation at 20 K. Two distinct isochronal annealing stages, the first centred at about 150 K and the other around 275 K, are to be observed as exactly the same in both the pure Pb and dilute alloys but the vacancy clustering over the second stage seen in lead and Pb-Au is completely suppressed in the Pb-Cd alloy. The results are discussed in terms of a high interaction between the cadmium atoms and vacancies in agreement with a probable presence of atomic excitons. (orig.)

  4. Positron lifetime study in dilute electron irradiated lead based alloys

    International Nuclear Information System (INIS)

    The recovery of defects in two dilute solute-lead based alloys (Pb-Au, Pb-Cd) has been followed by positron lifetime measurements after a 3 MeV electron irradiation at 20 K. Two distinct isochronal annealing stages, the first centered at about 150 K and the other around 275 K, are to be observed as exactly the same in both the pure Pb and dilute alloys but the vacancy clustering over the second stage seen in lead and Pb-Au is completely suppressed in the Pb-Cd alloy. The results are discussed in terms of a high interaction between the cadmium atoms and vacancies in agreement with a probable presence of atomic excitons. (authors). 3 figs., 9 refs

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

    International Nuclear Information System (INIS)

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

  6. Behavior under irradiation of zirconium alloy strips

    International Nuclear Information System (INIS)

    Tests on elementary cells in inconel 718, in zircaloy or zircaloy with inconel springs, have been made so as to evaluate the evolution under irradiation of the stress put on the fuel rod by the cell springs and the evolution of the boss form. That is why, after having given the essential of the results got during the bulk tests on unitary cells, we show more in detail the results given by relaxation and growth tests under neutron flux, realized with materials differing from materials used currently either because of final thermal treatment or because of their proper nature

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jun, E-mail: cjer19841011@163.com [State Key Laboratory of Rolling and Automation, Northeastern University, Shenyang 110819 (China); Lv, Mengyang [School of Materials and Metallurgy, Northeastern University, Shenyang 110819 (China); Tang, Shuai; Liu, Zhenyu; Wang, Guodong [State Key Laboratory of Rolling and Automation, Northeastern University, Shenyang 110819 (China)

    2015-08-15

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2004-08-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-05

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

  12. Austenite grain refinement during load-biased thermal cycling of a Ni49.9Ti50.1 shape memory alloy

    International Nuclear Information System (INIS)

    A near-equiatomic NiTi shape memory alloy was subjected to a variety of thermomechanical treatments including pure thermal cycling and load-biased thermal cycling to investigate microstructural evolution of the material under actuating conditions. In situ and post mortem scanning transmission electron microscopy (STEM) was used to study the effects of stress on the development of defect substructures during cycling through the martensitic transformation. High temperature observations of the austenite phase show rapid accumulation of dislocations and moderate deformation twinning upon thermomechanical cycling. Additionally, TEM-based orientation mapping suggests the emergence of fine crystallites from the original coarse austenite grain structure. A possible mechanism is proposed for the observed grain refinement based on the crystallographic theory of martensite transformation

  13. Pre-irradiation tests on U-Si alloys

    International Nuclear Information System (INIS)

    Pre-irradiation tests of hardness, density, electrical resistivity, and corrosion resistance as well as metallographic and X-ray examinations were undertaken on U-Si core material, which had been co-extruded in Zr--2, in order that the effect of irradiation on alloys in the epsilon range could be assessed. In addition, a study of the epsilonization of arc-melted material was undertaken in order to rain familiarity with the epsilonization process and to obtain information on the corrosion behaviour of epsilonized material. Sheathed U-Si samples in the epsilonized and de-epsilonized conditions have been irradiated in the X-2 loop, with a water temperature of 275oC. The samples have been examined after 250 MWD/Tonne and show no dimensional change. (author)

  14. Cavity formation by hydrogen injection in electron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Thin foil specimens of solution annealed Type 316, 316L and Ti-modified 316L steels, cold worked Type 316 steel, and 1 MV electron-irradiated Type 316 steel are injected with hydrogen ions to 3x1021-9x1021 ions/m2 at 673 K. A number of cavities are formed not only in the matrix but also on grain boundaries in all the solution annealed steels. In the 5% cold worked 316 steel, the cavity number density is more reduced than in the solution annealed 316 steel. Cavity formation in 1 MV electron irradiated regions depends sensitively on electron predamage given prior to hydrogen ion injection. From a comparison of the cavity formations in electron irradiated regions with predamaged structures, it is noted that vacancy clusters interact strongly with hydrogen atoms and that tangled dislocation networks enable hydrogen atom transport to the foil surface. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-11-01

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

  17. Influence of alloying elements on the irradiation hardening and environmental sensitivity of zirconium alloys

    International Nuclear Information System (INIS)

    Ten different alloys of zirconium have been tested with regard to the effect of irradiation on their mechanical properties and their sensitivity to environmentally induced failure. Two different environments were used: iodine vapour and liquid cesium with an addition of 2% cadmium. The neutron dose was 1021n/cm2 (E>1MeV) and the irradiation temperature was about 300 degrees C. All alloy additions increased the irradiation hardening. Especially notable was the large effect of titanium and tin on irradiation hardening. A limited amount of transmission electron microscopy was carried out in order to find an explanation to the effects. The testing in different environments showed that there is no clear correlation between environmental sensitivity and yield stress. For materials of similar yield stress an alloyed material tends to be more sensitive to environmental cracking than a material which only contains oxygen as an impurity. There also seems to be an effect of oxygen on the environmental cracking sensitivity. A material with 910 ppm oxygen was considerably more sensitive to cracking than a material with 470 ppm oxygen despite the fact that the yield stress values differed by only 90 MPa

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

    International Nuclear Information System (INIS)

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

  19. Low temperature irradiation effects on iron boron based amorphous metallic alloys

    International Nuclear Information System (INIS)

    Three Fe-B amorphous alloys (Fe80B20, Fe27Mo2B20 and Fe75B25) and the crystallized Fe3B alloy have been irradiated at the temperature of liquid hydrogen. Electron irradiation and irradiation by 10B fission fragments induce point defects in amorphous alloys. These defects are characterized by an intrinsic resistivity and a formation volume. The threshold energy for the displacement of iron atoms has also been calculated. Irradiation by 235U fission fragments induces some important structural modifications in the amorphous alloys

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Ren, W.

    1996-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  3. Radiation effects in the aluminium alloys irradiated with neutrons

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Crack growth behavior of irradiated austenitic stainless steels in BWR environments

    International Nuclear Information System (INIS)

    Crack growth tests have been performed in boiling water reactor (BWR) environments on Types 304 and 316 stainless steel that were irradiated to fluence levels up to 2.0 x 1021 n cm-2 (E > 1 MeV) at approx. 288 degC in a helium environment. Two waveforms were used in the tests, slow/fast sawtooth and trapezoidal. The cyclic loading was done with rise times between 30 and 1000 s. At the longer rise times, the environmental contributions to the crack growth rate dominate. The trapezoidal waveform essentially represents constant load with periodic unloading and loading. The results indicate significant enhancement of crack growth rates of the irradiated steel in the BWR environment with normal water chemistry. The effects of fluence and hydrogen water chemistry are presented. (author)

  6. Effect of alloying elements on the electronic properties of thin passive films formed on carbon steel, ferritic and austenitic stainless steels in a highly concentrated LiBr solution

    International Nuclear Information System (INIS)

    The influence of alloying elements on the electrochemical and semiconducting properties of thin passive films formed on several steels (carbon steel, ferritic and austenitic stainless steels) has been studied in a highly concentrated lithium bromide (LiBr) solution at 25 °C, by means of potentiodynamic tests and Mott–Schottky analysis. The addition of Cr to carbon steel promoted the formation of a p-type semiconducting region in the passive film. A high Ni content modified the electronic behaviour of highly alloyed austenitic stainless steels. Mo did not modify the electronic structure of the passive films, but reduced the concentration of defects. - Highlights: • The addition of Cr to carbon steel promotes p-type semiconductivity. • Passive films formed on stainless steels are made up of complex spinel oxides. • Ni modifies the electronic behaviour of highly alloyed austenitic stainless steels

  7. Evaluation of austenitic alloys abrasive wear of FeMnAlC system; Avaliacao de desgaste abrasivo de ligas austeniticas do sistema FeMnAlC

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Allan Ribeiro de; Acselrad, Oscar [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Metalurgica e de Materiais. Lab. de Processamento Termomecanico e Engenharia Microestrutural]. E-mail: allariba@metalmat.ufrj.br

    2003-07-01

    Alloys of the FeMnAlC system have been studied as an alternative to stainless steels applications. Such alloys, when solubilized, are non-magnetic and present an austenitic structure that can be modified by thermal treatments. In this way, a large spectrum of mechanical and physical properties can be obtained. They are oxidation-resistant alloys, and by 15 hours aging at 550 deg C mechanical strength can be as high as conventional structural alloy steels. Information concerning the performance of these alloys under wear conditions are still limited. The possibility of application in components exposed to cavitation or abrasive loads, such as pipes, pumps and drilling systems is still a subject for fundamental research, such as the one that is now reported. Samples of a FeMnAlC alloy have been submitted to different thermal processing, leading to microstructures that have been characterized by optical, transmission and atomic force microscopy and by X-ray diffraction. They were subsequently subjected to a micro-abrasion test in which the abrasive wear resistance could be determined. The results have been used to differentiate the performance of different microstructures and allowed also a comparative analysis with the performance of an AISI M2 tool steel. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  10. Microstructure optimization of austenitic Alloy 800H (Fe-21Cr-32Ni)

    International Nuclear Information System (INIS)

    Research highlights: → Presented a synergistic effect of TMP on microstructure and resulted properties. → Used AFM to quantitatively analyze geometry and distribution of GB precipitates. → Correlated GB characters with precipitates to interpret their effects on properties. → Provided evidence of coherent precipitates at coherent Σ3 boundaries. - Abstract: The microstructural evolution, specifically of grain boundaries, precipitates, and dislocations in thermomechanically processed (TMP) Alloy 800H samples was characterized by scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDS), electron backscattered diffraction (EBSD), transmission electron microscopy (TEM), and atomic force microscopy (AFM). The TMP not only significantly increased the fraction of low-Σ coincidence site lattice boundaries, but also introduced nanoscale precipitates in the matrix and altered the distribution of dislocations. Statistical analysis indicates that the morphology and distribution of grain boundary precipitates were dependent on grain boundary types. The microstructure optimization played a synergistic effect on the significantly increased strength with comparable ductility and enhanced intergranular corrosion resistance and creep-fatigue life compared to the as-received samples.

