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Sample records for austenitic alloys irradiated

  1. Irradiation-assisted stress corrosion cracking of austenitic alloys

    International Nuclear Information System (INIS)

    Was, G.S.; Atzmon, M.

    1991-01-01

    An experimental program has been conducted to determine the mechanism of irradiation-assisted stress-corrosion cracking (IASCC) in austenitic stainless steel. High-energy protons have been used to produce grain boundary segregation and microstructural damage in samples of controlled impurity content. The densities of network dislocations and dislocation loops were determined by transmission electron microscopy and found to resemble those for neutron irradiation under LWR conditions. Grain boundary compositions were determined by in situ fracture and Auger spectroscopy, as well as by scanning transmission electron microscopy. Cr depletion and Ni segregation were observed in all irradiated samples, with the degree of segregation depending on the type and amount of impurities present. P, and to a lesser extent P, impurities were observed to segregate to the grain boundaries. Irradiation was found to increase the susceptibility of ultra-high-purity (UHP), and to a much lesser extent of UHP+P and UHP+S, alloys to intergranular SCC in 288 degree C water at 2 ppm O 2 and 0.5 μS/cm. No intergranular fracture was observed in arcon atmosphere, indicating the important role of corrosion in the embrittlement of irradiated samples. The absence of intergranular fracture in 288 degree C argon and room temperature tests also suggest that the embrittlement is not caused by hydrogen introduced by irradiation. Contrary to common belief, the presence of P impurities led to a significant improvement in IASCC over the ultrahigh purity alloy

  2. Irradiation-assisted stress-corrosion cracking in austenitic alloys

    International Nuclear Information System (INIS)

    Was, G.S.; Andresen, P.L.

    1992-01-01

    Irradiation-assisted stress-corrosion cracking (IASCC) in austentic alloys is a complicated phenomenon that poses a difficult problem for designers and operators of nuclear plants. Because IASCC accelerates the deterioration of various reactor components, it is imperative that it be understood and modeled to maintain reactor safety. Unfortunately, the costs and dangers of gathering data on radiation effects are high, and the phenomenon itself is so complex that it is difficult to enumerate all of the causes. This article reviews current knowledge of IASCC and describes the goals of ongoing work

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

    International Nuclear Information System (INIS)

    Huguenin, D.

    1989-01-01

    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 [fr

  4. 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.

  5. Swelling of austenitic iron-nickelchromium ternary alloys during fast neutron irradiation

    International Nuclear Information System (INIS)

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

    1984-01-01

    Swelling data are now available for 15 iron-nickel-chromium ternary alloys irradiated to exposures as high as 110 displacements per atom (dpa) in Experimental Breeder Reactor-II (EBR-II) between 400 and 650 0 C. These data confirm trends observed at lower exposure levels and extend the generality of earlier conclusions to cover a broader range of composition and temperature. It appears that all austenitic iron-nickel-chromium ternary alloys eventually approach an intrinsic swelling rate of about1%/dpa over a range of temperature even wider than studied in this experiment. The duration of the transient regime that precedes the attainment of this rate is quite sensitive to nickel and chromium content, however. At nickel and chromium levels typical of 300 series steels, swelling does not saturate at engineering-relevant levels. However, there appears to be a tendency toward saturation that increases with declining temperature, increasing nickel and decreasing chromium levels. Comparisons of these results are made with those of similar studies conducted with charged particles. Conclusions are then drawn concerning the validity of charged particle simulation studies to determine the compositional and temperature dependence of swelling

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

    International Nuclear Information System (INIS)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.

    2002-01-01

    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

  7. 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.

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

    NARCIS (Netherlands)

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

    2003-01-01

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

  9. Effects of pulsed and/or dual ion irradiation on microstructural evolution in a Ti and Si modified austenitic alloy

    International Nuclear Information System (INIS)

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

    1983-01-01

    The influence of pulsed 4 MeV Ni-ion bombardment, with and without simultaneous helium injection, at 958 K and damage levels from 1 to 50 dpa has been studied in a low swelling, Ti- and Si- modified austenitic stainless steel. Compared to continuous irradiation, pulsing caused an increase in the number density of interstitial loops formed during irradiation. Helium also increased the nucleation of interstitial loops. The main precipitates formed were a large number of small TiC particles uniformly distributed in the matrix, and a small number of relatively large eta and G precipitates. These course precipitates were somewhat larger in the pulsed specimens. Pulsing appeared to produce no significant change in swelling compared to continuous irradiation. However, for one specimen irradiated to 54 dpa, pulsing concurrent with substantial temperature fluctuations caused by beam heating may have been responsible for a larger swelling compared to continuous irradiation

  10. Tough and corrosion resistant austenitic alloy

    International Nuclear Information System (INIS)

    Johnson, T.E.

    1977-01-01

    The invention concerns austenitic alloys of high corrosion resistance, which can be deformed hot and tempered, so that they can be forged, rolled, and drawn into tubes and other shapes. The alloys have a basis of nickel, chromium and iron. The silicon content is between 2 and 4% by weight, and the molybdenum content is between 0 and 2% by weight. The alloys can be hardened by ageing and contain up to 0.1% by weight of boron. The other alloying materials are 1 to 3.5% by weight of manganese, 4 to 7.5% by weight of cobalt, 2.5 to 8% by weight of copper and 0.05 to 0.25% by weight of carbon. (IHOE) [de

  11. 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.)

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

    International Nuclear Information System (INIS)

    Boulanger, L.

    1988-06-01

    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 [fr

  13. 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)

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

    International Nuclear Information System (INIS)

    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

  15. Perspective on present and future alloy development efforts on austenitic stainless steels for fusion application

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.

    1984-01-01

    The purpose of this paper is to address important questions concerning how to effect further alloy development of austenitic stainless steels for resistance, and to what extent the behavior of other properties under irradiation, such as strength/embrittlement, fatigue/irradiation creep, corrosion (under irradiation), and radiation-induced activation must be influenced. To summarize current understanding, helium has been found to have major effects on swelling and embrittlement, but several metallurgical avenues are available for significant improvement relative to type 316 stainless steel. Studies on fatigue and irradiation creep, particularly including helium effects, are preliminary but have yet to reveal engineering problems requiring additional alloy development remedies. The effects of irradiation on corrosion behavior are unknown, but higher alloy nickel contents make thermal corrosion in lithium worse. 67 refs. (JDB)

  16. Perspective on present and future alloy development efforts on austenitic stainless steels for fusion application

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1984-01-01

    The purpose of this paper is to address important questions concerning how to effect further alloy development of austenitic stainless steels for resistance, and to what extent the behavior of other properties under irradiation, such as strength/embrittlement, fatigue/irradiation creep, corrosion (under irradiation), and radiation-induced activation must be influenced. To summarize current understanding, helium has been found to have major effects on swelling and embrittlement, but several metallurgical avenues are available for significant improvement relative to type 316 stainless steel. Studies on fatigue and irradiation creep, particularly including helium effects, are preliminary but have yet to reveal engineering problems requiring additional alloy development remedies. The effects of irradiation on corrosion behavior are unknown, but higher alloy nickel contents make thermal corrosion in lithium worse. 67 refs

  17. Assessing SCC and IASCC of austenitic alloys for application to the SCWR concept

    International Nuclear Information System (INIS)

    Teysseyre, S.; Was, G.S.

    2008-01-01

    From the standpoint of environmental degradation of material, the selection of alloys for use as structural material in a supercritical water-cooled reactor (SCWR) must include assessment of the corrosion and stress corrosion cracking susceptibility of the alloys in supercritical water. Moreover, as experience in current reactors showed that irradiation-assisted stress corrosion cracking (IASCC) is a. major concern, a comprehensive study must include the assessment of the effect of irradiation on SCC in supercritical water. Therefore, such selection faces multiple obstacles. The first is the lack of data on the corrosion and SCC susceptibility of the candidate alloys in this environment. There is a need to produce basic data using complementary experimental techniques. The second is the difficulty to obtain material irradiated in conditions relevant for SCWR. Availability of such material is needed to determine the influence of irradiation and its influence on SCC. Techniques such as proton irradiation are appealing surrogates for neutron irradiation in assessing its effect of stress corrosion cracking initiation, and can be used for screening of various material and environmental conditions. However, neutron irradiation is required to confirm the role of in-core irradiation on crack growth and in performing final verification of the effect of alternative irradiation on candidate alloys. Another obstacle would be the lack of facilities for testing materials in the unirradiated and irradiated state in supercritical water. The University of Michigan has developed a comprehensive programme to assess the stress corrosion cracking susceptibility of austenitic alloys in supercritical water in unirradiated, proton-irradiated and neutron-irradiated state. The cracking susceptibility of unirradiated alloys has been evaluated by a set of constant extension rate tensile, CERT, experiments and by determination of the crack propagation rate by DCPD technique under constant K

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

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

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

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

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

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

  20. 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.

  1. Effect of neutron irradiation on vanadium alloys

    International Nuclear Information System (INIS)

    Braski, D.N.

    1986-01-01

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

  2. Effect of neutron irradiation on vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

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

  3. 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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    Fe-18Cr-9Ni-1,5Mn austenitic alloys containing phosphorus, silicon and sulphur were irradiated by 1 MeV electrons at 573-773 K. Phosphorus increased the interstitial loop nucleation and decreased the void swelling by increasing void number density and suppressing void growth. Silicon had a similar effect to phosphorus but its effect was weaker than phosphorus. Sulphur enhanced void swelling through increasing the void density. Nickel enrichment at grain boundaries was suppressed only in the alloy containing phosphorus. These phosphorus effects may be explained by a strong interaction with interstitials resulting in a high density of sinks for point defects. (orig.)

  5. The influence of combined addition of phosphorus and titanium on void swelling of austenitic Fe-Cr-Ni alloys at 646-700 K

    International Nuclear Information System (INIS)

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

    1994-01-01

    The influence of combined addition of phosphorus and titanium on void swelling of model Fe-Cr-Ni austenitic alloys at 646 to 700 K under fast neutron irradiation has been investigated, in comparison with that of a complex austenitic alloy (JPCA-2). In the model alloys, void swelling decreased with increasing phosphorus content. Void average size and density of JPCA-2 were comparable to those of the 0.024P alloy. The fact that these two alloys have the same phosphorus level suggests the void swelling of the model alloys would be strongly suppressed by increasing the phosphorus concentration and/or coaddition of phosphorus and titanium. The present study demonstrated that the phosphorus level is the strongest determinant of void swelling of both model and complex austenitic alloys. ((orig.))

  6. Development of advanced austenitic stainless steels resistant to void swelling under irradiation

    International Nuclear Information System (INIS)

    Rouxel, Baptiste

    2016-01-01

    In the framework of studies about Sodium Fast Reactors (SFR) of generation IV, the CEA is developing new austenitic steel grades for the fuel cladding. These steels demonstrate very good mechanical properties but their use is limited because of the void swelling under irradiation. Beyond a high irradiation dose, cavities appear in the alloys and weaken the material. The reference material in France is a 15Cr/15Ni steel, named AIM1, stabilized with titanium. This study try to understand the role played by various chemical elements and microstructural parameters on the formation of the cavities under irradiation, and contribute to the development of a new grade AIM2 more resistant to swelling. In an analytical approach, model materials were elaborated with various chemical compositions and microstructures. Ten grades were cast with chemical variations in Ti, Nb, Ni and P. Four specific microstructures for each alloy highlighted the effect of dislocations, solutes or nano-precipitates on the void swelling. These materials were characterized using TEM and SANS, before irradiation with Fe 2+ (2 MeV) ions in the order to simulate the damages caused by neutrons. Comparing the irradiated microstructures, it is demonstrated that the solutes have a dominating effect on the formation of cavities. Specifically titanium in solid solution reduces the swelling whereas niobium does not show this effect. Finally, a matrix enriched by 15% to 25% of nickel is still favorable to limit swelling in these advanced austenitic stainless steels. (author) [fr

  7. Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

    Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

  8. Resistance Element Welding of Magnesium Alloy/austenitic Stainless Steel

    Science.gov (United States)

    Manladan, S. M.; Yusof, F.; Ramesh, S.; Zhang, Y.; Luo, Z.; Ling, Z.

    2017-09-01

    Multi-material design is increasingly applied in the automotive and aerospace industries to reduce weight, improve crash-worthiness, and reduce environmental pollution. In the present study, a novel variant of resistance spot welding technique, known as resistance element welding was used to join AZ31 Mg alloy to 316 L austenitic stainless steel. The microstructure and mechanical properties of the joints were evaluated. It was found that the nugget consisted of two zones, including a peripheral fusion zone on the stainless steel side and the main fusion zone. The tensile shear properties of the joints are superior to those obtained by traditional resistance spot welding.

  9. Empirical relations for tensile properties of austenitic stainless steels irradiated in mixed-spectrum reactors

    International Nuclear Information System (INIS)

    Grossbeck, M.L.