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

    International Nuclear Information System (INIS)

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

  12. Post-irradiation examination of Ti or Nb stabilized austenitic steels irradiated as BN-600 reactor fuel pin claddings up to 87 dpa

    International Nuclear Information System (INIS)

    The results of postirradiation study of fuel pins with claddings fabricated from the 16Cr-15Ni-3Mo-Nb (EI-847), 16Cr-15Ni-3Mo-Nb-B (EP-172) and 16Cr-15Ni-2Mo-Ti-V-B (ChS-68) austenitic stainless steels in 20% cold-work condition are given. All fuel pins after irradiation in the BN-600 reactor to peak burn up of 11.6% (displacement dose of 83 dpa) and remained its tightness. At the same time, a number of fuel pins have failed during low-load handling in hot cells. Tensile mechanical tests revealed a drastic decrease in strength and a severe embrittlement of the cladding material taken from some parts of fuel pins. For these parts numerous deep microcracks at the inner surface of pin cladding have been observed. Locations of the maximum cladding property degradation coincides with locations of the peak diameter increase and peak swelling. The effects of high swelling and radiation-induced segregation on mechanical properties and corrosion resistance of the fuel pin cladding are discussed. (author)

  13. Irradiation-induced patterning in dilute Cu–Fe alloys

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Radiation defects formed in ion-irradiated 316L stainless steel model alloys with different Si additions

    International Nuclear Information System (INIS)

    The 304/316 series of austenite stainless steels are used in light water reactors as structural materials. As a result of the high temperatures and neutron irradiation in reactor, dislocation defects will form in stainless steel, causing an increase in the hardness and a decrease in the ductility of the material. In this work, high purity 316L stainless steel model alloys with three different Si contents were ion irradiated at 290°C or 400°C to investigate the black dot and Frank loop formation mechanism influenced by Si addition. Black dot defect formation mainly occurs at 290°C. It is Frank loop in nature with its formation not affected by Si addition. Frank loop is the main defect at 400°C, and both loop density and the average size are substantially suppressed by Si addition. This may be caused by silicon’s role in enhancing effective vacancy diffusivity and thus promoting recombination. The trend of irradiation hardening measured verses temperature matches the microstructure observed. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Modification of mechanical properties and microstructure of Ni-Cr-base alloy by continuous electron irradiation

    International Nuclear Information System (INIS)

    Using the methods of transmission and scanning electron microscopy and X-ray structure analysis investigation of 40CrNiAl alloy structure-phase state after different conditions of thermomechanical treatment (TMT) and electron irradiation is carried out. Correlation of microstructure parameters of irradiated alloy with its mechanical properties is ascertained as well as morphology of structural and phase transformations in alloy at continuous electron irradiation. Simultaneous increasing of strength characteristics and plasticity of 40CrNiAl alloy after certain conditions of TMT and electron irradiation is find out, the reasons of the phenomenon is analyzed. The scientifically-based schemes of 40CrNiAl alloy TMT are developed and choice of electron irradiation conditions for optimization of its mechanical properties is substantiated

  1. Correlation between physical and mechanical properties changes of austenitic steel ChS-68 under high dose irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ershova, O.V.; Shcherbakov, E.N.; Evseev, M.V.; Shihkalev, V.S.; Kozlov, A.V. [FSUE, Institute of Nuclear Materials, Zarechney, Sverdlovsk (Russian Federation); Garner, F. [Pacific Northwest National Laboratory, P.O. Box 999, Richland WA, AK 99352 (United States)

    2007-07-01

    Full text of publication follows: It is well known that void swelling at high levels exerts significant influence on physical, mechanical and creep properties of austenitic steels. For many fusion or fission reactor concepts it is desirable not only to characterize these relationships but also to develop nondestructive measurements to measure swelling without removing components from the reactor. Previous studies at this institute have shown that swelling can be estimated using changes in elastic moduli via ultrasonic techniques and electrical resistivity via electro-resistive methods. In this study we examined two pin claddings of ChS-68 (Fe-16Cr-15Ni-2Mo-2Mn-Ti-Si irradiated at somewhat different dpa rates in the high-flux BN-600 fast reactor, with temperatures ranging from 370-590 deg. C to maximum doses of 69 and 78 dpa. After removing the fuel, ring specimens were cut and used to conduct tensile tests using a standardized ring-pull test. Changes in density, elastic moduli and electrical resistivity were performed prior to tensile testing. Maximum swelling levels in the two pins reached {approx}7 and 12%, with strong consequences observed in mechanical properties. At the higher swelling level there was a total loss of ductility over a significant middle portion of the pin. In both the lower swelling and higher swelling pins there was a clear correlation between the local swelling along the pin length with declining ultimate strength and total elongation, providing clear evidence of void-induced embrittlement. Changes in electrical resistivity and elastic moduli correlated well with predictions based on void swelling at lower irradiation temperatures where precipitates were not a dominant part of the radiation-induced microstructure. At higher temperatures large precipitates of Ni-rich radiation-stable phases are a large portion of the microstructure and void-based predictions of elastic moduli and electrical resistivity do not agree well with the measurements

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

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, M.L.; Edwards, D.J. [Pacific Northwest Laboratory, Richland, WA (United States); Toloczko, M.B. [Univ. of California, Santa Barbara, CA (United States)] [and others

    1995-04-01

    This work was performed to determine whether shear punch and tensile data obtained on neutron irradiated aluminum alloys exhibited the same type of relationship as had been seen in other work and to assess the validity of extrapolating the results to proton-irradiated alloys. This work was also meant to be the first of a series of similar test matrices designed to determine whether the shear punch/tensile relationship varied or was the same for different alloy classes.

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

    International Nuclear Information System (INIS)

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

  4. Multi-scale approach of plasticity mechanisms in irradiated austenitic steels

    International Nuclear Information System (INIS)

    The plasticity in irradiated metals is characterized by the localization of the deformation in clear bands, defect free, formed by the dislocation passage. We investigated the clear band formation thanks to a multi-scale approach. Molecular dynamics simulations show that screw dislocations mainly un-fault and absorb the defects as helical turns, are strongly pinned by the helical turns and are remitted in new glide planes when they unpin whereas edge dislocations mainly shear the defects for moderate stresses and can drag the helical turns. The interaction mechanisms were implemented into the discrete dislocation dynamics code in order to study the clear band formation at the micron scale. As dislocations are issued from grain boundaries, we consider a dislocation source located on a box border that emits dislocations when the dislocation nucleation stress is reached. The hardening was seen mainly due to the screw dislocations that are strongly pinned by helical turns. Edge dislocations are less pinned and glide on long distances, letting long screw dislocation segments. As more dislocations are emitted, screw dislocation pile-ups form and this permits the unpinning of screw dislocations. They unpin by activating dislocation segments in new glide planes, which broadens the clear band. When the segments activate, they create edge parts that sweep the screw dislocation lines by dragging away the super-jogs towards the box borders where they accumulate, which clears the band. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

  7. Developments in austenitic steels containing manganese

    International Nuclear Information System (INIS)

    Two broad categories of austenitic steels are considered in this review: (a) alloys based on the Fe-Mn-C system, typified by austenitic wear resistant (Hadfield) steels and (b) alloys based on the Fe-Mn-Cr system, typified by austenitic corrosion resistant steels. Advances made in recent years in understanding and improving the relevant properties and manufacturing methods of these steels are critically appraised. The development of austenitic manganese bearing high technology steels for fusion reactor and other non-magnetic applications, as well as those that can be used in cryogenic structures, is also considered. (author)

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

    International Nuclear Information System (INIS)

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

  9. Resistance of (Fe, Ni)3V long-range-ordered alloys to neutron and ion irradiation

    International Nuclear Information System (INIS)

    A series of (Fe, Ni)3V long-range-ordered alloys were irradiated with neutrons in the Oak Ridge Research Reactor (ORR) and with 4 MeV Ni ions at temperatures above 2500C. The displacement damage levels for the two irradiations were 3.8 and 70 dpa, and the helium levels were 29 and 560 at. ppM, respectively. Irradiation in ORR generally increased the yield strength and lowered the ductility of an LRO alloy, but produced relatively little swelling. The LRO alloys retained their long-range order after ion irradiation below the critical ordering temperature, T/sub c/, and exhibited low swelling. Above T/sub c/ the alloys were disordered and showed greater swelling. Adjustment of alloy composition to prevent sigma phase formation reduced swelling

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

    International Nuclear Information System (INIS)

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

  11. Recovery of the electrical resistivity in electron-irradiated, concentrated silver-zinc alloys

    International Nuclear Information System (INIS)

    Silver and silver-zinc alloys with 5, 10, 20, 25 and 30 at.% Zn were irradiated with 3 MeV electrons at 4.7 K and isochronally annealed up to 573 K. The initial rate of resistivity increase with irradiation dose increases linearly with the Zn concentration. Stage I is strongly suppressed in the concentrated alloys. The first pronounced recovery stage is stage III. This is in agreement with the model proposed by Halbwachs, Hillairet and coworkers that in AgZn the vacancies are more mobile than the interstitials The annihilating vacancies change the short-range order of the alloys reducing by this the resistivity far below the value prior to the irradiation. This effect is independent of the irradiation dose, indicating a reaction of second order in stage III. Above 400 K the resistivity of the alloys increases drastically when thermal vacancies contribute in the disordering of the alloys. (author)

  12. Recovery of the electrical resistivity in electron-irradiated, concentrated silver-zinc alloys

    Energy Technology Data Exchange (ETDEWEB)

    Vaessen, P.; Lengeler, B.; Schilling, W.

    1984-05-01

    Silver and silver-zinc alloys with 5, 10, 20, 25 and 30 at.% Zn were irradiated with 3 MeV electrons at 4.7 K and isochronally annealed up to 573 K. The initial rate of resistivity increase with irradiation dose increases linearly with the Zn concentration. Stage I is strongly suppressed in the concentrated alloys. The first pronounced recovery stage is stage III. This is in agreement with the model proposed by Halbwachs, Hillairet and coworkers that in AgZn the vacancies are more mobile than the interstitials The annihilating vacancies change the short-range order of the alloys reducing by this the resistivity far below the value prior to the irradiation. This effect is independent of the irradiation dose, indicating a reaction of second order in stage III. Above 400 K the resistivity of the alloys increases drastically when thermal vacancies contribute in the disordering of the alloys.

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

    International Nuclear Information System (INIS)

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

  14. Mechanical property and conductivity changes in several copper alloys after 13.5 dpa neutron irradiation

    International Nuclear Information System (INIS)

    A scoping experiment in which 25 different copper materials of 17 alloy compositions were irradiated to approx.13.5 dpa approx.4000C in a fast reactor is described. The materials include rapidly solidified (RS) alloys, with and without oxide dispersion strengthening, as well as conventionally processed alloys. Immersion density (swelling), electrical conductivity (which can be related to thermal conductivity), and yield stress and ductility by miniature disk bend testing have been measured before and after irradiation. It was found, in general, that the Rs alloys are stable under irradiation to 13.5 dpa, showing small conductivity changes and little or no swelling. Reduction of strength and ductility, in post-irradiation tests at the irradiation temperature, are not generally observed. Some conventionally processed alloys also performed well, although irradiation softening and swelling of several percent were observed in some cases, and pure copper swelled in excess of 5%. It is concluded that a number of copper alloys should receive further study, and that higher dose irradiations will be required to establish the limits of swelling suppression in these alloys

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

    International Nuclear Information System (INIS)

    This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x1020 n/cm2 at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranular failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  17. Study of a type AISI 321 austenitic stainless steel with niobium additions, submitted to thermal and mechanical treatments and to fast neutron irradiation

    International Nuclear Information System (INIS)

    The strengthening mechanisms and improved corrosion-and swelling resistance of austenitic Ni-Fe-Cr stainless steel by Nb additions are properties of interest in Nuclear Technology. In this work, Nb additions were made in type 321 stainless steel for metallography, microhardness, electrical resistivity and radiation damage studies. The samples were fabricated in induction furnace on water cooled melting-pot in argon atmosphere. This work presents the results of experimental measurements from which an attempt is made to analyse the effects of Nb additions acting as microalloying element and of thermal and mechanical process (cold work in present work) on the microstructure, micorhardness and electrical resistivity properties of the 11%Ni-70%Fe-17%Cr austenitic stainless steel. The study of this properties before, during and after irradiation with fast neutrons, showed: - for the original composition of type 321 stainless steel the radiation damage peak is around 4950C; - the radiation damage peak for the composition with 0.05Wt.% of Nb addition is around 5000C; - the radiation damage peak for the composition with 0.1Wt.% of Nb addition is around 5650C. Results of vacancies supersaturation are present in the sense to contribute to the void formation studies in metals during irradiation with high energy particles. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  19. An effect of gamma-neutron irradiation and oxygen admixtures on interaction of hydrogen with austenitic 18Cr-10Ni-Ti steel

    International Nuclear Information System (INIS)

    Experimental data on permeability and diffusion of hydrogen through austenitic 18Cr-10Ni-0.65Ti steel with 0.08 and 0.12 wt.% of carbon have been obtained with special equipment designed and installed at the research reactor IVV-2M. It has been shown that parameters of hydrogen isotope transfer in the steel increased substantially during irradiation by fast neutrons of flux density ffast18 n m-2 per s, fluence F=9.5 x 1024 n m-2 (E>0.1 MeV) and absorbed gamma dose 3.6 W g-1 at temperatures of 573-1073 K. Under these irradiation conditions, even a small oxygen admixture content in hydrogen allowed creation of reliable surface barriers on the steel, which essentially decreased hydrogen permeability. At the same time, the physico-mechanical properties of the steel under study changed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-02-01

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

  5. Creep-rupture performance of 0.07C-23Cr-45Ni-6W-Ti,Nb austenitic alloy (HR6W) tubes

    Energy Technology Data Exchange (ETDEWEB)

    Evans, Neal D [ORNL; Shingledecker, John P [ORNL

    2010-01-01

    A series of base metal and cross-weld creep-rupture tests were conducted on the advanced austenitic alloy, HR6W, to evaluate the material for use at advanced ultrasupercritical (A-USC) steam conditions. Creep deformation and rupture were evaluated by traditional methods and data were compared with other studies to evaluate the creep response of the material. Optical and scanning electron microscopy revealed changes in failure mode and precipitation behavior. Thermodynamic predictions of phase stability were conducted and the results were compared with the experimental data. This research confirmed the important role of W and the precipitation of laves phase in the alloy system, but a direct relationship between laves phase content and creep strength was not observed. Furthermore, Cr content was investigated as an additional factor which may be important in the microstructural stability of the alloy which had not been previously considered. Finally, when compared to commercially available stainless steels, this heat of HR6W showed no creep strength advantage for A-USC application.

  6. Creep-rupture performance of 0.07C-23Cr-45Ni-6W-Ti,Nb austenitic alloy (HR6W) tubes

    Energy Technology Data Exchange (ETDEWEB)

    Shingledecker, J.P., E-mail: jshingledecker@epri.co [Electric Power Research Institute, Charlotte, NC (United States); Oak Ridge National Laboratory, Oak Ridge, TN (United States); Evans, N.D. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); University of Tennessee, Knoxville, TN (United States)

    2010-06-15

    A series of base metal and cross-weld creep-rupture tests were conducted on the advanced austenitic alloy, HR6W, to evaluate the material for use at advanced ultrasupercritical (A-USC) steam conditions. Creep deformation and rupture were evaluated by traditional methods and data were compared with other studies to evaluate the creep response of the material. Optical and scanning electron microscopy revealed changes in failure mode and precipitation behavior. Thermodynamic predictions of phase stability were conducted and the results were compared with the experimental data. This research confirmed the important role of W and the precipitation of laves phase in the alloy system, but a direct relationship between laves phase content and creep strength was not observed. Furthermore, Cr content was investigated as an additional factor which may be important in the microstructural stability of the alloy which had not been previously considered. Finally, when compared to commercially available stainless steels, this heat of HR6W showed no creep strength advantage for A-USC application.

  7. Creep-rupture performance of 0.07C-23Cr-45Ni-6W-Ti,Nb austenitic alloy (HR6W) tubes

    International Nuclear Information System (INIS)

    A series of base metal and cross-weld creep-rupture tests were conducted on the advanced austenitic alloy, HR6W, to evaluate the material for use at advanced ultrasupercritical (A-USC) steam conditions. Creep deformation and rupture were evaluated by traditional methods and data were compared with other studies to evaluate the creep response of the material. Optical and scanning electron microscopy revealed changes in failure mode and precipitation behavior. Thermodynamic predictions of phase stability were conducted and the results were compared with the experimental data. This research confirmed the important role of W and the precipitation of laves phase in the alloy system, but a direct relationship between laves phase content and creep strength was not observed. Furthermore, Cr content was investigated as an additional factor which may be important in the microstructural stability of the alloy which had not been previously considered. Finally, when compared to commercially available stainless steels, this heat of HR6W showed no creep strength advantage for A-USC application.

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

    International Nuclear Information System (INIS)

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

  9. Influence of the Amount of Master Alloy on the Properties of Austenitic Stainless Steel AISI 316L Powder Sintered in Hydrogen

    Directory of Open Access Journals (Sweden)

    Mateusz Skaloń

    2012-01-01

    Full Text Available AISI 316L austenitic stainless steel powder was modified with four different amounts of boron (0.1; 0.2; 0.3; 0.4 of wt. % in the form of MasterAlloy micro-powder, and was sintered in a pure dry hydrogen atmosphere in order to obtain high density sintered samples characterized by a thickened non-porous surface layer. We investigated the influence of the amount of boron on: density, hardness, grain microhardness, porosity, microstructure and surface quality. The study revealed that it is possible by a conventional compacting and sintering process to obtain near full-density sintered samples with a non-porous superficial layer without boride precipitations.

  10. Separation of radiation defects in Ni and Ni-C alloys under electron and neutron irradiation

    Science.gov (United States)

    Arbuzov, S. E.; Danilov, V. L.; Goshchitskii, B. N.; Kar'kin, A. E.; Parkhomenko, V. D.

    2016-02-01

    Complex investigations of radiation damage of Ni and Ni- 880 at. ppm C alloy under electron and neutron irradiation in the region of room temperature hardened and deformed state. In pure nickel, with the deformation microstructure, both in electron and in the neutron irradiation is observed separation of radiation-induced defects. When electron irradiation in the alloy Ni-C separation effect is observed, and when neutron irradiation there is no. This is due to the interaction of carbon atoms with radiation defects. The main sinks for radiation-induced defects are the areas with a high concentration of defects in cascades of atomic displacements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.

    1996-04-01

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

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

    OpenAIRE

    Sheakley, Brian J.

    2000-01-01

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

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

    OpenAIRE

    山田 弘一; 河村 弘

    2006-01-01

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

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

    OpenAIRE

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

    2014-01-01

    The interaction between an ultra-fast pulse laser and a material's surface has become a research hotspot in recent years. Micromachining of titanium alloy with an ultra-fast pulse laser is a very important research direction, and it has very important theoretical significance and application value in investigating the ablation threshold of titanium alloy irradiated by ultra-fast pulse lasers. Irradiated by a picosecond pulse laser with wavelengths of 1064 nm and 532 nm, the surface morphology...

  15. Irradiation induced stress relaxation and high temperature deformation behavior of neutron irradiated Ti based shape memory alloys

    International Nuclear Information System (INIS)

    Several tools using Ti based shape memory alloys (SMA) such as SMA coupler, connector, jack system and in-vacuum gate valve, have been developed to promote the remote maintenance and the quick replacement technology for fusion core parts. Recently, irradiation induced stress relaxation (IISR) has become a concern for components of the fusion core. IISR may be a severe problem for SMAs as well as the structural materials in the fusion reactor. The IISR of TiNi SMA and TiPd high temperature shape memory alloys (HTSMA), which have both transformation temperatures and working temperatures 400 K higher than those of TiNi alloys, may be controlled by the migration of vacancies rather than interstitials. This mechanism facilitates restoration of the damaged state to normal state under irradiation. TiPd HTSMAs may be used to fabricate irradiation-resistant shape memory devices for temperatures up to 800 K if proper heat treatments can be developed. ((orig.))