    1991-01-01

    An assessment has been made of available tensile property data relevant to the design of fusion reactors, especially near term devices expected to operate at lower temperatures than power reactors. Empirical relations have been developed for the tensile properties as a functions of irradiation temperature for neutron exposures of 10-15, 20, 30, and 50 dpa. It was found that yield strength depends little on the particular austenitic alloy and little on the helium concentration. Strength depends upon initial condition of the alloy only for exposures of less than 30 dpa. Uniform elongation was found to be more sensitive to alloy and condition. It was also more sensitive than strength to helium level. However, below 500deg C, helium only appeared to have an efect at 10-15 dpa. At higher temperatures, helium embrittlement was apparent, and its threshold temperature decreased with increasing neutron exposure level. (orig.)

  10. 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.

  11. Radiation induced phosphorus segregation in austenitic and ferritic alloys

    International Nuclear Information System (INIS)

    Brimhall, J.L.; Baer, D.R.; Jones, R.H.

    1984-01-01

    The radiation induced surface segregation (RIS) of phosphorus in stainless steel attained a maximum at a dose of 0.8 dpa then decreased continually with dose. This decrease in the surface segregation of phosphorus at high dose levels has been attributed to removal of the phosphorus layer by ion sputtering. Phosphorus is not replenished since essentially all of the phosphorus within the irradiation zone has been segregated to the surface. Sputter removal can explain the previously reported absence of phosphorus segregation in ferritic alloys irradiated at high dosessup(1,2) (>1 dpa) since irradiation of ferritic alloys to low doses has shown measurable RIS. This sputtering phenomenon places an inherent limitation to the heavy ion irradiation technique for the study of surface segregation of impurity elements. The magnitude of the segregation in ferritics is still much less than in stainless steel which can be related to the low damage accumulation in these alloys. (orig.)

  12. Void swelling behaviour of austenitic stainless steel during electron irradiation

    International Nuclear Information System (INIS)

    Sheng Zhongqi; Xiao Hong; Peng Feng; Ti Zhongxin

    1994-04-01

    The irradiation swelling behaviour of 00Cr17Ni14Mo2 austenitic stainless steel (AISI 316L) was investigated by means of high voltage electron microscope. Results showed that in solution annealed condition almost no swelling incubation period existed, and the swelling shifted from the transition period to the steady-state one when the displacement damage was around 40 dpa. In cold rolled condition there was evidently incubation period, and when the displacement damage was up to 84 dpa the swelling still remained in the transition period. The average size and density of voids in both conditions were measured, and the factors, which influenced the void swelling, were discussed. (3 figs.)

  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. Precipitation hardening in Fe--Ni base austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chang, K.M.

    1979-05-01

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

  15. 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.

  16. Internal Friction of Austenitic Fe-Mn-C-Al Alloys

    Science.gov (United States)

    Lee, Young-Kook; Jeong, Sohee; Kang, Jee-Hyun; Lee, Sang-Min

    2017-12-01

    The internal friction (IF) spectra of Fe-Mn-C-Al alloys with a face-centered-cubic (fcc) austenitic phase were measured at a wide range of temperature and frequency ( f) to understand the mechanisms of anelastic relaxations occurring particularly in Fe-Mn-C twinning-induced plasticity steels. Four IF peaks were observed at 346 K (73 °C) (P1), 389 K (116 °C) (P2), 511 K (238 °C) (P3), and 634 K (361 °C) (P4) when f was 0.1 Hz. However, when f increased to 100 Hz, whereas P1, P2, and P4 disappeared, only P3 remained without the change in peak height, but with the increased peak temperature. P3 matches well with the IF peak of Fe-high Mn-C alloys reported in the literature. The effects of chemical composition and vacancy (v) on the four IF peaks were also investigated using various alloys with different concentrations of C, Mn, Al, and vacancy. As a result, the defect pair responsible for each IF peak was found as follows: a v-v pair for P1, a C-v pair for P2, a C-C pair for P3, and a C-C-v complex (major effect) + a Mn-C pair (minor effect) for P4. These results showed that the IF peaks of Fe-Mn-C-Al alloys reported previously were caused by the reorientation of C in C-C pairs, not by the reorientation of C in Mn-C pairs.

  17. Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

    Directory of Open Access Journals (Sweden)

    Anna Hojná

    2017-09-01

    Full Text Available Austenitic stainless steels are normally ductile and exhibit deep dimples on fracture surfaces. These steels can, however, exhibit brittle intergranular fracture under some circumstances. The occurrence of intergranular fracture in the irradiated steels is briefly reviewed based on limited literature data. The data are sorted according to the irradiation temperature. Intergranular fracture may occur in association with a high irradiation temperature and void swelling. At low irradiation temperature, the steels can exhibit intergranular fracture at low or even at room temperatures during loading in air and in high temperature water (~300 °C. This paper deals with the similarities and differences for IG fractures and discusses the mechanisms involved. The intergranular fracture occurrence at low temperatures might be correlated with decohesion or twinning and strain martensite transformation in local narrow areas around grain boundaries. The possibility of a ductile-to-brittle transition is also discussed. In case of void swelling higher than 3%, quasi-cleavage at low temperature might be expected as a consequence of ductile-to-brittle fracture changes with temperature. Any existence of the change in fracture behavior in the steels of present thermal reactor internals with increasing irradiation dose should be clearly proven or disproven. Further studies to clarify the mechanism are recommended.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Effect of Heat Treatment on Conductivity of Metastable Alloys with Ferromagnetic and Paramagnetic Austenite

    Science.gov (United States)

    Uvarov, A. N.; Sandovskii, V. A.; Vil'danova, N. F.; Anufrieva, E. I.

    2013-11-01

    Invars N30K10T3 and N28K10T3 and nonmagnetic austenitic alloy N25Kh2T3 are studied after different kinds of treatment, i.e., quenching, mechanical phase hardening, and deformation followed by aging. The structure and the conductivity of the alloys are determined. An optimum treatment for providing high electric conductivity is suggested.

  20. Diffusion Couple Alloying of Refractory Metals in Austenitic and Ferritic/Martensitic Steels

    Science.gov (United States)

    2012-03-01

    temperature (DBTT) and lower upper shelf energy (USE) obtained via a Charpy impact test (austenitic steels , however, do not experience DBTT) as seen in...ALLOYING OF REFRACTORY METALS IN AUSTENITIC AND FERRITIC/MARTENSITIC STEELS by Alexander L. McGinnis March 2012 Thesis Advisor: Luke...Ferritic/Martensitic Steels 5. FUNDING NUMBERS 6. AUTHOR(S) Alexander L. McGinnis 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) Naval

  1. 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 〈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.

  2. Effects of alloying elements on the kinetics of austenitization from pearlite in Fe-C-M alloys

    Science.gov (United States)

    Xia, Yuan; Enomoto, Masato; Yang, Zhigang; Li, Zhaodong; Zhang, Chi

    2013-03-01

    The effects of alloying elements on the kinetics of austenitization from pearlite structure were studied by computer simulation in Fe-C-M ternary alloys, where M is Mn, Cr, Si or Ni, assuming local equilibrium conditions at all transformation interfaces. A thin austenite film was assumed to nucleate at ferrite/cementite interfaces and grow in one dimension. The existence of a partition to no-partition transition temperature (PNTT) was rationalized. Above the PNTT, the growth rate of austenite is governed by the difference in carbon activity between austenite/cementite and ferrite/austenite interfaces; a substitutional element influences the reaction rate by affecting carbon activity. Below the PNTT, redistribution of M is necessary. The PNTT increases with the concentration of all alloy elements except Ni, which has a large segregation tendency in austenite from both ferrite and cementite, as well as repulsive interaction with carbon. The amount of overheating at PNTT from Ae1 increases in the order Si (∼Ni), Mn and Cr, essentially in agreement with a recently reported experiment.

  3. The conflicting roles of boron on the radiation response of precipitate-forming austenitic alloys

    International Nuclear Information System (INIS)

    Okita, T.; Sekimura, N.; Garner, F.

    2007-01-01

    Full text of publication follows: Boron is often a deliberately added solute to improve the radiation resistance of austenitic structural alloys, with boron exerting its greatest influence on carbide precipitation. However, boron also a source of helium via transmutation and therefore tends to accelerate the onset of void nucleation. These conflicting contributions of boron with respect to radiation resistance are not easily separated, but are sometimes utilized to mimic fusion-relevant gas generation rates when testing in surrogate fission spectra. In an earlier study the authors demonstrated that in simple model ternary alloys that boron additions tended to homogenize swelling somewhat via increased helium generation but not to exert any significant influence on the total swelling. In these easily swelling alloys void nucleation was not significantly influenced by additional helium or by boron's chemical effect, with boron remaining primarily in solution. In the current study, Fe-15Cr-16Ni-0.25 Ti-0.05C alloys with four levels of natural boron addition (0, 100, 500, 2500 appm) were irradiated side-by-side at ∼400 deg. C in the Fast Flux Test Facility under active temperature control in the Materials Open Test Assembly. Although three sets of irradiation conditions were explored, the boron variation was the only variable operating in each data set. The bulk swelling was measured using an immersion density technique and electron microscopy was employed to determine the details of void, dislocation and precipitate microstructure. It was found that by 100 appm B the strongest and most immediate effect of boron was to reduce swelling at all irradiation conditions explored, but the boron-induced increases in overall helium content were rather small over the 0-100 appm B range. This indicates that boron's primary effect was chemical in nature, expressed via its effect on precipitation. As the boron level was progressively increased, however, there was a reversal in

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-31

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

  5. Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Sekio, Yoshihiro; Yamashita, Shinichiro; Takahashi, Heishichiro; Sakaguchi, Norihito

    2014-01-01

    Addition of minor elements to a base alloy is often applied with the aim of mitigating void swelling by decreasing the vacancy diffusivity and flux which influence vacancy accumulation behavior. However, the comparative evaluations of parameters, such as the diffusivity and flux, between a base alloy and modified alloys with specific additives have not been studied in detail. In this study, type 316 austenitic stainless steel as a base alloy and type 316 austenitic stainless steels modified with vanadium (V) or zirconium (Zr) additions were used to perform evaluations from the changes of widths of the void denuded zone (VDZ) formed near a random grain boundary during electron irradiation because these widths depend on vacancy diffusivity and flux. The formations of VDZs were observed in in-situ observations during electron irradiation at 723 K and the formed VDZ widths were measured from the transmission electron microscopic images after electron irradiation. As a result, the VDZs were formed in both steels without and with V, and respective widths were ∼119 and ∼100 nm. On the other hand, the VDZ formation was not observed clearly in the steel with Zr. From the measured VDZ widths in the steels without and with V addition, the estimated ratio of the vacancy diffusivity in the steel with V to that in the steel without V was about 0.50 and the estimated ratio of the vacancy flux in the steel with V to that in the steel without V was about 0.71. This result suggests that the effect of additional minor elements on vacancy accumulation behaviors under electron irradiation could be estimated from evaluations of the VDZ width changes among steels with and without minor elements. Especially, because void swelling is closely related with the vacancy diffusion process, the VDZ width changes would also be reflected on void swelling behavior. (author)

  6. Mechanical and irradiation properties of zirconium alloys irradiated in HANARO

    International Nuclear Information System (INIS)

    Kwon, Oh Hyun; Eom, Kyong Bo; Kim, Jae Ik; Suh, Jung Min; Jeon, Kyeong Lak

    2011-01-01

    These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, 1.1 10 21 n/cm 2 ). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed

  7. Natural composite in austenitic alloys with a structure of the complete discontinuous decomposition

    International Nuclear Information System (INIS)

    Zemtsova, N.D.; Anufrieva, E.I.; Uvarov, A.I.; Vasechkina, T.P.; Sandovskij, V.A.

    2002-01-01

    The papers are reviewed on heat treatment conditions resulting in a 100% degree of discontinuous precipitation in austenite on ageing metastable Fe-Ni-Ti alloys (N2KhT2, N26T3, N25T5, N24Kh2T3, N29T3, N29KhT3, N26Kh5T3). Mechanical properties and structure of the alloys are investigated after various heat and thermomechanical treatments. It is revealed that discontinuous precipitation is accomplished by way of migration of low-angle boundaries as the matrix is supersaturated essentially with alloying elements. The alloys with the structure of 100% discontinuous precipitation can be treated as natural composites consisting of alternating plates of intermetallic compound and austenite. Temperature dependences of strength and plastic properties of a composite material and a hardened alloy are compared [ru

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

    OpenAIRE

    A. Janus; A. Kurzawa

    2011-01-01

    Within the research, determined were direction and intensity of alloying elements influence on solidification way (directional orvolumetric) of primary austenite dendrites in hypoeutectic austenitic cast iron Ni-Mn-Cu. 50 cast shafts dia. 20 mm were analysed.Chemical composition of the alloy was as follows: 1.7 to 3.3 % C, 1.4 to 3.1 % Si, 2.8 to 9.9 % Ni, 0.4 to 7.7 % Mn, 0 to 4.6 % Cu, 0.14 to0.16 % P and 0.03 to 0.04 % S. The discriminant analysis revealed that carbon influences solidifica...

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

    International Nuclear Information System (INIS)

    Grandjean, Y.

    1996-01-01

    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 50 Ni 50 and Fe 49 Ni 50 Hf 1 alloys. (author)

  10. 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.

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

    International Nuclear Information System (INIS)

    Dalton, J.H.