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-14

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

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

    Science.gov (United States)

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

    2015-10-01

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

  19. Restoration properties of neutron irradiated Ti-Ni shape memory alloys

    International Nuclear Information System (INIS)

    Transformation properties and deformation behavior of Ti-Ni shape memory alloys which were irradiated at 323 and 520 K up to a maximum fast neutron fluence of 1025 m-2 and subsequently annealed above 523 K, were examined by electrical resistance measurements and tensile tests. When irradiation was performed at 323 K, Ms temperature of irradiated specimens abruptly decreased at a dose over 10-2 dpa. This shows that the irradiation has a great influence on transformation properties of specimens. After post-irradiation annealing above 523 K, the Ms temperature of specimens which were irradiated with a dose of 10-1 dpa, increased to that of unirradiated ones. When irradiation was performed at 520 K, the decrease in Ms temperature was negligibly small regardless of the magnitude of damage. It is clear that at irradiation temperature of 520 K the irradiation has no influence on transformation properties of Ti-Ni alloys. In the Ti-Ni alloys two conflicting processes take place during irradiation: disordering and ordering. The migration of vacancies is enhanced by thermal activation and ordering becomes predominant over the disordering and restoration phenomena occur. The phenomena can be described as a function of temperature, displacement and displacement rate by the theory of order-disorder transformation under irradiation. It is confirmed that the threshold temperature at which the restoration phenomena take place is about 520 K. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Institute, St. Petersburg (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

    The objective of this study is to evaluate the properties of several copper alloys following fission reactor irradiation at ITER-relevant temperatures of 80 to 200{degrees}C. This study provides some of the data needed for the ITER research and development Task T213. These low temperature irradiations caused significant radiation hardening and a dramatic decrease in the work hardening ability of copper and copper alloys. The uniform elongation was higher at 200{degree}C compared to 100{degree}C, but still remained below 1% for most of the copper alloys.

  3. Enhanced ion irradiation resistance of bulk nanocrystalline TiNi alloy

    Energy Technology Data Exchange (ETDEWEB)

    Kilmametov, A.R. [Institute for Nanotechnology, Forschungszentrum Karlsruhe GmbH, Postfach 3640, 76021 Karlsruhe, Buchig (Germany); Institute of Physics of Advanced Materials, Ufa State Aviation Technical University, 12 K. Marx Street, 450000 Ufa (Russian Federation)], E-mail: ascar2@yandex.ru; Gunderov, D.V.; Valiev, R.Z. [Institute of Physics of Advanced Materials, Ufa State Aviation Technical University, 12 K. Marx Street, 450000 Ufa (Russian Federation); Balogh, A.G. [Institute for Materials Science, Technische Universitaet Darmstadt, Petersenstr. 23, 64287 Darmstadt (Germany); Hahn, H. [Institute for Nanotechnology, Forschungszentrum Karlsruhe GmbH, Postfach 3640, 76021 Karlsruhe, Buchig (Germany); Joint Research Laboratory Nanomaterials, Technische Universitaet Darmstadt, Petersenstr. 23, 64287 Darmstadt (Germany)

    2008-11-15

    Bulk ordered nanocrystalline Ti{sub 49.4}Ni{sub 50.6} alloys with a grain size of 23-31 nm were prepared using severe plastic deformation. Nanocrystalline and coarse-grained alloys were subjected to 1.5 MeV Ar{sup +} ion irradiation at room temperature. At the same damage dose the nanocrystalline alloy retained the long-range order while the coarse-grained counterpart was amorphized. In contrast to former irradiation studies, fully dense nanocrystalline materials are used in the present study for the first time.

  4. Defect microstructure in copper alloys irradiated with 750 MeV protons

    DEFF Research Database (Denmark)

    Zinkle, S.J.; Horsewell, A.; Singh, B.N.; Sommer, W.F.

    1994-01-01

    Transmission electron microscopy (TEM) disks of pure copper and solid solution copper alloys containing 5 at% of Al, Mn, or Ni were irradiated with 750 MeV protons to damage levels between 0.4 and 2 displacements per atom (dpa) at irradiation temperatures between 60 and 200 degrees C. The defect...

  5. The influence of pulsed plasma jet irradiation on the mechanical properties of KH16N15M3B and KH18N10T austenitic stainless steels

    International Nuclear Information System (INIS)

    The influence of pulsed jets of hydrogen, helium and mixed hydrogen-helium plasmas with a specific power of 2-3 MW/cm2 on the mechanical properties of austenitic stainless steels of type KH16N15M3B and KH18N10T has been investigated. It was found that the irradiation results in a 1.8 fold increase in the yield point of the above steels and, more importantly, that the elongation values decreased 2.3-2.7 times. It is suggested that such modifications of mechanical properties are caused by the formation of a cellular structure in the subsurface layers of the materials. (orig.)

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

    International Nuclear Information System (INIS)

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al2O3-dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 1024n/m2(E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

  7. Irradiation Tests of Alloy 690 Steam Generator Tube Material of the SMART in HANARO

    International Nuclear Information System (INIS)

    The System-integrated Modular Advanced ReacTor (SMART) is one of the most advanced SMRs. The Korean government decided to obtain the standard design approval on SMART from the Korean licensing authority by 2011. Because the SMART steam generators are located inside the reactor vessel, the degradation of the fracture toughness of the Alloy 690 heat exchanger tube should be clearly determined for a design lifetime neutron fluence. However, the neutron irradiation characteristics of the alloy are barely known. Therefore, an irradiation plan of the Alloy 690 materials to obtain the neutron irradiation characteristics of the alloy using the HANARO irradiation capsules was planned. The target of fast neutron fluence of Alloy 690 was determined to be 1x1018 n/cm2, 1x1019 n/cm2, and 1x1020 n/cm2 (E>1.0 MeV), considering the maximum lifetime neutron fluence of 1.1x1018 n/cm2 of the SMART steam generator. To obtain these neutron fluences, three different irradiation capsules were scheduled and successfully irradiated in the OR5 and CT test holes of the HANARO. The target of irradiation temperature of the specimens was determined as 250 ± 10 .deg. C, considering the operating temperature of 247 .deg. C∼282 .deg. C of the steam generator tube having the highest neutron fluence. Generally, the neutron irradiation degradation effect appears more clearly in a lower temperature. The obtained material properties of the irradiated Alloy 690 specimens will be very valuable to acquire the standard design approval of SMART from the Korean licensing authority

  8. Evidence for a mechanism of swelling variation with composition in irradiated Fe-Cr-Ni alloys

    International Nuclear Information System (INIS)

    Irradiations with 4 MeV Ni ions and 200 to 400 keV He ions were carried out on two alloys, Fe-15Cr-15Ni and Fe-15Cr-35Ni, at 675 C and doses up to 84 dpa. Both dual-ion irradiation experiments and sequenced He injection-anneal-Ni irradiations were used. The dual-ion experiment showed that the two alloys exhibited large differences in microstructural development, with the low nickel alloy having significantly greater swelling. The injection-anneal-irradiation experiment was designed to test the hypothesis, suggested by our earlier work, that the lower swelling of the high nickel alloy may result from a larger critical radius/critical number of gas atoms required to achieve bias driven swelling. This experiment provided a direct measurement of these critical quantities by the induction of a bimodal cavity size distribution. The measurement gave minimum critical radii of about 5 nm for the high nickel alloy and < 0.5 nm for the low nickel alloy, values consistent with the hypothesized mechanism. The basis of this difference in critical quantities was further investigated. Evidence suggests that interstitial absorption at interstitial type dislocation loops is significantly more difficult in the high nickel alloy. (author)

  9. Investigation on corrosion resistance of Hastelloy N alloy after He+ ion irradiation

    International Nuclear Information System (INIS)

    Background: The invalidation problems of irradiated Hastelloy N alloy caused by high temperature, intense irradiation and severe corrosion are the key factors to the service life of structural materials of molten salt reactor (MSR). Purpose: The aim is to investigate the effect of absorbed dose on the corrosion resistance of Hastelloy N alloy. Methods: Hastelloy N alloy was irradiated by 4.5-MeV He+ ions, and the absorbed doses were 0 He+·cm-2, 1x1015 He+·cm-2, 5×1015 He+·cm-2 and 1×1016 He+·cm-2 respectively. The virgin and irradiated specimens were immersed into molten fluoride salts at 700℃ for 300 h. Then the corroded specimens were imaged by scanning electron microscopy and analyzed by synchrotron radiation microbeam X-ray fluorescence (μ-XRF). Results: The weight-loss results showed that the corrosion generally correlated with the absorbed dose of the alloy. The μ-XRF results indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Conclusion: The density of dislocations of Hastelloy N alloy increased with the absorbed dose, which acted as quick paths for Cr element diffusion, and the diffusion of element Cr out of matrix became easier. Finally became weak of the corrosion resistance of Hastelloy N alloy. (authors)

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

    Science.gov (United States)

    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.

  11. The mechanical properties of 316L/304L stainless steels, Alloy 718 and Mod 9Cr-1Mo after irradiation in a spallation environment

    Science.gov (United States)

    Maloy, S. A.; James, M. R.; Willcutt, G.; Sommer, W. F.; Sokolov, M.; Snead, L. L.; Hamilton, M. L.; Garner, F.

    2001-07-01

    The Accelerator Production of Tritium (APT) project proposes to use a 1.0 GeV, 100 mA proton beam to produce neutrons via spallation reactions in a tungsten target. The neutrons are multiplied and moderated in a lead/aluminum/water blanket and then captured in 3He to form tritium. The materials in the target and blanket region are exposed to protons and neutrons with energies into the GeV range. The effect of irradiation on the tensile and fracture toughness properties of candidate APT materials, 316L and 304L stainless steel (annealed), modified (Mod) 9Cr-1Mo steel, and Alloy 718 (precipitation hardened), was measured on tensile and fracture toughness specimens irradiated at the Los Alamos Neutron Science Center accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The irradiation temperatures ranged from 50°C to 164°C, prototypic of those expected in the APT target/blanket. The maximum achieved proton fluence was 4.5×10 21 p/ cm2 for the materials in the center of the beam. This maximum exposure translates to a dpa of 12 and the generation of 10 000 appm H and 1000 appm He for the Type 304L stainless steel tensile specimens. Specimens were tested at the irradiation temperature of 50-164°C. Less than 1 dpa of exposure reduced the uniform elongation of the Alloy 718 (precipitation hardened) and Mod 9Cr-1Mo to less than 2%. This same dose reduced the fracture toughness by 50%. Approximately 4 dpa of exposure was required to reduce the uniform elongation of the austenitic stainless steels (304L and 316L) to less than 2%. The yield stress of the austenitic steels increased to more than twice its non-irradiated value after less than 1 dpa. The fracture toughness reduced significantly by 4 dpa to ˜100 MPa m 1/2. These results are discussed and compared with results of similar materials irradiated in fission reactor environments.

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

    Science.gov (United States)

    Chen, H. C.; Li, D. H.; Lui, R. D.; Huang, H. F.; Li, J. J.; Lei, G. H.; Huang, Q.; Bao, L. M.; Yan, L.; Zhou, X. T.; Zhu, Z. Y.

    2016-06-01

    Under Xe ion irradiation, the microstructural evolution of a nickel based alloy, Hastelloy N (US N10003), was studied. The intrinsic dislocations are decorated with irradiation induced interstitial loops and/or clusters. Moreover, the intrinsic dislocations density reduces as the irradiation damage increases. The disappearance of the intrinsic dislocations is ascribed to the dislocations climb to the free surface by the absorption of interstitials under the ion irradiation. Moreover, the in situ annealing experiment reveals that the small interstitial loops and/or clusters induced by the ion irradiation are stable below 600 °C.

  13. On the precipitation of magnesium silicide in irradiated aluminum-magnesium alloys

    International Nuclear Information System (INIS)

    Thermal neutron irradiation of aluminium or its alloys causes the production of silicon by transmutation. In aluminium-magnesium alloys, the transmutation-produced silicon reacts with magnesium and forms small precipitates. The precipitation in irradiated Al-Mg alloys is similar to the early stage of aging in thermally treated Al-Mg-Si alloys. This study evidences the simultaneous generation of two crystallographically different precipitate types. On the basis of electron diffraction patterns, unit cell parameters are derived and compared with structures found in thermally aged alloys. One of the two precipitate types has an Mg2Si composition, while the other is an Al-Si-Mg intermetallic compound with high aluminium and silicon but low magnesium content. The formation of magnesium poor precipitates is important since it indicates that the threshold neutron fluence for grain boundary precipitation of silicon may be much higher than estimated in the past

  14. Fatigue performance of copper and copper alloys before and after irradiation with fission neutrons

    International Nuclear Information System (INIS)

    The fatigue performance of pure copper of the oxygen free, high conductivity (OFHC) grade and two copper alloys (CuCrZr and CuAl-25) was investigated. Mechanical testing and microstructural analysis were carried out to establish the fatigue life of these materials in the unirradiated and irradiated states. The present report provides the first information on the ability of these copper alloys to perform under cyclic loading conditions when they have undergone significant irradiation exposure. Fatigue specimens of OFHC-Cu, CuCrZr and CuAl-25 were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of ∼2.5 x 1017 n/m2s (E > 1 MeV) to fluence levels of 1.5 - 2.5 x 1024 n/m2s (E > 1 MeV) at ∼47 and 100 deg. C. Specimens irradiated at 47 deg. C were fatigue tested at 22 deg. C, whereas those irradiated at 100 deg. C were tested at the irradiation temperature. The major conclusion of the present work is that although irradiation causes significant hardening of copper and copper alloys, it does not appear to be a problem for the fatigue life of these materials. In fact, the present experimental results clearly demonstrate that the fatigue performance of the irradiated CuAl-25 alloy is considerably better in the irradiated than that in the unirradiated state tested both at 22 and 100 deg. C. This improvement, however, is not so significant in the case of the irradiated OFHC-copper and CuCrZr alloy tested at 22 deg. C. These conclusions are supported by the microstructural observations and cyclic hardening experiments. (au) 4 tabs., 26 ills., 10 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Tensile properties at cryogenic temperatures of austenitic stainless steel and high manganese steel irradiated by fast neutrons at 5K

    International Nuclear Information System (INIS)

    The tensile tests of experimental SUS304 series stainless steels which have nitrogen contents of 0.003, 0.132 and 0.230 were carried out at liquid helium temperature after fast neutron irradiation at 5K, without any intermediate warm-up. The 0.2% yield strength at liquid helium temperature after the irradiation increased for the no-additional-nitrogen stainless steel and, however, the yield strength became independent of the irradiation with nitrogen content. The tensile strength and martensite volume after the irradiation were not affected by the nitrogen content. A high manganese steel (25 Mn-Cr) was tested at liquid helium temperature after the irradiation at 5K, warm-up to liquid nitrogen temperature. The tensile strength and the 0.2% yield strength have a tendency to increase with irradiation. The effects of irradiation on tensile properties are discussed. (author)

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

    International Nuclear Information System (INIS)

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

  4. Microstructural development in irradiated U-7Mo/6061 Al alloy matrix dispersion fuel

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1996-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

    Christine Eckstein; Heinz- Joachim Spies; Jochen Albrecht

    2004-01-01

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

  7. The role of microchemical and microstructural effects in the IASCC of high purity austenitic stainless steels

    International Nuclear Information System (INIS)

    The role of chromium depletion and radiation hardening on the irradiation assisted stress corrosion cracking in CERT tests in high purity 288 degrees C water following proton irradiation at either 400 degrees C or 200 degrees C has been examined using ultra high purity 304L stainless steel and austenitic Fe/xCr/24Ni (x=15, 20, 24) alloys. No intergranular cracking was found in any of the irradiated 254 wt% nickel alloys after CERT tests in 2 ppm O2 water at 288 degrees C, with 0.5, 1.0 or 3.2 μS/cm conductivity, while the UHP 304L alloy cracked extensively. Since the 24 wt% Ni alloys experienced severe grain boundary Cr depletion (from 6.3 at% to 13 at% below bulk), these results suggest that Ni improves the resistance of the irradiated alloys to cracking. Conversely, these results also show little correlation with grain boundary Cr depletion. Cracking of the UHP 304L alloy still occurred, although to a lesser extent, when the sample was irradiated at 200 degrees C where radiation induced segregation was expected to be significantly suppressed. This indicated that radiation hardening may play a role in IASCC in high temperature water

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

    Science.gov (United States)

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

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 3.30 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm-3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Science.gov (United States)

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

    2013-08-01

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

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

    International Nuclear Information System (INIS)

    Helium embrittlement resistance of Fe-25% Ni-15% Cr austenitic alloys with various MX (M=V, Ti, Nb, Zr; X=C, N) stabilizers was compared through post helium implantation creep testing at 923 K. While significant deterioration by helium in terms of creep rupture time and elongation occurred for all materials investigated, the suppression of the deterioration, especially in rupture time, was discerned for the materials in which semi-coherent MC (M=Ti, Ti+Nb, V+Ti) particles were distributed at high density. The material which contains the incoherent M23C6 as predominant precipitates seems to be less degraded by helium than those containing the MXs (M=Zr, V; X=C, N), if compared at the same number density of precipitates. Therefore, it is suggested that the high density dispersion of incoherent M23C6 as well as semi-coherent Ti containing MC particles would be beneficial in reducing the detrimental helium influences on mechanical properties. (orig.)

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

    International Nuclear Information System (INIS)

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

  14. Neutron diffraction analysis of Cr–Ni–Mo–Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    International Nuclear Information System (INIS)

    A quantitative assessment is presented of the dislocation density and relative fractions of edge and screw dislocations in reactor-steel samples 16Cr–15Ni–3Mo–1Ti subjected to preliminary cold deformation by rolling and subsequent fast neutron irradiation using neutron diffraction analysis. The Williamson–Hall modified method was used for calculations. It is shown that the fast neutron irradiation leads to a decrease in the density of dislocations that appeared after samples deformation. The applicability of neutron diffraction analysis to the examination of dislocation structure of deformed and irradiated materials is shown

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

    International Nuclear Information System (INIS)

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

  16. Mechanical properties and damage behavior of non-magnetic high manganese austenitic steels

    International Nuclear Information System (INIS)

    Fe-Cr-Mn steels have been considered as materials of structural components for fusion reactor because of their low induced-radio-activity compared with SUS316 stainless steels. It has been expected to develop a non-magnetic steel with a high stability of the austenitic phase and a strong resistance to irradiation environments. For these objectives, a series of the Fe-Cr-Mn steels have been examined by tensile tests and simulation irradiation by electrons. The main alloying compositions of the steels developed are: C:0.02-0.2 wt%, Mn: 15 wt%, Cr: 15-16 wt%, N: 0.2 wt%. These steels were heat-treated at 1323 K for 1 h. The structure of the steels after the heat-treatment was austenite single phase. The yield stress of the steels was 350-450 MPa and the elongation were 55-60%. When the steels of high C and N was electron-irradiated at below 673 K, no voids were nucleated and only small dislocation loops were formed with high density. The austenite phase was also stable during irradiation below 673 K. Thus, newly developed high manganese steels have excellent mechanical proprieties and high irradiation resistance at relatively low temperature. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  18. One-pot synthesis of AuPt alloyed nanoparticles by intense x-ray irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Cheng-Liang; Hsao, Bo-Jun; Lai, Sheng-Feng; Chen, Hsiang-Hsin; Chen, Yi-Yun; Chien, Chia-Chi; Cai, Xiaoqing; Kempson, Ivan M; Hwu, Y [Institute of Physics, Academia Sinica, Nankang, Taipei 11529, Taiwan (China); Chen, Wen-Chang [Department of Chemical and Materials Engineering, National Yulin University of Science and Technology, Douliou, Yunlin 64002, Taiwan (China); Margaritondo, G, E-mail: clwang@phys.sinica.edu.tw, E-mail: phhwu@.sinica.edu.tw [Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2011-02-11

    We synthesized AuPt alloyed nanoparticles in colloidal solution by a one-pot procedure based on synchrotron x-ray irradiation in the presence of PEG (polyethylene glycol). The exclusive presence of alloyed nanoparticles with fcc structure was confirmed by several different experiments including UV-vis spectroscopy, x-ray diffraction (XRD) and transmission electron microscopy (TEM). The composition of the AuPt alloyed nanoparticles can be varied in a continuous fashion by simply varying the feed ratios of Au and Pt precursors. The nanoparticles exhibited colloidal stability and biocompatibility, important for potential applications.