    1978-01-01

    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 [af

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

    International Nuclear Information System (INIS)

    Yamagata, Ichiro; Konno, Shotaro; Hayashi, Takehiro; Takaya, Shigeru

    2012-01-01

    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)

  13. Recent experimental and theoretical insights on the swelling of austenitic alloys

    International Nuclear Information System (INIS)

    Garner, F.A.; Wolfer, W.G.

    1983-01-01

    Once void nucleation subsides, the swelling rate of many austenitic alloys becomes rather insensitive to the variables that determine the duration of the transient regime of swelling. Models are presented which describe the roles of nickel, chromium and silicon in void nucleation. The relative insensitivity of steady-state swelling to temperature and composition is also discussed

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

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

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

  15. 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

  16. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part I. Ductility and fracture toughness

    Energy Technology Data Exchange (ETDEWEB)

    Margolin, B., E-mail: mail@crism.ru; Sorokin, A.; Shvetsova, V.; Minkin, A.; Potapova, V.; Smirnov, V.

    2016-11-15

    The radiation swelling effect on the fracture properties of irradiated austenitic steels under static loading has been studied and analyzed from the mechanical and physical viewpoints. Experimental data on the stress-strain curves, fracture strain, fracture toughness and fracture mechanisms have been represented for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various swelling. Some phenomena in mechanical behaviour of irradiated austenitic steels have been revealed and explained as follows: a sharp decrease of fracture toughness with swelling growth; untypical large increase of fracture toughness with decrease of the test temperature; some increase of fracture toughness after preliminary cyclic loading. Role of channel deformation and channel fracture has been clarified in the properties of irradiated austenitic steel and different tendencies to channel deformation have been shown and explained for the same austenitic steel irradiated at different temperatures and neutron doses.

  17. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    Science.gov (United States)

    Lu, Qiaofeng; Su, Qing; Wang, Fei; Zhang, Chenfei; Lu, Yongfeng; Nastasi, Michael; Cui, Bai

    2017-06-01

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments.

  18. Hydrogen solubility in austenite of Fe-Ni-Cr alloys

    International Nuclear Information System (INIS)

    Zhirnova, V.V.; Mogutnov, B.M.; Tomilin, I.A.

    1981-01-01

    Hydrogen solubility in Fe-Ni-Cr alloys at 600-1000 deg C is determined. Hydrogen solubility in ternary alloys can not be predicted on the basis of the data on its solubility in binary Fe-Ni, Fe-Cr alloys. Chromium and nickel effect on hydrogen solubility in iron is insignificant in comparison with the effect of these elements on carbon or nitrogen solubility [ru

  19. Fundamental irradiation studies on vanadium alloys

    International Nuclear Information System (INIS)

    Loomis, B.A.; Garner, F.A.; Ermi, A.M.

    1985-05-01

    A joint experiment on the irradiation response of simple vanadium alloys has been initiated under the auspices of the DAFS and BES progams. Specimen fabrication is nearly complete and the alloys are expected to be irradiated in lithium in FFTF-MOTA Cycles 7 and 8

  20. Evaluation of High Temperature Corrosion Resistance of Finned Tubes Made of Austenitic Steel And Nickel Alloys

    Directory of Open Access Journals (Sweden)

    Turowska A.

    2016-06-01

    Full Text Available The purpose of the paper was to evaluate the resistance to high temperature corrosion of laser welded joints of finned tubes made of austenitic steel (304,304H and nickel alloys (Inconel 600, Inconel 625. The scope of the paper covered the performance of corrosion resistance tests in the atmosphere of simulated exhaust gases of the following chemical composition: 0.2% HCl, 0.08% SO2, 9.0% O2 and N2 in the temperature of 800°C for 1000 hours. One found out that both tubes made of austenitic steel and those made of nickel alloy displayed good resistance to corrosion and could be applied in the energy industry.

  1. Interpretation of the influences of irradiation upon fatigue crack propagation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Lloyd, G.J.

    1982-04-01

    An interpretation of the influences of neutron irradiation upon fatigue crack propagation in austenitic stainless steels is given. The approach has been to extend a previously developed rationalisation of the effects of various test and materials variables upon fatigue crack propagation in unirradiated stainless steels to include irradiated stainless steels. Irradiation has diverse influences upon the rate of fatigue crack propagation depending on the exact irradiation and test conditions. It has been shown that by considering the underlying mechanisms of failure, some confidence is established in trends in data in a subject where information is very scarce and difficult to obtain. (author)

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

    Science.gov (United States)

    Kaynak, Yusuf

    2014-09-01

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

  3. Irradiation effects in magnesium and aluminium alloys

    International Nuclear Information System (INIS)

    Sturcken, E.F.

    1979-01-01

    Effects of neutron irradiation on microstructure, mechanical properties and swelling of several magnesium and aluminium alloys were studied. The neutron fluences of 2-3 X 10 22 n/cm 2 , >0.2 MeV produced displacement doses of 20 to 45 displacements per atom (dpa). Ductility of the magnesium alloys was severely reduced by irradiation induced recrystallization and precipitation of various forms. Precipitation of transmuted silicon occurred in the aluminium alloys. However, the effect on ductility was much less than for the magnesium alloys. The magnesium and aluminium alloys had excellent resistance to swelling: The best magnesium alloy was Mg/3.0 wt% Al/0.19 wt% Ca; its density decreased by only 0.13%. The best aluminium alloy was 6063, with a density decrease of 0.22%. (Auth.)

  4. Swelling analysis of austenitic stainless steels by means of ion irradiation and kinetic modeling

    International Nuclear Information System (INIS)

    Kohyama, Akira; Donomae, Takako

    1999-03-01

    The influences of irradiation environment on the swelling behavior of austenitic stainless steel has been studied, to aid understanding the origin of the difference in swelling response of PNC316 stainless steel in fuel-pin environment and in materials irradiation capsules, in terms of irradiation conditions, damage mechanism and material conditions. This work focused on the theoretical investigation of the influence of temperature variation on microstructural development of austenitic stainless steels during irradiation, using a kinetic rate theory model. A modeling and calculation on non-steady irradiation effects were first carried out. A fully dynamic model of point defect evolution and extended defect development, which accounts for cascade damage, was developed and successfully applied to simulate the interstitial loop evolution in low temperature regimes. The influence of cascade interstitial clustering on dislocation loop formation has also been assessed. The establishment of a basis for general assessment of non-steady irradiation effects in austenitic stainless steels was advanced. The developed model was applied to evaluate the influences of temperature variation in formerly carried out CMIR and FFTF/MFA-1 FBR irradiation experiments. The results suggested the gradual approach of microstructural features to equilibrium states in all the temperature variation conditions and no sign of anomalous behavior was noted. On the other hand, there is the influence of temperature variation on microstructural development under the neutron irradiation, like CMIR. So there are some possibilities of the work of mechanism which is not taken care on this model, for example the effect of the precipitate behavior which is sensitive to irradiation temperature. (author)

  5. 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.

  6. Development of a high strength, hydrogen-resistant austenitic alloy

    International Nuclear Information System (INIS)

    Chang, K.M.; Klahn, D.H.; Morris, J.W. Jr.

    1980-08-01

    Research toward high-strength, high toughness nonmagnetic steels for use in the retaining rings of large electrical generators led to the development of a Ta-modified iron-based superalloy (Fe-36 Ni-3 Ti-3 Ta-0.5 Al-1.3 Mo-0.3 V-0.01 B) which combines high strength with good toughness after suitable aging. The alloy did, however, show some degradation in fatigue resistance in gaseous hydrogen. This sensitivity was associated with a deformation-induced martensitic transformation near the fracture surface. The addition of a small amount of chromium to the alloy suppressed the martensite transformation and led to a marked improvement in hydrogen resistance

  7. Carburization of austenitic alloys by gaseous impurities in helium

    International Nuclear Information System (INIS)

    Lai, G.Y.; Johnson, W.R.

    1980-03-01

    The carburization behavior of Alloy 800H, Inconel Alloy 617 and Hastelloy Alloy X in helium containing various amounts of H 2 , CO, CH 4 , H 2 O and CO 2 was studied. Corrosion tests were conducted in a temperature range from 649 to 1000 0 C (1200 to 1832 0 F) for exposure time up to 10,000 h. Four different helium environments, identified as A, B, C, and D, were investigated. Concentrations of gaseous impurities were 1500 μatm H 2 , 450 μatm CO, 50 μatm CH 4 and 50 μatm H 2 O for Environment A; 200 μatm H 2 , 100 μatm CO, 20 μatm CH 4 , 50 μatm H 2 O and 5 μatm CO 2 for Environment B; 500 μatm H 2 , 50 μatm CO, 50 μatm CH 4 and 2 O for Environment C; and 500 μatm H 2 , 50 μatm CO, 50 μatm CH 4 and 1.5 μatm H 2 O for Environment D. Environments A and B were characteristic of high-oxygen potential, while C and D were characteristic of low-oxygen potential. The results showed that the carburization kinetics in low-oxygen potential environments (C and D) were significantly higher, approximately an order of magnitude higher at high temperatures, than those in high-oxygen potential environments (A and B) for all three alloys. Thermodynamic analyses indicated no significant differences in the thermodynamic carburization potential between low- and high-oxygen potential environments. It is thus believed that the enhanced carburization kinetics observed in the low-oxygen potential environments were related to kinetic effects. A qualitatively mechanistic model was proposed to explain the enhanced kinetics. The present results further suggest that controlling the oxygen potential of the service environment can be an effective means of reducing carburization of alloys

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

    Energy Technology Data Exchange (ETDEWEB)

    Delia, M.; Alaalam, M.; Grech, M. [Univ. of Malta (Malta). Dept. of Metallurgy and Materials

    1998-04-01

    The effect of austenitizing conditions on the microstructure and impact properties of an austempered ductile iron (ADI) containing 1.6% Cu and 1.6% Ni as the main alloying elements was investigated. Impact tests were carried out on samples of initially ferritic matrix structure and which had been first austenitized at 850, 900, 950, and 1,000 C for 15 to 360 min and austempered at 360 C for 180 min. Results showed that the austenitizing temperature, T{sub {gamma}}, and time, t{sub {gamma}} have a significant effect on the impact properties of the alloy. This has been attributed to the influence of these variables on the carbon kinetics. Microstructures of samples austenitized at 950 and 1,000 C contain no pro-eutectoid ferrite. The impact properties of the former structures are independent of t{sub {gamma}}, while those solution treated at 1,000 C are generally low and show wide variation over the range of soaking time investigated. For fully ausferritic structures, impact properties fall with an increase in T{sub {gamma}}. This is particularly evident at 1,000 C. As the T{sub {gamma}} increases, the amount of carbon dissolved in the original austenite increases. This slows down the rate of austenite transformation and results in coarser structures with lower mechanical properties. Optimum impact properties are obtained following austenitizing between 900 and 950 C for 120 to 180 min.

  9. Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

    Energy Technology Data Exchange (ETDEWEB)

    Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao

    1996-04-01

    SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young`s modulus lowers. In addition, its composition elements have the large (n,{alpha}) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

  10. Improving of wear resistance of alloys with metastable austenite structure in abrasion wearing

    International Nuclear Information System (INIS)

    Popov, V.S.; Brykov, N.N.; Pugachev, G.A.

    1979-01-01

    The effect of grain composition of abrasive masses upon the wear resistance of alloys having a metastable austenitic structure is studied. The investigations have been carried out on Kh12F1 steel, using diffraction and hardness measurements and the metallographic analysis. Experimental data indicate that the specific wear of the stable alloys increases substantially with the size of abrasive particles. As regards the metastable alloys, the increase in the size of the abrasive grains has little effect upon the specific wear, as the increasing abradability of the grains is compensated for by the strengthening of the rubbing surfaces, this resulting from the ability of the metal surface layer to underao structure transformations in the course of wear

  11. Swelling in neutron-irradiated titanium alloys

    International Nuclear Information System (INIS)

    Peterson, D.T.

    1982-04-01

    Immersion density measurements have been performed on a series of titanium alloys irradiated in EBR-II to a fluence of 5 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 and 550 0 C. The materials irradiated were the near-alpha alloys Ti-6242S and Ti-5621S, the alpha-beta alloy Ti-64, and the beta alloy Ti-38644. Swelling was observed in all alloys with the greater swelling being observed at 550 0 C. Microstructural examination revealed the presence of voids in all alloys. Ti-38644 was found to be the most radiation resistant. Ti-6242S and Ti-5621S also displayed good radiation resistance, whereas considerable swelling and precipitation were observed in Ti-64 at 550 0 C

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

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Garner, F.; Okita, T.

    2007-01-01

    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 -9 to 4x10 -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)

  13. Influence of Hold Time on Creep-Fatigue Behavior of an Advanced Austenitic Alloy

    International Nuclear Information System (INIS)

    Carroll, Mark; Carroll, Laura

    2011-01-01

    An advanced austenitic alloy, HT-UPS (high temperature-ultrafine precipitate strengthened), is a candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS provides improved creep resistance through a composition based on 316 stainless steel (SS) with additions of Ti and Nb to form nano-scale MC precipitates in the austenitic matrix. The low cycle fatigue and creep-fatigue behavior of a HT-UPS alloy has been investigated at 650 C, 1.0% total strain, and an R ratio of -1 with hold times as long as 9000 sec at peak tensile strain. The cyclic deformation response of HT-UPS is compared to that of 316 SS. The cycles to failure are similar, despite differences in peak stress profiles and the deformed microstructures. Cracking in both alloys is transgranular (initiation and propagation) in the case of continuous cycle fatigue, while the primary cracks also propagate transgranularly during creep-fatigue cycling. Internal grain boundary damage as a result of the tensile hold is present in the form of fine cracks for hold times of 3600 sec and longer and substantially more internal cracks are visible in 316 SS than HT-UPS. The dislocation substructures observed in the deformed material are different. An equiaxed cellular structure is observed in 316 SS, whereas tangles of dislocations are present at the nanoscale MC precipitates in HT-UPS and no cellular substructure is observed.

  14. Mechanical properties microstructure correlation in neutron irradiated heat-affected zones of austenitic stainless steels

    Science.gov (United States)

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

    2007-05-01

    The effects of neutron irradiation on austenitic stainless steels, usually used for the manufacturing of internal elements of nuclear reactors (e.g. the core shrouds), are the alteration of the microstructure, and, as a consequence, of the mechanical properties. The present study is aimed at extending knowledge upon the impact of neutron-irradiation on the heat-affected zone of welded materials, which was influenced by the thermal cycles upon fusion welding. An austenitic stainless steel weld type AISI 304 from a decommissioned experimental pressurised water reactor has been used in the present study. The welded material has been irradiated during 11 reactor cycles to a maximum dpa dose of 0.35 and a temperature of around 573 K. The mechanical properties and microstructure are determined on specimens from heat-affected zone and base materials, with different dose levels. The mechanical properties were determined by performing tensile tests on small flat specimens at two deformation temperatures: room temperature and about 573 K. The characterisation of the microstructure was made by transmission electron microscopy. The correlation between mechanical properties and microstructure after neutron irradiation is made using the dispersed obstacle hardening model. It was found that the measured radiation hardening cannot be explained solely by the presence of the irradiation-induced defects observed in TEM. Smaller irradiation-induced features not resolvable in TEM may also contribute to radiation hardening.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

    2007-01-01

    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 ∼ 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)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  19. 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.

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

    International Nuclear Information System (INIS)

    Carvalho e Camargo, M.U. de; Lucki, G.

    1979-01-01

    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) [pt

  1. Microstructural design of PCA austenitic stainless steel for improved resistance to helium embrittlement under HFIR irradiation

    International Nuclear Information System (INIS)

    Maziasz, P.J.; Braski, D.N.

    1983-01-01

    Several variants of Prime Candidate Alloy (PCA) with different preirradiation thermal-mechanical treatments were irradiated in HFIR and were evaluated for embrittlement resistance via disk-bend tensile testing. Comparison tests were made on two heats of 20%-cold-worked type 316 stainless steel. None of the alloys were brittle after irradiation at 300 to 400 0 C to approx. 44 dpa and helium levels of 3000 to approx.3600 at. ppm. However, all were quite brittle after similar exposure at 600 0 C. Embrittlement varied with alloy and pretreatment for irradiation to 44 dpa at 500 0 C and to 22 dpa at 600 0 C. Better relative embrittlement resistance among PCA variants was found in alloys which contained prior grain boundary MC carbide particles that remained stable under irradiation

  2. Influence of Short Austenitization Treatments on the Mechanical Properties of Low-Alloy Steels for Hot Forming Applications

    Science.gov (United States)

    Holzweissig, Martin Joachim; Lackmann, Jan; Konrad, Stefan; Schaper, Mirko; Niendorf, Thomas

    2015-07-01

    The current work elucidates an improvement of the mechanical properties of tool-quenched low-alloy steel by employing extremely short austenitization durations utilizing a press heating arrangement. Specifically, the influence of different austenitization treatments—involving austenitization durations ranging from three to 15 seconds—on the mechanical properties of low-alloy steel in comparison to an industrial standard furnace process was examined. A thorough set of experiments was conducted to investigate the role of different austenitization durations and temperatures on the resulting mechanical properties such as hardness, bending angle, tensile strength, and strain at fracture. The most important finding is that the hardness, the bending angle as well as the tensile strength increase with shortened austenitization durations. Furthermore, the ductility of the steels exhibits almost no difference following the short austenitization durations and the standard furnace process. The enhancement of the mechanical properties imposed by the short heat treatments investigated, is related to a refinement of microstructural features as compared to the standard furnace process.

  3. Fundamental flow and fracture analysis of prime candidate alloy (PCA) for path a (austenitics)

    International Nuclear Information System (INIS)

    Lucas, G.E.; Jayakumar, M.; Maziasz, P.J.

    1982-01-01

    Room temperature microhardness tests have been performed on samples of Prime Candidate Alloy (PCA) for the austenitics (Path A) subjected to various thermomechanical treatments (TMT). The TMTs have effected various microstructures, which have been well characterized by optical metallography and TEM. For comparison, microhardness tests have been performed on samples of N-lot, DO heat and MFE 316 stainless steel with similar TMTs. The results indicate that the TMTs investigated can significantly alter the microhardness of the PCA in a manner which is consistent with microstructural changes. Moreover, while PCA had the lowest microhardness of the four alloys types after cold working, its microhardness increased while the others decreased to comparable values after aging for 2 h at 750 0 C

  4. Influence of Ti(C,N precipitates on austenite growth of micro-alloyed steel during continuous casting

    Directory of Open Access Journals (Sweden)

    Liu Yang

    2017-11-01

    Full Text Available Austenite grain size is an important influence factor for ductility of steel at high temperatures during continuous casting. Thermodynamic and kinetics calculations were performed to analyze the characteristics of Ti(C,N precipitates formed during the continuous casting of micro-alloyed steel. Based on Andersen-Grong equation, a coupling model of second phase precipitation and austenite grain growth has been established, and the influence of second precipitates on austenite grain growth under different cooling rates is discussed. Calculations show that the final sizes of austenite grains are 2.155, 1.244, 0.965, 0.847 and 0.686 mm, respectively, under the cooling rate of 1, 3, 5, 7, and 10 ℃·s-1, when ignoring the pinning effect of precipitation on austenite growth. Whereas, if taking the pinning effect into consideration, the grain growth remains stable from 1,350 ℃, the calculated final sizes of austenite grains are 1.46, 1.02, 0.80, 0.67 and 0.57 mm, respectively. The sizes of final Ti(C,N precipitates are 137, 79, 61, 51 and 43 nm, respectively, with the increase of cooling rate from 1 to 10 ℃·s-1. Model validation shows that the austenite size under different cooling rates coincided with the calculation results. Finally, the corresponding measures to strengthen cooling intensity at elevated temperature are proposed to improve the ductility and transverse crack of slab.

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

    International Nuclear Information System (INIS)

    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). In addition, several phases develop at significantly lower temperaures during neutron irradiation (radiation-enhanced phases). 18 figures, 9 tables

  6. 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.

  7. Ordering in alloys under electron irradiation

    International Nuclear Information System (INIS)

    Tendoloo, G. van; Amelinckx, S.

    1981-01-01

    Different alloys with a face centered cubic disordered structure have been electron irradiated in the quenched-in short range ordered state using a high voltage electron microscope. Care has been taken to avoid ordering due to thermal effects of beam heating. The influence of the irradiation temperature has been illustrated for Ni 4 Mo. Ordering due to 1 MeV irradiation has been observed for Au 4 Mn, Ni 4 Mo and Cu 3 Pd. The diffuse intensity associated with the irradiation ordering can be interpreted in function of predominant clusters occuring in the transition state between short range order and long range order

  8. Microstructural observation of ion-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Sawai, T.; Hamada, S.; Hishinuma, A.

    1992-01-01

    Type 316 stainless steel, base metal and weld metal obtained from an electron beam weld joint, was irradiated with 90 MeV Br +6 in the JAERI tandem accelerator. Cross-sectional TEM specimens were obtained by nickel plating. Microstructural observation revealed a band of tiny dislocation loops was observed around the mean projected range and the measured distance from the surface was 6.75±0.15μm. This is slightly larger than the calculated value using Ziegler's stopping power. Defect clusters were also observed around defect sinks within the mean projected range, suggesting cascade-sink interaction. These sinks are the grain boundary in the base metal specimen and preexisting dislocation lines in the weld metal specimen. Surface roughness of polished specimen was detected at the shallower side of the peak damage band, although no visible crystalline defect cluster was observed. This suggests radiation-induced microchemical evolution prior to sever microstructural evolution. (author)

  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...

  10. Effect of phosphorus on the swelling and precipitation behavior of austenitic stainless steels during irradiation

    International Nuclear Information System (INIS)

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

    1983-01-01

    It has been observed that increasing the volume fraction of the needle-shaped iron phosphide phase in austenitic stainless steels tends to inhibit void swelling during neutron irradiation. An earlier analysis showed that this effect could not be accounted for in terms of enhanced point defect recombination at particle-matrix interfaces. The behavior of the iron phosphide phase has been further examined using dual ion beam irradiations. It was found that the particle-matrix interface serves as a site for the nucleation of a very fine dispersion of helium bubbles. It is thought that since a high number density of cavities lowers the number of helium atoms per cavity, the irradiation time for the cavities to accumulate the critical number of gas atoms for bias-driven growth is correspondingly increased. Although the phosphide phase nucleates rapidly, it eventually undergoes dissolution if either the G or Laves phase develops with increasing dose

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Science.gov (United States)

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

    2007-02-01

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

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

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Futakawa, Masatoshi; Nanjyo, Yoshiyasu; Kiuchi, Kiyoshi; Anegawa, Takefumi

    2003-01-01

    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(ε 0 + ε) n , A: strength coefficient, ε 0 : 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)

  14. 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.

  15. Void formation in irradiated binary nickel alloys

    International Nuclear Information System (INIS)

    Shaikh, M.A.; Ahmed, M.; Akhter, J.I.

    1994-01-01

    In this work a computer program has been used to compute void radius, void density and swelling parameter for nickel and binary nickel-carbon alloys irradiated with nickel ions of 100 keV. The aim is to compare the computed results with experimental results already reported

  16. Manufacturing and characterization of Ni-free N-containing ODS austenitic alloy

    Science.gov (United States)

    Mori, A.; Mamiya, H.; Ohnuma, M.; Ilavsky, J.; Ohishi, K.; Woźniak, Jarosław; Olszyna, A.; Watanabe, N.; Suzuki, J.; Kitazawa, H.; Lewandowska, M.

    2018-04-01

    Ni-free N-containing oxide dispersion strengthened (ODS) austenitic alloys were manufactured by mechanical alloying (MA) followed by spark plasma sintering (SPS). The phase evolutions during milling under a nitrogen atmosphere and after sintering were studied by X-ray diffraction (XRD). Transmission electron microcopy (TEM) and alloy contrast variation analysis (ACV), including small-angle neutron scattering (SANS) and ultra-small-angle X-ray scattering (USAXS), revealed the existence of nanoparticles with a diameter of 3-51 nm for the samples sintered at 950 °C. Sintering at 1000 °C for 5 and 15 min caused slight growth and a significant coarsening of the nanoparticles, up to 70 nm and 128 nm, respectively. The ACV analysis indicated the existence of two populations of Y2O3, ε-martensite and MnO. The dispersive X-ray spectrometry (EDS) confirmed two kinds of nanoparticles, Y2O3 and MnO. The material was characterized by superior micro-hardness, of above 500 HV0.1.

  17. Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hug, E., E-mail: eric.hug@ensicaen.fr [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Prasath Babu, R. [School of Materials, University of Manchester, M13 9PL (United Kingdom); Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Monnet, I. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Etienne, A. [Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Moisy, F. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Pralong, V. [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Enikeev, N. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); Saint Petersburg State University, Laboratory of the Mechanics of Bulk Nanostructured Materials, 198504 St. Petersburg (Russian Federation); Abramova, M. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); and others

    2017-01-15

    Highlights: • Impacts of nanostructuration and irradiation on the properties of 316 stainless steels are reported. • Irradiation of nanostructured samples implies chromium depletion as than depicted in coarse grain specimens. • Hardness of nanocrystalline steels is only weakly affected by irradiation. • Corrosion resistance of the nanostructured and irradiated samples is less affected by the chromium depletion. - Abstract: The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV {sup 56}Fe{sup 5+} ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

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

    International Nuclear Information System (INIS)

    Han, X.

    2012-01-01

    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) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.B.; Solly, B.

    1966-12-15

    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.

  20. Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi

    1987-09-01

    The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

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

    International Nuclear Information System (INIS)

    Vilpas, M.; Haenninen, H.

    1999-01-01

    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.)

  2. 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)

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

    Science.gov (United States)

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

    2007-03-01

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

  4. The natural aging of austenitic stainless steels irradiated with fast neutrons

    Science.gov (United States)

    Rofman, O. V.; Maksimkin, O. P.; Tsay, K. V.; Koyanbayev, Ye. T.; Short, M. P.

    2018-02-01

    Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens "naturally aged" in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10-25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials "on the shelf" do not evolve.