  19. Mechanical properties and microscopic structure of Al-Mg-Li alloys after 14 MeV neutron irradiation

    International Nuclear Information System (INIS)

    Al-4.1 wt% Mg-Li alloys containing 0.6, 1.1 and 1.7 wt% of Li were tested by an Instron type tensile machine, after 14 MeV neutron irradiation at RTNS-II for neutron fluence up to 4x1022 n/m2 at room temperature. No appreciable changes in yield stress, tensile strength and elongation were observed for the irradiated Al-Mg-Li alloys, in contrast with other metals and alloys such as Cu, Mo, Ni, Fe, Au, TZM, ferritic steel and SUS 316, which were irradiated and tested at the same time. Microscopic observations by TEM on the specimens of irradiated alloys showed that there were no irradiation defects in the microstructure. It is concluded that Al-Mg-Li alloys are promising materials for the D-T neutron environment for the fluence levels examined in this study. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Carbon transfer between 2 1/4 Cr 1 Mo alloy and austenitic steels (experiments in anisothermal loops)

    International Nuclear Information System (INIS)

    Studies on carbon transfer between the ferritic steel 2 1/4 Cr 1 Mo and the austenitic steels 316L and 321H have shown that there is not any measurable carbon transfer in the operating conditions of the secondary circuit of PHENIX (475 deg C was the maximal temperature of the 2 1/4 Cr 1 Mo steel). A significant carbon transfer has been observed between the ferritic steel and the 316L steel when the 321H was replaced by the 2 1/4 Cr 1 Mo steel in the same thermohydraulic conditions (the ferritic steel was then used up to 545 deg C). This experiment has demonstrated the importance of the temperature and the initial carbon content of the ferritic steel as parameters in the decarburization process. It appears that decarburization may not be sensitive to the thermohydraulic conditions at least in the range investigated in those experiments. In the other hand the 316L steel is observed to have been carburized, the degree of carburization remaining appreciably constant and independent on the temperature between 400 deg C and 550 deg C

  3. Relation between the punctual defects dynamics and irradiation growth in Zr and its alloys

    International Nuclear Information System (INIS)

    The knowledge of the static properties punctual defects and their interaction with the drains present in the material, such as dislocations and grain borders, is of fundamental importance for the correct prediction of the reactor's components behaviour under irradiation. This work calculates the Zr alloys growth under irradiation and results are compared with the stretching values of the cooling channels measured at the Atucha I nuclear power plant. (Author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1979-01-01

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

  6. Processing of Refractory Metal Alloys for JOYO Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    RF Luther; ME Petrichek

    2006-02-21

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

  7. Processing of Refractory Metal Alloys for JOYO Irradiations

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Białobrzeska B.; Dudziński W.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  11. Low temperature irradiation of FeB amorphous alloys

    International Nuclear Information System (INIS)

    These experiments show that low temperature electron irradiation induce localized defects in the short range order of the amorphous structure. These defects are assumed to be of Frenkel pair type. At low temperature, 2.5 MeV electron irradiation induces an higher concentration of defects in the amorphous than in its crystallized counterpart

  12. Ion irradiation testing and characterization of FeCrAl candidate alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-29

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

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

    International Nuclear Information System (INIS)

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 1021 fissions cm−3 at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 1021 fissions cm−3 and approximately 45% for a fission density of 4.52 × 1021 fissions cm−3 from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements

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

    Science.gov (United States)

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

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  18. Simulation of the nanostructure evolution under irradiation in Fe-C alloys

    Science.gov (United States)

    Jansson, V.; Malerba, L.

    2013-11-01

    Neutron irradiation induces in steels nanostructural changes, which are at the origin of the mechanical degradation that these materials experience during operation in nuclear power plants. Some of these effects can be studied by using as model alloy the iron-carbon system.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Work hardening characteristics of gamma-ray irradiated Al-5356 alloy

    International Nuclear Information System (INIS)

    Effects of γ-irradiation and deformation temperatures on the hardening behavior of Al-5356 alloy have been investigated by means of stress–strain measurements. Wire samples irradiated with different doses (ranging from 500 to 2000 kGy) were strained at different deformation temperatures Tw (ranging from 303 to 523 K) and a constant strain rate of 1.5×10−3 s−1. The effect of γ-irradiation on the work-hardening parameters (WHP): yield stress σy, fracture stress σf, total strain εT and work-hardening coefficient χp of the given alloy was studied at the applied deformation temperature range. The obtained results showed that γ-irradiation exhibited an increase in the WHP of the given alloy while the increase in its deformation temperature showed a reverse effect. The mean activation energy of the deformation process was calculated using an Arrhenius-type relation, and was found to be ∼80 kJ/mole, which is close to that of grain boundary diffusion in aluminum alloys

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  4. Magneto-elastic attenuation in austenitic phase of Ni-Mn-Ga alloy investigated by ultrasonic methods

    Czech Academy of Sciences Publication Activity Database

    Seiner, Hanuš; Bicanová, Lucie; Sedlák, Petr; Landa, Michal; Heller, Luděk; Aaltio, I.

    521-522, - (2009), s. 205-208. ISSN 0921-5093 R&D Projects: GA ČR GA101/06/0768 Institutional research plan: CEZ:AV0Z20760514 Keywords : ultrasonics methods * Shape memory alloys * RUS * magnetoelasticity Subject RIV: BI - Acoustics Impact factor: 1.901, year: 2009 http:// apps .isiknowledge.com/full_record.do?product=WOS&search_mode=GeneralSearch&qid=1&SID=S1446KoaJ84G2G4LchI&page=1&doc=1

  5. Auto-oscillations of temperature and defect density in ordered binary alloys under irradiation

    International Nuclear Information System (INIS)

    A manifestation of antistructural defects created by nuclear irradiation in ordered binary alloys is investigated. Calculations show that the concentration of such defects can be large at typical values of the intensity of irradiation and temperature. The appearance of structural defects can cause instability in a crystal during irradiation. The instability is connected with the acceleration of antistructural defect relaxation due to the heat which releases during this relaxation. The instability leads to the appearance of self-oscillations of the defect density and temperature of a crystals. The manifestation of self-oscillations is investigated

  6. Cast alumina forming austenitic stainless steels

    Science.gov (United States)

    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; 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.

  7. The research of corrosion defects in aluminium alloy SAV-1, irradiated by neutrons

    International Nuclear Information System (INIS)

    The study of corrosion resistance of rod from low aluminium alloy SAV-1 after a long term operating in the nuclear reactor WWR-K and in water pool storage are resulted. The corrosion tests executed with usage of chemical and electrochemical methods of an estimation of a fixed potential and corrosion rate in chlorine solution to the environment on an accelerated mode on samples cut from the top and bottom end of a atomic reactor rod. With application of methods of a volume determination by hydrostatic weighting of corrosion of aluminium alloy SAV-1 irradiated by a different fluency of neutrons. Is showing, that the irradiation decrease periods of passivation and accelerates a rate corrosion of aluminium alloy SAV-1. (author)

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

    DEFF Research Database (Denmark)

    Barlow, P.; Leffers, Torben

    1977-01-01

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

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

    CERN Document Server

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

  11. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

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

    International Nuclear Information System (INIS)

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

  13. The effect of alloyed nitrogen or dissolved nitrate ions on the anodic behaviour of austenitic stainless steel in hydrochloric acid

    International Nuclear Information System (INIS)

    The anodic behaviour of high purity stainless steels, based on a 316L composition, has been studied at room temperature in HCl solutions from 1 to 6 M. For all acid concentrations, the presence of 0.22% nitrogen has little or no effect on the active dissolution kinetics at low over-potentials. The effect on the critical current density for passivation is also small for low HCl concentrations (4.5 M), no passivation occurs and again nitrogen has little effect. However, for HCl concentrations around 4 M nitrogen reversibly impedes active dissolution at a few hundred mA cm-2. The effect does not appear to be an oxide passivation, but is more likely to be due to surface enrichment of nitrogen atoms. Implications for localized corrosion are discussed. An effect similar to that of nitrogen alloying is reproduced on a nitrogen free alloy by adding 2 M NaNO3 to a 4M HCl solution. This effect is distinct from the passivation of salt-covered surfaces and may be preferable to the latter as an explanation of the increase in pitting potential by nitrate additions to NaCl solutions. Passivation under a salt film is retained to explain the passivation of growing pits above the inhibition potential. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R ampersand D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication

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

    Directory of Open Access Journals (Sweden)

    Buxiang Zheng

    2014-02-01

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

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

    International Nuclear Information System (INIS)

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

  20. Austenite formation during intercritical annealing

    Directory of Open Access Journals (Sweden)

    A. Lis

    2008-07-01

    Full Text Available Purpose: of this paper is the effect of the soft annealing of initial microstructure of the 6Mn16 steel on the kinetics of the austenite formation during next intercritical annealing.Design/methodology/approach: Analytical TEM point analysis with EDAX system attached to Philips CM20 was used to evaluate the concentration of Mn, Ni and Cr in the microstructure constituents of the multiphase steel and mainly Bainite- Martensite islands.Findings: The increase in soft annealing time from 1-60 hours at 625°C increases Mn partitioning between ferrite and cementite and new formed austenite and decreases the rate of the austenite formation during next intercritical annealing in the (α+γ temperature range at 700 and 750°C. The general equations for carbide dissolution and austenite formation in intercritical temperature range were established.Research limitations/implications: The final multiphase microstructure can be optimised by changing the time / temperature parameters of the intercritical heating in the (α+γ temperature range.Originality/value: The knowledge of partitioning of alloying elements mainly Mn during soft annealing and intercritical heating is very important to optimise the processing technology of intercritical annealing for a given amount of the austenite.