  5. Elastic properties of paramagnetic austenitic steel at finite temperature: Longitudinal spin fluctuations in multicomponent alloys

    Science.gov (United States)

    Dong, Zhihua; Schönecker, Stephan; Chen, Dengfu; Li, Wei; Long, Mujun; Vitos, Levente

    2017-11-01

    We propose a first-principles framework for longitudinal spin fluctuations (LSFs) in disordered paramagnetic (PM) multicomponent alloy systems and apply it to investigate the influence of LSFs on the temperature dependence of two elastic constants of PM austenitic stainless steel Fe15Cr15Ni. The magnetic model considers individual fluctuating moments in a static PM medium with first-principles-derived LSF energetics in conjunction with describing chemical disorder and randomness of the transverse magnetic component in the single-site alloy formalism and disordered local moment (DLM) picture. A temperature-sensitive mean magnetic moment is adopted to accurately represent the LSF state in the elastic-constant calculations. We make evident that magnetic interactions between an LSF impurity and the PM medium are weak in the present steel alloy. This allows gaining accurate LSF energetics and mean magnetic moments already through a perturbation from the static DLM moments instead of a tedious self-consistent procedure. We find that LSFs systematically lower the cubic shear elastic constants c' and c44 by ˜6 GPa in the temperature interval 300-1600 K, whereas the predominant mechanism for the softening of both elastic constants with temperature is the magneto-volume coupling due to thermal lattice expansion. We find that non-negligible local magnetic moments of Cr and Ni are thermally induced by LSFs, but they exert only a small influence on the elastic properties. The proposed framework exhibits high flexibility in accurately accounting for finite-temperature magnetism and its impact on the mechanical properties of PM multicomponent alloys.

  6. Low Temperature Diffusion Transformations in Fe-Ni-Ti Alloys During Deformation and Irradiation

    Science.gov (United States)

    Sagaradze, Victor; Shabashov, Valery; Kataeva, Natalya; Kozlov, Kirill; Arbuzov, Vadim; Danilov, Sergey; Ustyugov, Yury

    2018-03-01

    The deformation-induced dissolution of Ni3Ti intermetallics in the matrix of austenitic alloys of Fe-36Ni-3Ti type was revealed in the course of their cascade-forming neutron irradiation and cold deformation at low temperatures via employment of Mössbauer method. The anomalous deformation-related dissolution of the intermetallics has been explained by the migration of deformation-induced interstitial atoms from the particles into a matrix in the stress field of moving dislocations. When rising the deformation temperature, this process is substituted for by the intermetallics precipitation accelerated by point defects. A calculation of diffusion processes has shown the possibility of the realization of the low-temperature diffusion of interstitial atoms in configurations of the crowdions and dumbbell pairs at 77-173 K. The existence of interstitial atoms in the Fe-36Ni alloy irradiated by electrons or deformed at 77 K was substantiated in the experiments of the electrical resistivity measurements.

  7. Effect of composition on corrosion resistance of high-alloy austenitic stainless steel weld metals

    International Nuclear Information System (INIS)

    Marshall, P.I.; Gooch, T.G.

    1993-01-01

    The corrosion resistance of stainless steel weld metal in the ranges of 17 to 28% chromium (Cr), 6 to 60% nickel (Ni), 0 to 9% molybdenum (Mo), and 0.0 to 0.37% nitrogen (N) was examined. Critical pitting temperatures were determined in ferric chloride (FeCl 3 ). Passive film breakdown potentials were assessed from potentiodynamic scans in 3% sodium chloride (NaCl) at 50 C. Potentiodynamic and potentiostatic tests were carried out in 30% sulfuric acid (H 2 SO 4 ) ar 25 C, which was representative of chloride-free acid media of low redox potential. Metallographic examination and microanalysis were conducted on the test welds. Because of segregation of alloying elements, weld metal pitting resistance always was lower than that of matching composition base steel. The difference increased with higher Cr, Mo, and N contents. Segregation also reduced resistance to general corrosion in H 2 SO 4 , but the effect relative to the base steel was less marked than with chloride pitting. Segregation of Cr, Mo, and N in fully austenitic deposits decreased as the Ni' eq- Cr' eq ratio increased. Over the compositional range studied, weld metal pitting resistance was dependent mainly on Mo content and segregation. N had less effect than in wrought alloys. Both Mo and N enhanced weld metal corrosion resistance in H 2 SO 4

  8. Accelerated irradiation growth of zirconium alloys

    International Nuclear Information System (INIS)

    Griffiths, M.; Gilbert, R.W.; Fidleris, V.

    1989-01-01

    This paper discusses how sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 10 25 n · m -2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or breakaway growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature. Transmission electron microscopy has shown that there is a correlation between accelerated irradiation growth and the appearance of c-component vacancy loops on basal planes. Measurements in some specimens indicate that a significant fraction of the strain can be directly attributed to the loops themselves. There is considerable evidence to show that their formation is dependent both on the specimen purity and on the irradiation temperature. Materials that have a high interstitial-solute content contain c-component loops and exhibit high growth rates even at low fluences ( 2 :5 n · m -2 , E > 1 MeV). For sponge zirconium and the Zircaloys, c-component loop formation and the associated acceleration of growth (breakaway) during irradiation occurs because the intrinsic interstitial solute (mainly, oxygen, carbon and nitrogen) in the zirconium matrix is supplemented by interstitial iron, chromium, and nickel from the radiation-induced dissolution of precipitates. (author)

  9. Atom redistribution and multilayer structure in NiTi shape memory alloy induced by high energy proton irradiation

    Science.gov (United States)

    Wang, Haizhen; Yi, Xiaoyang; Zhu, Yingying; Yin, Yongkui; Gao, Yuan; Cai, Wei; Gao, Zhiyong

    2017-10-01

    The element distribution and surface microstructure in NiTi shape memory alloys exposed to 3 MeV proton irradiation were investigated. Redistribution of the alloying element and a clearly visible multilayer structure consisting of three layers were observed on the surface of NiTi shape memory alloys after proton irradiation. The outermost layer consists primarily of a columnar-like TiH2 phase with a tetragonal structure, and the internal layer is primarily comprised of a bcc austenite phase. In addition, the Ti2Ni phase, with an fcc structure, serves as the transition layer between the outermost and internal layer. The above-mentioned phenomenon is attributed to the preferential sputtering of high energy protons and segregation induced by irradiation.

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

    Directory of Open Access Journals (Sweden)

    A. Janus

    2011-04-01

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

  11. The influence of nitrogen, phosphorus, sulphur and nickel on the stress corrosion cracking of austenitic Fe-Ni-Cr alloys

    International Nuclear Information System (INIS)

    Cihal, V.

    1985-01-01

    From the results of the stress corrosion cracking tests it is evident that austenitic alloys with a phosphorus content 0.01% causes a strong increase in stress corrosion cracking susceptibility of alloys with a nickel content in the range 33 to 38%. With a nickel content of approx. 35%, an increase of nitrogen concentration to approx. 0.15% also produces a significant effect on stress corrosion cracking susceptibility. A sulphur content up to 0.033% does not produce a significant effect on stress corrosion cracking. (author)

  12. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part II. Fatigue crack growth rate

    Energy Technology Data Exchange (ETDEWEB)

    Margolin, B., E-mail: margolinbz@yandex.ru; Minkin, A.; Smirnov, V.; Sorokin, A.; Shvetsova, V.; Potapova, V.

    2016-11-15

    The experimental data on the fatigue crack growth rate (FCGR) have been obtained for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various radiation swelling. The performed study of the fracture mechanisms for cracked specimens under cyclic loading has explained why radiation swelling affects weakly FCGR unlike its effect on fracture toughness. Mechanical modeling of fatigue crack growth has been carried out and the dependencies for prediction of FCGR in irradiated austenitic steel with and with no swelling are proposed and verified with the obtained experimental results. As input data for these dependencies, FCGR for unirradiated steel and the tensile mechanical properties for unirradiated and irradiated steels are used.

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

    International Nuclear Information System (INIS)

    Abramov, V.Ya.; Ionajtis, R.R.; Kotov, V.V.; Loguntsev, E.N.; Ushakov, V.P.

    1996-01-01

    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 10 19 - 10 20 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 [ru

  14. Hydrogen-Induced Delayed Cracking in TRIP-Aided Lean-Alloyed Ferritic-Austenitic Stainless Steels.

    Science.gov (United States)

    Papula, Suvi; Sarikka, Teemu; Anttila, Severi; Talonen, Juho; Virkkunen, Iikka; Hänninen, Hannu

    2017-06-03

    Susceptibility of three lean-alloyed ferritic-austenitic stainless steels to hydrogen-induced delayed cracking was examined, concentrating on internal hydrogen contained in the materials after production operations. The aim was to study the role of strain-induced austenite to martensite transformation in the delayed cracking susceptibility. According to the conducted deep drawing tests and constant load tensile testing, the studied materials seem not to be particularly susceptible to delayed cracking. Delayed cracks were only occasionally initiated in two of the materials at high local stress levels. However, if a delayed crack initiated in a highly stressed location, strain-induced martensite transformation decreased the crack arrest tendency of the austenite phase in a duplex microstructure. According to electron microscopy examination and electron backscattering diffraction analysis, the fracture mode was predominantly cleavage, and cracks propagated along the body-centered cubic (BCC) phases ferrite and α'-martensite. The BCC crystal structure enables fast diffusion of hydrogen to the crack tip area. No delayed cracking was observed in the stainless steel that had high austenite stability. Thus, it can be concluded that the presence of α'-martensite increases the hydrogen-induced cracking susceptibility.

  15. Hydrogen-Induced Delayed Cracking in TRIP-Aided Lean-Alloyed Ferritic-Austenitic Stainless Steels

    Directory of Open Access Journals (Sweden)

    Suvi Papula

    2017-06-01

    Full Text Available Susceptibility of three lean-alloyed ferritic-austenitic stainless steels to hydrogen-induced delayed cracking was examined, concentrating on internal hydrogen contained in the materials after production operations. The aim was to study the role of strain-induced austenite to martensite transformation in the delayed cracking susceptibility. According to the conducted deep drawing tests and constant load tensile testing, the studied materials seem not to be particularly susceptible to delayed cracking. Delayed cracks were only occasionally initiated in two of the materials at high local stress levels. However, if a delayed crack initiated in a highly stressed location, strain-induced martensite transformation decreased the crack arrest tendency of the austenite phase in a duplex microstructure. According to electron microscopy examination and electron backscattering diffraction analysis, the fracture mode was predominantly cleavage, and cracks propagated along the body-centered cubic (BCC phases ferrite and α’-martensite. The BCC crystal structure enables fast diffusion of hydrogen to the crack tip area. No delayed cracking was observed in the stainless steel that had high austenite stability. Thus, it can be concluded that the presence of α’-martensite increases the hydrogen-induced cracking susceptibility.

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

    Energy Technology Data Exchange (ETDEWEB)

    Brady, M.P.; Matthews, W.J. (Capstone Turbine Corp.)

    2010-09-15

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

  17. 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)

    Huebner, R.; Ehrlich, K.

    1998-01-01

    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 dpa NRT at 420 deg. C, 81 dpa NRT 500 deg. C and 61 dpa NRT 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 dpa NRT at 420 deg. C. The formerly observed inter correlation between swelling and in-pile creep is confirmed up to 106 dpa NRT . 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 dpa NRT and 120 MPa hoop

  18. Defect clustering in concentrated alloys during irradiation

    International Nuclear Information System (INIS)

    Hashimoto, T.; Shigenaka, N.; Fuse, M.

    1992-01-01

    A rate theory based model is presented to investigate the kinetics of interstitial clustering processes in a face-centered cubic (fcc) binary alloy containing A- and B-atoms. Three types of interstitial dumbbells, AA-, BB- and AB-type dumbbells, are considered. Conversions between these interstitial dumbbells are explicitly introduced into the formulation, based on the consideration of dumbbell configurations and movements. A di- interstitial is assumed to be the nucleus of a dislocation loop. Reactions of point defect production by irradiation, mutual recombination of an interstitial and a vacancy, dislocation loop nucleation and their growth are included in the model. Parameter values are chosen based on the atom size of the alloy elements, and dislocation loop formation kinetics are investigated while varying alloy compositions. Two different types of kinetics are obtained in accordance with the dominant loop nucleus types. Conversions between interstitial dumbbells are important in the determination of the interstitial dumbbell concentration ratios, of the dominant nucleus types, and consequently, the loop formation kinetics. Dislocation loop concentration decreases with increasing undersized atom content, but dose rate and temperature dependence of loop concentration are insensitive to alloy compositions. (author)

  19. Diffusive Phenomena and the Austenite/Martensite Relative Stability in Cu-Based Shape-Memory Alloys

    Science.gov (United States)

    Pelegrina, J. L.; Yawny, A.; Sade, M.

    2018-02-01

    The main characteristic of martensitic phase transitions is the coordinate movement of the atoms which takes place athermally, without the contribution of diffusion during its occurrence. However, the impacts of diffusive phenomena on the relative stability between the phases involved and, consequently, on the associated transformation temperatures and functional properties can be significant. This is particularly evident in the case of Cu-based shape-memory alloys where atomic diffusion in both austenite and martensite metastable phases might occur even at room-temperature levels, giving rise to a variety of intensively studied phenomena. In the present study, the progresses made in the understanding of three selected diffusion-related effects of importance in Cu-Zn-Al and Cu-Al-Be alloys are reviewed. They are the after-quench retained disorder in the austenitic structure and its subsequent reordering, the stabilization of the martensite, and the effect of applied stress on the austenitic order. It is shown how the experimental results obtained from tests performed on single crystal material can be rationalized under the shed of a model developed to evaluate the variation of the relative stability between the phases in terms of atom pairs interchanges.