  1. Evolution of commercial titanium alloys microstructure under neutron and ion irradiation

    International Nuclear Information System (INIS)

    Results of neutron and ion irradiation on microstructure of titanium alloys of V T-9, V T-12, O T4-1 trade-marks are considered. Samples with 8·4·2 mm3 sizes were irradiated by neutrons in wet channel of WWR-K reactor (E>0,1 MeV) under temperatures below than 80 deg C up different influences. Neutron energy is made up 2,35 MeV (flow 6,5·1016 m-2·c-1). Ion irradiation was carried out on impulse accelerator with energy of accelerated ions 60 keV (impulse duration 250 μs, frequency 3-25 Hz). It was defined that phase recrystallization is taking place in the result of neutron irradiation influence on structure of phases in dilute heterogeneous commercial α+β alloys of titanium-aluminium system. The recrystallization has both the quantitative ratio of phase partials and characteristics of fine structure each of it. It was shown that irradiation of V T-14 alloys by ions on the initial stages up to fluence 2·1017 ion/sm2 conduits to increasing of microhardness up to 60 %. 7 refs., 7 figs

  2. Irradiation effect on the precipitation in Fe-Cr model alloys with around 15% of chromium

    International Nuclear Information System (INIS)

    The ferritic-martensitic steels containing around 12% of chromium are considered for nuclear applications. But, under working reactor conditions, they can become brittle because of the precipitation of a new chromium rich phase called α'. To answer this question, we study this phase separation in Fe-Cr (10 to 25%) model alloys under irradiation at 300 C with a weak flux of electron and under thermal annealing at 500 C. When the precipitation of the α' phase occurs, the alloys become harder. We measured the hardening by Vickers testings. The precipitates are detected by small-angle neutron scattering. Analysis of the intensities with a hard sphere model gives the mean precipitate size and density. These parameters obtained that way can explain the hardening. Under irradiation at 300 C, the growth kinetic is very slow - the precipitates remain very small with a typical radius of 7-8 Angstroms - and the density of precipitates rises up 1019 per cm3. On the other hand, when the alloys are annealed at 500 C, the precipitates grow with a coarsening kinetic. Assuming that the only effect of irradiation is to enhance the diffusion, we calculate the precipitation kinetic with the cluster dynamic model. It is able to reproduce the thermal precipitation at 500 C but not the precipitation at 300 C. An other mechanism, induced by a coupling between fluxes of point defects and solute atoms, is clearly relevant under irradiation. The precipitation kinetic observed in the alloys irradiated at 300 C could relate to this mechanism: the negative coupling of fluxes in Fe-Cr alloys could slow down the precipitate growth. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    Science.gov (United States)

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

    2016-02-01

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

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

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, B.N.

    The electrical conductivity of three different types of copper alloys, viz. CuNiBe, CuCrZr and Cu-Al(2)O(3) as well as of pure copper are reported. The alloys have undergone different pre-irradiation heat treatments and have been fission-neutron irradiated up to 0.3 dpa. In some cases post......-irradiation annealing has been carried out. The results are discussed with reference to equivalent Transmission Electron Microscopy results on the microstructure of the materials. The CuNiBe has the lowest conductivity (less than or equal to 55% of that of pure Cu), and Cu-Al(2)O(3) the highest (75-90% of pure Cu). (C...

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

    Science.gov (United States)

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

    2014-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-01

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

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

    Science.gov (United States)

    Zemtsova, N. D.

    2014-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Motta, A.T.; Faldowski, J.A. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering; Howe, L.M. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Okamoto, P.R. [Argonne National Lab., IL (United States)

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

  18. Test matrices for irradiation of Path A Prime Candidate and developmental alloys in FFTF

    International Nuclear Information System (INIS)

    These irradiations are generally intended to evaluate the void swelling resistance of the Path A Prime Candidate Alloy (PCA) at high fluence (>100 dpa) relative to 20%-cold-worked type 316 after side-by-side irradiation in the Fast Flux Test Facility (FFTF), as well as to evaluate the improvements achieved by additional minor compositional variations of the PCA. Further, these irradiations also serve as a low helium base line against which to gage more accurately the effects of higher helium generation for similar irradiation of exactly the same alloys in the High Flux Isotope Reactor (HFIR). An initial set of transmission electron microscopy (TEM) disk specimens was assembled for irradiation at approximately 420, 520, and 6000C to fluences of approximately 15, 45, and 75 dpa. Some TEM specimens were discharged at approximately 9.5 to 15.6 dpa at all temperatures, and additional specimens were then reloaded to achieve fluences well beyond 100 dpa. Tensile specimens were also included in the reload for irradiation in an above-core position at approximately 6000C. The general goals of the experiments are outlined and detailed specimen loadings are described. 9 references, 4 tables

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

    Science.gov (United States)

    Liu, Xiangbing; Wang, Rongshan; Ren, Ai; Jiang, Jing; Xu, Chaoliang; Huang, Ping; Qian, Wangjie; Wu, Yichu; Zhang, Chonghong

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

  2. On fracture toughness decrease in low alloy steel under irradiation

    International Nuclear Information System (INIS)

    Mechanism of fracture toughness decrease under the effect of irradiation in steels 15Kh2MFA and 15Kh3NMFA-A has been investigated. Samples with cracks irradiated in BOR-60 and WWR-M type reactos were tested for eccentric tension at the temperature from - 196 deg C to 100 deg C. A decrease in the value KIcmin under the effect of irradiation, which agrees with the notions on relation of the phenomenon to the relaxation of residual stresses in the crack vertex, is detected. Resumption of KIcmin value can not be attained even in the case of radiation embrittlement elimination by annealing. The value can be increased only by means of loading up to KI > KIcmin in the temperature region of ductile fracture

  3. Voids in neutron-irradiated metals and alloys

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Alloys under irradiation are continuously driven away from equilibrium: Every time an external particle interacts with the atoms in the solid, a perturbation very localized in space and time is produced. Under this external forcing, phase and microstructural evolution depends ultimately on the dynamical interaction between the external perturbation and the internal recovery kinetics of the alloy. We consider the nonequilibrium steady state of an immiscible binary alloy subject to mixing by heavy-ion irradiation. It has been found that the range of the forced atomic relocations taking place during collision cascades plays an important role on the final microstructure: when this range is large enough, it can lead to the spontaneous formation of compositional patterns at the nanometer scale. These results were rationalized in the framework of a continuum model solved by deriving a nonequilibrium thermodynamic potential. Here we derive the nonequilibrium structure factor by including the role of fluctuations. In order to consider an experimentally relevant situation, we perform kinetic Monte Carlo simulations of temperature-controlled heavy-ion irradiation in the immiscible model system Ag-Cu, where we obtain specific irradiation data by molecular dynamics simulations. We find that irradiation with 1 MeV Kr ions can indeed induce the spontaneous formation of a nanocomposite, in agreement with recent experimental results. Our results suggest, on the other hand, that this type of microstructure could not be obtained by irradiation with lighter particles, such as He ions. The simulation results predict, as well, the dependency of the structure factor on the type of irradiation particle used and are in qualitative agreement with the formula derived analytically. We propose that these statements can be tested by small angle scattering experiments

  5. Helium effects on irradiation dmage in V alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  6. Positron annihilation study of neutron irradiated pure Fe and Fe-Cu binary alloys

    International Nuclear Information System (INIS)

    The hardening and embrittlement of Reactor Pressure Vessel (RPV) steels is of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, like vacancies, interstitials, solutes and their clusters. Fe-Cu binary alloys are often used to mimic the behaviour of such steels. Their study allows identifying some of the defects responsible of the hardening, especially when compared to pure iron or C-micro-alloyed iron. Owing to their self-seeking and selective trapping, positrons are used to determine the nature of these defects. Recently, at SCK.CEN, a new Positron Annihilation Spectroscopy (PAS) setup has been built, calibrated and optimized to measure the Coincidence Doppler Broadening (CDB) and Lifetime (LT) of neutron irradiated materials. This set-up has been used to measure the CDB and LT of n-irradiated pure Fe and Fe-Cu alloys. It is found that the clustering of Cu take place at the very early stages of irradiation using the CDB while LT measurement are showing much more vacancy clustering for low Cu alloys than in the higher ones. Increasing the neutron dose up to 1.3 x 1020 n/m2, allows the follow up of the kinetic of Cu and V-clustering especially in Fe-Cu alloys. It is found that both copper and carbon decrease the size of vacancy-cluster, when added to iron. (copyright 2007 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  7. Effect of strain on ferrite transformation from super-cooled austenite in Fe-0. 5%C alloy. Fe-0. 5%C gokin no karei osutenaito/feraito hentai ni oyobosu kako no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, K.; Ito, Y.; Narita, T. (Hokkaido Univ., Sapporo (Japan). Faculty of Engineering)

    1993-08-01

    During the cooling of a steel, when austenite is applied by strain, the temperature of ferrite transformation would increase accompanied with decrease of its given temperature and increase of strain. In this study, the isothermal transformation behaviour from austenite to ferrite applied by strain in the super-cooled state was investigated, effect of strain on size of ferrite particles and increase of volume rate during transformation were explained by using the velocity theory. That is, concerning to the alloy of two-elemental system Fe-0.51%C cooled at 0.3[degree]C/s and applied by strain at 710[degree]C, at which austenite was super-cooled by 55[degree]C, its isothermal transformation behaviour was investigated. As a result, the following conclusions were obtained. Time required for the transformation remarkably decreased and the size of ferrite particles became ultra-fine subjected to strain. The nucleation rate of ferrite particles remarkably increased with increasing strain. 14 refs., 11 figs., 1 tab.

  8. Modification of a tool hard alloys by high power ion beams irradiation and following thermal annealing

    International Nuclear Information System (INIS)

    Perspective technological decision of the tool materials wear resistance increasing problem is offered. The method combines surface hardening by high power ion beams irradiation and following high temperature annealing. In this paper structure and phase changes in the near-surface layers of the hard alloys W-Co under high power pulsed ion beams irradiation and its evolution in the course of high temperature annealing have been investigated. As a result of thermal effect in near-surface of the area the layer with high strength is formed.This layer defines tribological properties of a material. It excludes fragile destruction of a tool material and promotes increase of wear resistance. The kinetic of wear of modified tool alloys during cutting is established. New method of modification provides the improvement of wear resistance of toll at cutting steel, nickel and titanium alloys. Distinctive features of the thermal annealing: availability of many phases outside of dependence on regimes of the preliminary ion beam irradiation; the low density of defects of a W-phase in comparison with the irradiated samples only (extended mosaic structure, low dislocation density and defects of packing, the reduced micro distortions of a crystal lattice); high degree of perfection of a crystal structure of a Co-phase

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

    International Nuclear Information System (INIS)

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

  10. Corrosion and creep properties of M5 alloy: optimization and validation under irradiation

    International Nuclear Information System (INIS)

    The development program of advanced zirconium alloys for cladding was defined by FRAMATOME during the eighties to meet utility requirements regarding fuel rod corrosion, hydriding, creep and growth performance at high burn-up and under severe operating conditions in PWR environment. From the irradiation in commercial plants of claddings made with 4 pre-selected alloys, M5 alloy was proposed as the best candidate early in the nineties. A complete qualification process for M5 cladding was engaged at that time in order to satisfy licensing requirements of Safety Authorities. Regarding performance under normal operating conditions, the process for M5 qualification included two steps: An extensive optimization program based on out-of-pile properties. Optimization was intended to assess the impact of chemical composition and manufacturing parameters on out-of-pile properties. Possible causes for variability in tube final properties were determined by tube production runs at CEZUS and ZIRCOTUBE facilities. The results of the optimization program allowed to define the corrosion - creep optimum and were implemented in the definition of M5 product within the scope of its industrialization. The validation under irradiation involving 15 plants in 8 countries. The analysis of irradiation feed-back is supported by post-irradiation examinations performed after several campaigns in 10 commercial plants. Validation included the main manufacturing features selected for the industrial product and was obtained within a large range of operating conditions. This paper is focused on these two qualification steps, regarding more particularly corrosion and creep performance (author) (ml)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (ed.)