  20. Ab initio investigation of the surface properties of austenitic Fe-Ni-Cr alloys in aqueous environments

    Energy Technology Data Exchange (ETDEWEB)

    Rák, Zs., E-mail: zrak@ncsu.edu; Brenner, D.W.

    2017-04-30

    Highlights: • The trend in the surface energies of austenitic stainless steels is: (111) < (100) < (110). • On the (111) orientation Ni segregates to the surface and Cr segregates into the bulk. • The surface stability of the alloys in contact with water decrease with temperature and pH. - Abstract: The surface energetics of two austenitic stainless steel alloys (Type 304 and 316) and three Ni-based alloys (Alloy 600, 690, and 800) are investigated using theoretical methods within the density functional theory. The relative stability of the low index surfaces display the same trend for all alloys; the most closely packed orientation and the most stable is the (111), followed by the (100) and the (110) surfaces. Calculations on the (111) surfaces using various surface chemical and magnetic configurations reveal that Ni has the tendency to segregate toward the surface and Cr has the tendency to segregate toward the bulk. The magnetic frustration present on the (111) surfaces plays an important role in the observed segregation tendencies of Ni and Cr. The stability of the (111) surfaces in contact with aqueous solution are evaluated as a function of temperature, pH, and concentration of aqueous species. The results indicate that the surface stability of the alloys decrease with temperature and pH, and increase slightly with concentration. Under conditions characteristic to an operating pressurized water reactor, the Ni-based alloy series appears to be of better quality than the stainless steel series with respect to corrosion resistance and release of aqueous species when in contact with aqueous solutions.

  1. Corrosion behaviour of austenitic stainless steel, nickel-base alloy and its weldments in aqueous LiBr solutions

    Energy Technology Data Exchange (ETDEWEB)

    Blasco-Tamarit, E.; Igual-Munoz, A.; Garcia Anton, J.; Garcia-Garcia, D. [Departamento de Ingenieria Quimica y Nuclear. E.T.S.I.Industriales, Universidad Politecnica de Valencia, P.O. Box 22012 E-46071 Valencia (Spain)

    2004-07-01

    With the advances in materials production new alloys have been developed, such as High- Alloy Austenitic Stainless Steels and Nickel-base alloys, with high corrosion resistance. These new alloys are finding applications in Lithium Bromide absorption refrigeration systems, because LiBr is a corrosive medium which can cause serious corrosion problems, in spite of its favourable properties as absorbent. The objective of the present work was to study the corrosion resistance of a highly alloyed austenitic stainless steel (UNS N08031) used as base metal, a Nickel-base alloy (UNS N06059) used as its corresponding filler metal, and the weld metal obtained by the Gas Tungsten Arc Welding (GTAW) procedure. The materials have been tested in different LiBr solutions (400 g/l, 700 g/l, 850 g/l and a commercial 850 g/l LiBr heavy brine containing Lithium Chromate as corrosion inhibitor), at 25 deg. C. Open Circuit Potential tests and potentiodynamic anodic polarization curves have been carried out to obtain information about the general electrochemical behaviour of the materials. The polarization curves of all the alloys tested were typical of passivable materials. Pitting corrosion susceptibility has been evaluated by means of cyclic potentiodynamic curves, which provide parameters to analyse re-passivation properties. The galvanic corrosion generated by the electrical contact between the welded and the base material has been estimated from the polarization diagrams according to the Mixed Potential Method. Samples have been etched to study the microstructure by Scanning Electron Microscopy (SEM). The results demonstrate that the pitting resistance of all these materials increases as the LiBr concentration decreases. In general, the presence of chromate tended to shift the pitting potential to more positive values than those obtained in the 850 g/l LiBr solution. (authors)

  2. Alloy development for irradiation performance: program strategy

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  4. Irradiation growth in zirconium alloys: a review

    International Nuclear Information System (INIS)

    Fidleris, V.

    1980-09-01

    The change in shape during irradiation without external stress, irradiation growth, was first discovered in uranium and later in graphite, zirconium and other core materials which exhibit anisotropic physical properties. The direction of maximum growth of metals invariably corresponds with the direction of minimum thermal expansion. In polycrystalline zirconium alloys growth is positive in the direction of maximum deformation during fabrication and in other directions it can be either positive or negative depending on the preferred orientation of grains (crystallographic texture). Growth increases gradually with temperature between 300 K and 620 K and rapidly with fluence up to about 1 x 10 25 n.m. -2 (Eμ1 MeV). At higher fluences the growth appears to saturate in annealed materials and reach a steady rate approximately proportional to dislocation density in cold-worked materials. Above 600 K both annealed and cold-worked materials have similar steady growth rates. Irradiation growth is caused by the segregation to different sinks of the vacancies and interstitials generated by irradiation, but the dominant types of sinks for each type of point defect and the mode of transport of the point defects to sinks cannot therefore be predicted theoretically. For the purpose of designing reactor core components empirical equations have been derived that can satisfactorily predict the steady state growth behaviour from texture and microstructure. (auth)

  5. Self-irradiation study of plutonium alloys

    International Nuclear Information System (INIS)

    Oudot, B.

    2005-02-01

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

  6. Nucleation and swelling in electron irradiated austenitic stainless steels in temperature range 400∼720 degree C

    International Nuclear Information System (INIS)

    Qian Jiapu; Lu Liping; Chen Jiming; Sun Jiguang; Zhao Zhuoyong

    1994-10-01

    A study of the influence of temperature on swelling behavior in electron irradiated austenitic stainless steels has been performed with a high voltage electron microscope (HVEM) in the temperature range 400 to 720 degree C. The specimen materials were solution annealed (SA) 316 stainless steel (SS), cold worked (CW) 316 SS and Ti-modified austenitic stainless steel (Ti-mod. SS). The electron energy in HVEM was 1 MeV. The results of mean void density, mean void diameter, void swelling, swelling rate and incubation dose vs. dose and temperature are presented. It is suggested that the irradiation temperature influenced the microstructure, overall sink strength and relative strength of neutral sinks to that of biased sinks in the specimens and so made the void nucleation, void growth and swelling behavior differently in 316 SA, 316 CW and Ti-mod. SS. The influence of He pre-implantation on void nucleation in electron irradiated austenitic stainless steels has also been studied. The experiments indicated that the helium was active void nucleus but had no direct impact on void growth

  7. 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.

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

    International Nuclear Information System (INIS)

    Maziasz, P.J.; Braski, D.N.

    1983-01-01

    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 300 0 C after approx. 44 dpa. At 500 to 600 0 C 25%-cold-worked PCA showed better void swelling resistance than type 316 at approx. 44 dpa. There was less swelling variability among alloys at 400 0 C, 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)

  9. Measurement of irradiation temperature and study of local overheating of claddings of austenitic steel, by X-ray diffraction

    International Nuclear Information System (INIS)

    Rousset, P.; Cadalbert, R.

    1983-01-01

    The development of fuel elements of fast neutron reactors makes the CEA effectuate numerous irradiations of steel in the RAPSODIE and PHENIX reactor. Detailed knowledge and control of the irradiation temperature pose an important problem both for experimental claddings and for fuel claddings used in these reactors. The authors utilize the relation that exists between the irradiation temperature and the failure density present in the austenitic steels, for measuring irradiation temperatures and for studying local overheating of the claddings by means of X-ray diffraction. In effect, the line widths of X-ray diffraction is a measure of the degree of lattice distortion. Quantitative analysis of this failure density enables to study the thermal profiles of the claddings in the assembly; to study local overheating (hot spots); and to measure the irradiation temperature in capsules. (Auth.)

  10. 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.

  11. A preliminary investigation of the initiation of pitting corrosion in austenitic stainless steels and nickel-based alloys

    International Nuclear Information System (INIS)

    Higginson, A.

    1984-01-01

    Pitting corrosion in a number of austenitic stainless steels and nickel-based alloys that differ widely in their resistance to corrosion was studed by electrochemical and electron-optical techniques. The effect of contamination of the sulphuric acid electrolyte by chloride ions was also investigated. Preliminary results for the surface analysis of samples of 316 stainless steel by Auger electron spectroscopy are presented, and suggestions are included for further application of this technique to the examination of pitting corrosion. A comprehensive review of the literature concerning the initiation of pitting corrosion is included

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-08-28

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. Modelling of radiation induced segregation in austenitic Fe alloys at the atomistic level

    International Nuclear Information System (INIS)

    Piochaud, Jean-Baptiste

    2013-01-01

    In pressurized water reactors, under irradiation internal structures are subject of irradiation assisted stress corrosion cracking which is influenced by radiation induced segregation (RIS). In this work RIS of 316 stainless steels is modelled considering a model ternary Fe-10Ni-20Cr alloy. For this purpose we have built an Fe-Ni-Cr pair interaction model to simulate RIS at the atomistic level using an atomistic kinetic Monte Carlo approach. The pair interactions have been deduced from density functional theory (DFT) data available in the pure fcc systems but also from DFT calculations we have performed in the Fe-10Ni-20Cr target alloy. Point defect formation energies were calculated and found to depend strongly on the local environment of the defect. As a consequence, a rather good estimation of these energies can be obtained from the knowledge of the number and respective positions of the Ni and Cr atoms in the vicinity of the defect. This work shows that a model based only on interaction parameters between elements positioned in perfect lattice sites (solute atoms and vacancy) cannot capture alone both the thermodynamic and the kinetic aspect of RIS. A more accurate of estimating the barriers encountered by the diffusing species is required than the one used in our model, which has to depend on the saddle point environment. This study therefore shows thus the need to estimate point defect migration energies using the DFT approach to calibrate a model that can be used in the framework of atomic kinetic Monte Carlo simulations. We also found that the reproduction by our pair interaction model of DFT data for the self-interstitial atoms was found to be incompatible with the modelling of RIS under electron irradiation. (author)

  15. Irradiation temperature determination and study of local overheating of fuel cans in austenitic steel by X-ray diffraction

    International Nuclear Information System (INIS)

    Rousset, P.; Cadalbert, R.

    1982-09-01

    The great advantage is demonstrated of measuring the widths of X ray diffraction lines on fast neutron irradiated austenitic steel cladding, for studying (1) the thermic shape of fuel rods and any overheating that might occur and (2) for measuring irradiation temperatures in capsules. This very sensitive method is easy to use on a counter diffraction meter, in a shielded cell, in conjunction with a calculator and an automatic sample holder, and enables temperature distribution charts to be quickly plotted. It is also extensively applicable to other components of the core such as hexagonal tubes, control rods, etc [fr

  16. 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.

  17. Severe Embrittlement of Neutron Irradiated Austenitic Steels Arising from High Void Swelling

    International Nuclear Information System (INIS)

    Neustroev, V.S.; Garner, F.

    2007-01-01

    Full text of publication follows: Data are presented from BOR-60 irradiations showing that significant radiation-induced swelling causes severe embrittlement in austenitic stainless steels, reducing the service life of structural components. Similar loss of ductility is expected when swelling arises in fusion and light water reactor environments. Above 7-16% swelling there is complete loss of ductility, with the onset of ductility loss beginning at lower swelling in ring-pull tensile tests than for flat tensile specimens. For steels that develop extensive precipitation during irradiation, the critical swelling level is even lower. A model is presented to demonstrate the effect of voids acting alone to produce the embrittlement. Although voids are not very effective hardeners, they are very effective to generate stress concentrations between voids. The stress concentration ratio increases strongly when the void diameter exceeds ∼40% of the void-to-void separation distance. When the volume fraction of voids is rather high (about 16 % and higher), a geometric situation develops where it is possible to create an intense field of deformation glide planes residing at an angle of 45 deg. to the void-to-void axis. Significant localized flow then proceeds on these planes for specimen stress levels that are significantly lower than the yield stress. Voids also segregate nickel to their surfaces such that flow localization occurs in the low-nickel inter-void regions to produce strain-induced martensite, which is further accelerated by stress concentrations at the advancing crack tip, leading to catastrophic failure. (authors)

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Roy, Marion, E-mail: marion.roy@cea.fr [CEA, DEN, DPC, SCCME, Laboratoire d’Etude de la Corrosion Non Aqueuse, F-91191 Gif-sur-Yvette (France); Martinelli, Laure, E-mail: laure.martinelli@cea.fr [CEA, DEN, DPC, SCCME, Laboratoire d’Etude de la Corrosion Non Aqueuse, F-91191 Gif-sur-Yvette (France); Ginestar, Kevin, E-mail: kevin.ginestar@cea.fr [CEA, DEN, DPC, SCCME, Laboratoire d’Etude de la Corrosion Non Aqueuse, F-91191 Gif-sur-Yvette (France); Favergeon, Jérôme, E-mail: jerome.favergeon@utc.fr [Laboratoire Roberval, UMR 7337, Université de Technologie de Compiègne, Centre de Recherche de Royallieu, CS 60319, 60203 Compiègne Cedex (France); Moulin, Gérard [Laboratoire Roberval, UMR 7337, Université de Technologie de Compiègne, Centre de Recherche de Royallieu, CS 60319, 60203 Compiègne Cedex (France)

    2016-01-15

    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{sup −9} and 5 10{sup −4} g kg{sup −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{sup −1} K{sup −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. - Highlights: • 10 austenitic steels and Ni rich alloys were tested in LBE at 520 °C with dissolved oxygen content between 10{sup -9} and 5 10{sup -4} wt%. • It is shown that only thermodynamics cannot explain the Ni rich alloys corrosion behaviour in LBE. • The role of oxygen on corrosion behaviour in LBE was highlighted. • An equilibrium line was defined above which only oxidation has occurred on 316L: RTln[O](wt%) = -57584/T(K)-55.876T(K)+254546. • 18Cr-15Ni-3.7Si, 21Cr-11Ni-1.6Si and 14Cr-25Ni-3.5Al

  20. A new high-strength iron base austenitic alloy with good toughness and corrosion resistance (GE-EPRI alloy-TTL)

    International Nuclear Information System (INIS)

    Ganesh, S.