    1980-10-01

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

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

    International Nuclear Information System (INIS)

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

  16. Precipitation behaviour of a ferritic FeNiAl alloy under irradiation

    International Nuclear Information System (INIS)

    Coarsening of coherent NiAl (β') precipitates in a ferritic Fe-8.3Al-3.ONi-O.1Nb (at.%) alloy was investigated during thermal annealing and during irradiation with 6.2 MeV protons at temperatures from 673 K to 973 K. The matrix concentration of Ni and Al was traced by resistivity measurements while the growth of the precipitates was studied quantitatively by transmission electron microscopy (TEM). Additional Vickershardness measurements gave information on precipitation strengthening in the alloy. (orig.)

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

    International Nuclear Information System (INIS)

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

  18. Power ultrasound irradiation during the alkaline etching process of the 2024 aluminum alloy

    Science.gov (United States)

    Moutarlier, V.; Viennet, R.; Rolet, J.; Gigandet, M. P.; Hihn, J. Y.

    2015-11-01

    Prior to any surface treatment on an aluminum alloy, a surface preparation is necessary. This commonly consists in performing an alkaline etching followed by acid deoxidizing. In this work, the use of power ultrasound irradiation during the etching step on the 2024 aluminum alloy was studied. The etching rate was estimated by weight loss, and the alkaline film formed during the etching step was characterized by glow discharge optical emission spectrometry (GDOES) and scanning electron microscope (SEM). The benefit of power ultrasound during the etching step was confirmed by pitting potential measurement in NaCl solution after a post-treatment (anodizing).

  19. Irradiation Embritlement in Alloy HT-­9

    Energy Technology Data Exchange (ETDEWEB)

    Serrano De Caro, Magdalena [Los Alamos National Laboratory

    2012-08-27

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

    International Nuclear Information System (INIS)

    The objective of this effort is to provide data on the swelling and segregation response of alloys used in ongoing fission-fusion correlation efforts. Additions of silicon and aluminum suppress the neutron-induced swelling of pure nickel, but to different degrees. Silicon is much more effective initially when compared to aluminum on a per-atom basis, but silicon exhibits a non-monotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminum segregates away from these boundaries. Whereas the formation of the Ni3Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation

  6. Study of the precipitation and of the hardening microscopic mechanisms under irradiation in dilute ferritic alloys

    International Nuclear Information System (INIS)

    The copper precipitation plays a significant role in the embrittlement process of reactor vessel steels under neutron irradiation at 300 deg C. In order to understand the copper precipitation mechanisms, we have studied model ferritic binary FeCu and ternary alloys FeCuX (X=Mn,Ni, Cr, P). These materials have been either Irradiated with 2.5 MeV electrons In the 175-360 deg C temperature range or thermal aged at 500 deg C. The evolution of materials has been followed by resistivity measurements under irradiation, by small angle neutron scattering and by Vickers microhardness measurements. We have shown the similarity of copper precipitation under thermally ageing at 500 deg C and electron Irradiation at 300 deg C, in FeCu1,34%. This result confirms that the main effect of electronic irradiation is to accelerate precipitation. Nevertheless, we have observed that irradiation induces an additional contribution to hardening attributed to point defect clusters. Concerning the ternary alloys, we observed that at 300 deg C the addition of a third element has no significant effect on the copper precipitation kinetic under irradiation but that at lower temperature manganese slows down precipitation kinetic. In order to reproduce the experimental results obtained on FeCu1,34% by using a cluster kinetics model, we have to suppose that the precipitation is heterogeneous and controlled by interface reactions for the small size clusters. In addition, neutron or electron irradiated industrial steels have been studied by small angle neutron scattering. The results revealed the presence of nano-metric solute clusters which contain few copper atoms and which are not linked to the formation of displacement cascades. (author)

  7. Defects in hyperpure Fe-based alloys created by 3MeV e--irradiation

    International Nuclear Information System (INIS)

    Information about vacancy defects created in RPV (Reactor Pressure Vessels) steels after neutron irradiations are obtained via a simulation: The RPV steels are simulated by a series of high purity Fe-based alloys representing the industrial alloy composition. The neutron irradiation is simulated by a 3MeV electron irradiation. Vacancy defects characteristics are obtained by positron lifetime techniques. Irradiations are made at 150 C or 288 C, with a dose of 4*1019 e-/cm2, and followed by isochronal annealing in the range 20 -500 C. The observed vacancy defects are single trapped vacancies and small vacancy clusters, the size of which being lower than 10 empty atomic volumes (Vacancy clusters containing more than 50 empty atomic volumes were never found). A large recovery step is observed between 200 and 400 C, after 150 C irradiation and attributed to vacancy-impurity detrapping, and also, vacancy cluster evaporation. The influence of C, Cu and Mo are presented. These results are in agreement with a model supposing, in pure Fe, single vacancy migration at -50 C and vacancy-impurity detrapping at 200 C. (orig.)

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

    Science.gov (United States)

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

    2010-01-01

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

  9. A brief review of cavity swelling and hardening in irradiated copper and copper alloys

    International Nuclear Information System (INIS)

    The literature on radiation-induced swelling and hardening in copper and its alloy is reviewed. Void formation does not occur during irradiation of copper unless suitable impurity atoms such as oxygen or helium are present. Void formation occurs for neutron irradiation temperatures of 180 to 550 degree C, with peak swelling occurring at ∼320 degree C for irradiation at a damage rate of 2 x 10-7 dpa/s. The post-transient swelling rate has been measured to be ∼0.5%/dpa at temperatures near 400 degree C. Dispersion-strengthened copper has been found to be very resistant to void swelling due to the high sink density associated with the dispersion-stabilized dislocation structure. Irradiation of copper at temperatures below 400 degree C generally causes an increase in strength due to the formation of defect clusters which inhibit dislocation motion. The radiation hardening can be adequately described by Seeger's dispersed barrier model, with a barrier strength for small defect clusters of α ∼ 0.2. The radiation hardening apparently saturates for fluences greater than ∼1024 n/m2 during irradiation at room temperature due to a saturation of the defect cluster density. Grain boundaries can modify the hardening behavior by blocking the transmission of dislocation slip bands, leading to a radiation- modified Hall-Petch relation between yield strength and grain size. Radiation-enhanced recrystallization can lead to softening of cold-worked copper alloys at temperatures above 300 degree C

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-22

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

  12. Investigation of point defects after low temperature irradiation and qenching in diluted W57Co alloys

    International Nuclear Information System (INIS)

    In high purity tungsten foils and monocrystals with implanted 57Co probe atoms (diluted W-Co alloys) were point defects investigated using Moessbauer spectroscopy and resistivity measurements. The point defects were produced by irradiation with reactor neutrons, with 3 MeV electrons, or by quenching. It is shown that in irradiated samples moving interstitials and vacancies can be trapped by motionless 57Co atoms. Annealing of point defects was performed in the temperature range from 4.2 K to 1683 K. (GSCH)

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

    Science.gov (United States)

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

    2011-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-01

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

  15. Behavior of Zr1 %Nb Alloy Under Swift Kr Ion and Intense Electron Irradiation

    Directory of Open Access Journals (Sweden)

    V.F. Klepikov

    2015-12-01

    Full Text Available In this paper there were studied the physical and mechanical properties of Zr1 % Nb alloy irradiated by the Kr ion beam with energy of 107 MeV, fluences of 1·1013 and 1014 ion/cm2, and exposed to the microsecond high-current pulsed electron beam with the energy of 370 keV, incident energy fluence in the range of 20…200 J/cm2. The low-intense Kr ion implantation induced softening of the alloy. The high-intense ion beam irradiation resulted in creation of a surface strengthened layer with nanohardness of 4.5 GPa and the elastic modulus of 120 GPa. The high-current electron beam exposure lead to the macroscopic surface melting of the sample, which provoked formation of the coarse structure with predominantly brittle fracture character.

  16. Internal friction study of dislocation dynamics in neutron irradiated iron, and iron-copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Konstantinovic, M.J., E-mail: mkonstan@sckcen.b [Studiecentrum voor Kernenergie/Centre d' Etude de l' Energie Nucleaire (SCK-CEN), Boeretang 200, B-2400 Mol (Belgium)

    2009-12-15

    The temperature dependent internal friction spectra of cold-worked and neutron irradiated iron and iron-copper binary alloys are investigated. By increasing dose, both gamma- and Snoek-Koester-relaxation peaks exhibit strong shift towards low temperatures, as a consequence of the reduction of double kink activation energy. This shift is found to be the largest in alloys with the highest copper content. Besides, new modes appear in the spectra at energies of about 410 and 540 K. The 410 K peak intensity increases at the expense of Snoek-Koester peak intensity, indicating that redistribution of carbon takes place under irradiation, most probably as a result of grain boundary segregation. The presence of copper impedes the carbon redistribution by influencing the formation of carbon-vacancy complexes, which causes the grain boundary segregation, and activation of the 410 K relaxation process at larger neutron fluence in comparison with pure iron.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

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

  19. Proton-irradiation-induced damage in Fe–0.3 wt.%Cu alloys characterized by positron annihilation and nanoindentation

    International Nuclear Information System (INIS)

    Proton irradiation in combination with positron annihilation and nanoindentation offers the possibility to characterize irradiation damage in a range of dose. Proton irradiation experiments for Fe–0.3 wt.%Cu alloys were carried out at about 100 °C with an energy of 240 keV. Slow positron beam Doppler broadening spectroscopy was used to detect the depth dependence of microstructure evolution. The constant stiffness measurement (CSM) with a diamond Berkovich indenter was used to obtain the depth profile of hardness. The results showed that S-parameter in the analysis of positron annihilation increased with dose after irradiation owing to open-volume defects induced by proton irradiation. For specimens irradiated with different dose, hardness values exceeded that of un-irradiated alloys. The correlation between positron parameters and hardness was found. The hardness of any dpa was also calculated by Kasada method

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

    International Nuclear Information System (INIS)

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