    1989-01-01

    A new high strength, iron based, austenitic alloy has been successfully developed by GE-EPRI to satisfy the strength and corrosion resistance requirements of large retaining rings for high capacity generators (>840Mw). This new alloy is a modified version of the EPRI alloy-T developed by the University of California, Berkeley, in an earlier EPRI program. It is age hardenable and has the nominal composition (weight %): 34.5 Ni, 5Cr, 3Ti, 1Nb, 1Ta, 1Mo, .5Al, .3V, .01B. This composition was selected based on detailed metallurgical and processing studies on modified versions of alloy-T. These studies helped establish the optimum processing conditions for the new alloy and enabled the successful scale-up production of three large (50-52 inch dia) test rings from a 5,000 lb VIM-VAR billet. The rings were metallurgically sound and exhibited yield strength capabilities in the range 145 to 220 ksi depending on the extent of hot/cold work induced. The test rings met or exceeded all the property goals. The above alloy can provide a good combination of strength, toughness and corrosion resistance and, through an suitable modification of chemistry or processing conditions, could be a viable candidate for high strength LWR internal applications. 3 figs

  1. Neutron diffraction and electron microscopic investigation of decomposition and radiation-induced ageing of Cr-Ni-Ti austenitic alloys

    International Nuclear Information System (INIS)

    Alyab'yev, V.M.; Vologin, V.G.; Dubinin, S.F.; Lapin, S.S.; Parkhomenko, V.D.; Sagaradze, V.V.

    1990-01-01

    The kinetic and structural features of intermetallic ageing of Cr-Ni-Ti austenitic steels are investigated over a wide time interval with neutron diffraction and electron microscopic methods. The size and volume fraction of γ'-particles at the initial (0.25 hr) and later (up to 1000 hr) stages of ageing are determined. From superlattice reflections in neutron diffraction a γ'-phase is identified after irradiation by fast neutrons of stainless steel Kh16N15M3T1, which possesses a high resistance to radiation swelling. (author)

  2. The irradiation behaviour of the austenitic stainless steel DIN 1.4970

    International Nuclear Information System (INIS)

    Huebner, R.

    2000-06-01

    The irradiation behaviour of the austenitic stainless steel DIN 1.4970 (15% Cr, 15% Ni, 1,6% Mn, 1,5% Mo, 0,4 - 1% Si, 0,3 - 0,5% Ti) has been examined in the irradiation experiment PFR-M2. The samples have been irradiated as pressurised capsules in the prototype fast reactor at Dounreay, Scotland, at 420, 500 and 600 C with maximum doses of 106, 81 and 62 dpa NRT . The stress-free and the stress-induced swelling and the irradiation induced creep could be determined by the diameter and length measurements. After the irradiation the density and the mechanical properties have been determined. Additionally the microstructure was investigated with a transmission electron microscope (TEM). All four lots show the maximum amount of stress-free swelling at 420 C. At 600 C no swelling can be detected up to the maximum dose of 62 dpa NRT . The lot with the lowest Si-content exhibits the highest amount of swelling at 420 C as well as at 500 C. Increasing the Si-content from 0.4% to 1% increases at 420 C the incubation dose from 20 to 40 dpa NRT , but has no influence on the swelling rate. Increasing not only the Si-content but also reducing the Ti-content from 0.5 to 0.3% increases not only the incubation dose but also reduces the swelling rate. At 500 C both the increase of the Si-content and the reduction of the Ti-content result in a reduced swelling rate without any effect on the incubation dose. Therefore the best swelling resistance is achieved with a high Si-content and a low Ti-content, resulting in an understabilised condition. The TEM analysis of the microstructure reveals the mechanisms by which the minor elements influence swelling. Increasing the Si-content increases the vacancy mobility and reduces therefore the void nucleation rate. This yields the increased incubation dose. But this mechanism is only working if the silicon is dissolved in the matrix. Swelling starts when the amount of dissolved silicon is reduced under a certain amount. The lots with a high Si

  3. Fundamental Studies of Irradiation-Induced Modifications in Microstructural Evolution and Mechanical Properties of Advanced Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James; Heuser, Brent; Hosemann, Peter; Liu, Xiang

    2018-04-24

    This final technical report summarizes the research performed during October 2014 and December 2017, with a focus on investigating the radiation-induced microstructural and mechanical property modifications in optimized advanced alloys for sodium-cooled fast reactor (SFR) structural applications. To accomplish these objectives, the radiation responses of several different advanced alloys, including austenitic steel Alloy 709 (A709) and 316H, and ferritic/ martensitic Fe–9Cr steels T91 and G92, were investigated using a combination of microstructure characterizations and nanoindentation measurements. Different types of irradiation, including ex situ bulk ion irradiation and in situ transmission electron microscopy (TEM) ion irradiation, were employed in this study. Radiation-induced dislocations, precipitates, and voids were characterized by TEM. Scanning transmission electron microscopy with energy dispersive X-ray spectroscopy (STEM-EDS) and/or atom probe tomography (APT) were used to study radiation-induced segregation and precipitation. Nanoindentation was used for hardness measurements to study irradiation hardening. Austenitic A709 and 316H was bulk-irradiated by 3.5 MeV Fe++ ions to up to 150 peak dpa at 400, 500, and 600°. Compared to neutron-irradiated stainless steel (SS) 316, the Frank loop density of ion-irradiated A709 shows similar dose dependence at 400°, but very different temperature dependence. Due to the noticeable difference in the initial microstructure of A709 and 316H, no systematic comparison on the Frank loops in A709 vs 316H was made. It would be helpful that future ion irradiation study on 316 stainless steel could be conducted to directly compare the temperature dependence of Frank loop density in ion-irradiated 316 SS with that in neutron-irradiated 316 SS. In addition, future neutron irradiation on A709 at 400–600° at relative high dose (>10 dpa) can be carried out to compare with ion-irradiated A709. The radiation

  4. Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

    International Nuclear Information System (INIS)

    Lin Jianbo; Yu Xiaohan; Li Aiguo; He Shangming; Cao Xingzhong; Wang Baoyi; Li Zhuoxin

    2014-01-01

    With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He + ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He + ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He + ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy. (author)

  5. BWRVIP-140NP: BWR Vessel and Internals Project Fracture Toughness and Crack Growth Program on Irradiated Austenitic Stainless Steel

    International Nuclear Information System (INIS)

    Gilman, J.

    2005-01-01

    To prepare for this project, EPRI and BWRVIP conducted a workshop at Ponte Vedra Beach, Florida during February 19-21, 2003 (EPRI report 1007822). Attendees were invited to exchange relevant information on the effects of irradiation on austenitic materials in light water reactors and to produce recommendations for further work. EPRI reviewed the data, recommendations, and conclusions derived from the workshop and developed prioritized test matrices defining new data needs. Proposals were solicited, and selected proposals are the basis for the program described in this report. Results The planned test matrix for fracture toughness testing includes 21 tests on 5 materials

  6. Recovery of electron irradiated V-Ga alloys

    International Nuclear Information System (INIS)

    Leguey, T.; Monge, M.; Pareja, R.; Hodgson, E.R.

    2000-01-01

    The recovery characteristics of electron-irradiated V-Ga alloys with 1.2 and 4.6 at.% Ga have been investigated by positron annihilation spectroscopy (PAS). It is found that vacancies created by electron irradiation become mobile in these alloys at ∼293 K. This temperature is noticeably lower than that in pure V and V-Ti alloys. The vacancies aggregate into microvoids in V-4.6Ga, but do not in V-1.2Ga. The results indicate that vacancies are bound to Ga-interstitial impurity pairs

  7. Three-dimensional modeling for deformation of austenitic NiTi shape memory alloys under high strain rate

    Science.gov (United States)

    Yu, Hao; Young, Marcus L.

    2018-01-01

    A three-dimensional model for phase transformation of shape memory alloys (SMAs) during high strain rate deformation is developed and is then calibrated based on experimental results from an austenitic NiTi SMA. Stress, strain, and martensitic volume fraction distribution during high strain rate deformation are simulated using finite element analysis software ABAQUS/standard. For the first time, this paper presents a theoretical study of the microscopic band structure during high strain rate compressive deformation. The microscopic transformation band is generated by the phase front and leads to minor fluctuations in sample deformation. The strain rate effect on phase transformation is studied using the model. Both the starting stress for transformation and the slope of the stress-strain curve during phase transformation increase with increasing strain rate.

  8. Intergranular corrosion in unserviced austenitic stainless steel pipes made of alloy 904L; Kornzerfall in nicht betriebsbeanspruchten rostfreien austenitischen Rohren aus Alloy 904L

    Energy Technology Data Exchange (ETDEWEB)

    Neidel, Andreas; Cagliyan, Erhan; Fischer, Boromir; Giller, Madeleine; Riesenbeck, Susanne [Siemens AG, Energy Sector, Berlin (Germany). Gasturbinenwerk Berlin

    2017-09-01

    Seamless tubes of the highly corrosion resistant austenitic steel 1.4539, X1NiCrMoCu25-20-5 (Alloy 904L) were observed to exhibit signs of inter-crystalline damage to a depth of several layers of grains and in particular on their internal surface. The material had been stored and had not been put into service. A number of hypotheses had been discussed to explain the predominant cause of the damage. Using optical light and scanning electron microscopy investigation techniques, clear evidence was obtained indicating it to be inter-crystalline corrosion due to the sensitisation of the grain boundaries. The most probable cause of this was determined to be the presence of residual deposits from the rolling process, which due to poor cleaning, had not been completely removed prior to the final solution annealing treatment. This explaining why predominantly the internal surface of the tubes was affected.

  9. Reirradiation in FFTF of swelling-resistant Path A alloys previously irradiated in HFIR

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1985-01-01

    Disks of Path A Prime Candidate Alloys (in several pretreatment conditions) and several heats of cold-worked (CW) type 316 and D9 type austenitic stainless steels have been irradiated in HFIR at 300, 500, and 600 0 C to fluences producing about 10 to 44 dpa and 450 to 3600 at. ppm He. These samples are being reirradiated in the Materials Open Test Assembly (MOTA) in FFTF at 500 and 600 0 C, together (side by side) with previously unirradiated disks of exactly the same materials, to greater than 100 dpa. These samples many of which have either very fine helium cluster or helium bubble distributions after HFIR irradiation, are intended to test the possibility and magnitude of a helium-induced extension of the initial low-swelling transient regime relative to the void swelling behavior normally found during FFTF irradiation. Further, these samples will reveal the microstructural stability or evolution differences that correlate with such helium effects. 17 references, 4 tables

  10. Swift heavy ion irradiation of Cu-Zn-Al and Cu-Al-Ni alloys.

    Science.gov (United States)

    Zelaya, E; Tolley, A; Condo, A M; Schumacher, G

    2009-05-06

    The effects produced by swift heavy ions in the martensitic (18R) and austenitic phase (β) of Cu based shape memory alloys were characterized. Single crystal samples with a surface normal close to [210](18R) and [001](β) were irradiated with 200 MeV of Kr(15+), 230 MeV of Xe(15+), 350 and 600 MeV of Au(26+) and Au(29+). Changes in the microstructure were studied with transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM). It was found that swift heavy ion irradiation induced nanometer sized defects in the 18R martensitic phase. In contrast, a hexagonal close-packed phase formed on the irradiated surface of β phase samples. HRTEM images of the nanometer sized defects observed in the 18R martensitic phase were compared with computer simulated images in order to interpret the origin of the observed contrast. The best agreement was obtained when the defects were assumed to consist of local composition modulations.

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

    Science.gov (United States)

    Chiang, Jasmine Sheree

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

  12. The effect of prior cold-work on the deformation behaviour of neutron irradiated AISI 304 austenitic stainless steel

    Science.gov (United States)

    Karlsen, Wade; Van Dyck, Steven

    2010-11-01

    Cold-work is intentionally employed to increase the yield strength of austenitic stainless steels and also occurs during fabrication processes, but it has also been associated with greater incidence of stress corrosion cracking. This study examined the effect of up to 3.85 dpa neutron irradiation on the deformation behaviour and microstructures of 30% cold-worked AISI 304 material tensile tested at 300 °C. While the deformation behaviour of 0.07 dpa material was similar to non-irradiated material tested at the same temperature, its stress-strain curve was shifted upwards by about 200 MPa. Materials irradiated to over 2 dpa hardened some 400-500 MPa, but showed limited strain hardening capacity, exhibiting precipitous softening with further straining beyond the yield point. The observed behaviour is most likely a consequence of planar deformation products serving as strengtheners to the unirradiated bulk on the one hand, while promoting strain localization on the other, behaviour exacerbated by the subsequent neutron irradiation.

  13. Potential irradiation of Cu alloys and tungsten samples in DONES

    Science.gov (United States)

    Mota, F.; Palermo, I.; Laces, S.; Molla, J.; Ibarra, A.

    2017-12-01

    Tungsten and Cu alloys are currently proposed as reference candidate material for ITER and DEMO first wall and divertor. Tungsten is proposed for its high fusion temperature and CuCrZr alloys for their high thermal conductivity together with good mechanical properties. However its behaviour under the extreme irradiation conditions as expected in ITER or DEMO fusion reactors is still unknown. Due to the determinant role of H and He played in the material behaviour any irradiation experiment must take into account the amount of these gases produced during the irradiation in Fusion reactors with high-energy neutrons. DONES (DEMO oriented neutron source) has been conceived as a simplified IFMIF (International Fusion Material Irradiation Facility) like plant to provide in a reduced time scale and with a reduced budget—both compared to IFMIF—the basic information on materials damage. The objective of DONES-IFMIF in its first stage will be to test structural materials under similar neutron irradiation nuclear fusion conditions as expected in fusion reactors. These tests will be carried out with specimens irradiated in the so called high flux test module (HFTM). The objective of this paper is to assess on the potential use of DONES to irradiate copper (Cu) alloys and tungsten (W) in the HFTM together with reduced activation ferritic martensitic steel like for example EUROFER (9%-Cr-steel). The presence of Cu alloys or W specimens may have an effect in the irradiation parameters of the EUROFER samples placed also in the HFTM and in the samples of the creep fatigue test module (CFTM). McDeLicious code is used for neutron transport calculations. Damage dose rate and H and He production are analysed in the different locations and compared with the irradiation conditions in first wall and divertor in fusion machines.

  14. Formation of an L10 superstructure in austenite upon the α → γ transformation in the invar alloy Fe-32% Ni

    Science.gov (United States)

    Kabanova, I. G.; Sagaradze, V. V.; Kataeva, N. V.

    2011-09-01

    Structure of a metastable austenitic invar alloy Fe-32% Ni preliminarily quenched for martensite and subjected to α → γ transformation using slow heating to various temperatures (430-500°C) with the formation of variously oriented nanocrystalline lamellar austenite, which was subjected to an additional annealing at 280°C (below the calculated temperature of ordering of the γ phase), has been studied electron-microscopically. An electron diffraction analysis revealed the presence of an L10 superstructure in the disperse nickel-enriched nanocrystalline γ phase both after annealing at 280°C and in the unannealed alloy immediately after α → γ transformation upon slow heating to 430°C.

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

    International Nuclear Information System (INIS)

    Barre, Bertrand

    1969-02-01

    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

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

    International Nuclear Information System (INIS)

    Glass, R.S.

    1985-09-01

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

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

    International Nuclear Information System (INIS)

    Kalin, B.A.; Korshunov, S.N.; Chernov, I.I.

    1987-01-01

    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

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

    Science.gov (United States)

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

    2010-07-06

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

  19. Solid state alloying by plasma nitriding and diffusion annealing treatment for austenitic stainless steel

    International Nuclear Information System (INIS)

    Pinedo, C.E.; Vatavuk, J.; Oliveira, S.D. de; Tschiptschin, A.P.

    1999-01-01

    Nitrogen has been added to stainless steels to improve mechanical strength and corrosion resistance. High nitrogen steel production is limited by high gas pressure requirements and low nitrogen solubility in the melt. One way to overcome this limitation is the addition of nitrogen in solid state because of its higher solubility in austenite. However, gas and salt bath nitriding have been done at temperatures around 550 C, where nitrogen solubility in the steel is still very low. High temperature nitriding has been, thus proposed to increase nitrogen contents in the steel but the presence of oxide layers on top of the steel is a barrier to nitrogen intake. In this paper a modified plasma nitriding process is proposed. The first step of this process is a hydrogen plasma sputtering for oxide removal, exposing active steel surface improving nitrogen pickup. This is followed by a nitriding step where high nitrogen contents are introduced in the outermost layer of the steel. Diffusion annealing is then performed in order to allow nitrogen diffusion into the core. AISI 316 austenitic stainless steel was plasma nitrided and diffusion annealed at 1423K, for 6 hours, with 0.2 MPa nitrogen pressure. The nitrided steel presented ∝60 μm outermost compact layer of (Fe,Cr) 3 N and (Fe,Cr) 4 N with 11 wt.% N measured by surface depth profiling chemical analysis - GDS system. During the annealing treatment the nitride layer was dissolved and nitrogen diffused to the core of the sample leaving more even nitrogen distribution into the steel. Using this technique one-millimetre thick sample were obtained having high nitrogen content and uniform distribution through the thickness. (orig.)

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

    International Nuclear Information System (INIS)

    Gessel, G.R.

    1978-06-01

    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 730 0 C 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 850 0 C. 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 650 0 C, the hardness data was found to fit an equation of the form: H = H 0 + b/T; where H is the hardness, T is the temperature, and H 0 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

  1. The irradiation hardening of Ni-Mo-Cr and Ni-W-Cr alloy under Xe26+ ion irradiation

    Science.gov (United States)

    Chen, Huaican; Hai, Yang; Liu, Renduo; Jiang, Li; Ye, Xiang-xi; Li, Jianjian; Xue, Wandong; Wang, Wanxia; Tang, Ming; Yan, Long; Yin, Wen; Zhou, Xingtai

    2018-04-01

    The irradiation hardening of Ni-Mo-Cr and Ni-W-Cr alloy was investigated. 7 MeV Xe26+ ion irradiation was performed at room temperature and 650 °C with peak damage dose from 0.05 to 10 dpa. With the increase of damage dose, the hardness of Ni-Mo-Cr and Ni-W-Cr alloy increases, and reaches saturation at damage dose ≥1 dpa. Moreover, the damage dose dependence of hardness in both alloys can be described by the Makin and Minter's equation, where the effective critical volume of obstacles can be used to represent irradiation hardening resistance of the alloys. Our results also show that Ni-W-Cr alloy has better irradiation hardening resistance than Ni-Mo-Cr alloy. This is ascribed to the fact that the W, instead of Mo in the alloy, can suppress the formation of defects under ion irradiation.

  2. Neutron irradiation damage of a stress relieved TZM alloy

    International Nuclear Information System (INIS)

    Abe, K.; Masuyama, T.; Satou, M.; Hamilton, M.L.

    1992-01-01

    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 10 22 n/cm 2 . 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

  3. Effect of silicon on stability of austenite during isothermal annealing of low-alloy steel with medium carbon content in the transition region between pearlitic and bainitic transformation

    Science.gov (United States)

    Jeníček, Š.; Vorel, I.; Káňa, J.; Ibrahim, K.; Kotěšovec, V.

    2017-02-01

    In a vast majority of steels, a prerequisite to successful heat treatment is the phase transformation of initial austenite to the desired type of microstructure which may consist of ferrite, pearlite, bainite, martensite or their combinations. Diffusion plays an important role in this phase transformation. Together with enthalpy and entropy, two thermodynamic quantities, diffusion represents the decisive mechanism for the formation of the particular phase. The basis of diffusion is the thermally-activated movement of ions of alloying and residual elements. It is generally known that austenite becomes more stable during isothermal treatment in the transitional region between pearlitic and bainitic transformation. This is due to thermodynamic processes which arise from the chemical composition of the steel. The transformation of austenite to pearlite or bainite is generally accompanied by formation of cementite. The latter can be suppressed by adding silicon to the steel because this element does not dissolve in cementite, and therefore prevents its formation. The strength of this effect of silicon depends mainly on the temperature of isothermal treatment. If a steel with a sufficient silicon content is annealed at a temperature, at which silicon cannot migrate by diffusion, cementite cannot form and austenite becomes stable for hours.

  4. 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.

  5. 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)

    Yonezawa, Toshio; Iwamura, Toshihiko; Fujimoto, Koji; Ajiki, Kazuhide

    2000-01-01

    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 M 23 C 6 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 M 23 C 6 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Brimbal, Daniel, E-mail: Daniel.brimbal@areva.com [AREVA NP, Tour AREVA, 1 Place Jean Millier, 92084 Paris La Défense (France); Fournier, Lionel [AREVA NP, Tour AREVA, 1 Place Jean Millier, 92084 Paris La Défense (France); Barbu, Alain [Alain Barbu Consultant, 6 Avenue Pasteur Martin Luther King, 78230 Le Pecq (France)

    2016-01-15

    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. - Highlights: • Irradiation of steels with helium is studied through a new cluster dynamics model. • There is only a small effect of helium on cavity distributions in PWR conditions. • An increase in helium production causes an increase in cavity density over 500 °C. • The role of helium is to stabilize cavities via reduced emission of vacancies.

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

    International Nuclear Information System (INIS)

    Zelaya, Maria Eugenia

    2006-01-01

    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

  8. Effect of neutron irradiation on Mo-Si amorphous alloys

    International Nuclear Information System (INIS)

    Ito, Fumitake; Hasegawa, Masayuki; Suzuki, Kenji; Honda, Toshihisa; Fukunaga, Toshiharu.

    1982-01-01

    The irradiation effects on Mo-Si amorphous alloys were investigated by means of X-ray diffraction and positron annihilation, and their electric resistance at low temperature was measured to examine the superconductivity of the alloys. The specimens of Mo 68 Si 32 and Mo 45 Si 55 were irradiated with the neutron fluence (E > 1 MeV) of about 9 x 10 18 n/cm 2 without temperature control in the Japanese Material Testing Reactor (JMTR). For these irradiated specimens, the X-ray diffraction experiment was performed to examine the irradiation effects on the radial distribution function, and the angular correlation curves for the positron annihilation were also measured. Both experiments showed that there was almost no irradiation effect. However, the width of the superconductive transition measured in Mo 68 Si 32 became extremely narrow due to neutron irradiation, and the transition temperature rose from 6.89 K to 7.03 K. On the other hand, in Mo 45 Si 55 , the width showed a tendency to become somewhat narrow, but the transition temperature shifted to the lower side. (Asami, T.)

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

    International Nuclear Information System (INIS)

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

    1990-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  11. Topological model of austenite-martensite interfaces in Cu-Al-Ni alloy

    Czech Academy of Sciences Publication Activity Database

    Ostapovets, Andriy; Zárubová, Niva; Paidar, Václav

    2012-01-01

    Roč. 122, č. 3 (2012), s. 493-496 ISSN 0587-4246. [International Symposium on Physics of Materials, ISPMA /12./. Praha, 04.09.2011-08.09.2011] R&D Projects: GA AV ČR IAA100100920 Institutional research plan: CEZ:AV0Z10100520 Keywords : CuAlNi * alloy * experimental data * in-situ * topological models * transmission electron microscopy Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.531, year: 2012

  12. Magnetic properties near the ferromagnetic-paramagnetic transformation in the austenite phase of Ni43Mn44X2Sn11 (X = Fe and Co) Heusler alloys

    Science.gov (United States)

    Nan, W. Z.; Thanh, T. D.; You, T. S.; Piao, H. G.; Yu, S. C.

    2018-03-01

    In this work, we present a detail study on the magnetic properties in the austenitic phase (A phase) Ni43Mn44X2Sn11 alloy with X = Fe and Co, which were prepared by an arc-melting method in an argon atmosphere. The M(T) curves of two samples exhibits a single magnetic phase transition at the Curie temperature of the ferromagnetic (FM) austenitic phase with TCA = 298 K and 334k for (X = Fe and Co) respectively. Based on the Landau theory and M(H) data measured at different temperatures, we found that the FM-PM phase transitions around TCA in both samples were the second-order phase transition. Under an applied field change of 30 kOe, around TCA , the magnetic entropy changes were found to be 0.66 J Kg-1 K-1 and 1.62 J Kg-1 K-1 for (X = Fe and Co) respectively.

  13. Influence of neutron irradiation at 550C on the properties of austenitic stainless steels

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.

    1981-01-01

    Types 316 and 316 + 0.23 wt % Ti stainless steels and 16-8-2 weldment were irradiated in HFIR at 55 0 C to fluences up to 1.35 x 10 26 neutrons/m 2 ( 0 C strength properties, with the weldments the weakest of the materials. The ductility of all materials was reduced by the irradiation, the uniform elongation to only 0.4% in the cold-worked material. Tests at temperatures above the irradiation temperature showed an approach to unirradiated properties as the temperature was increased from 200 to 600 0 C. Helium embrittlement at 700 0 C severely reduced elongation

  14. Design of a single variable helium effects experiment for irradiation in FFTF [Fast Flux Test Facility] using alloys enriched in nickel 59

    International Nuclear Information System (INIS)

    Simons, R.L.; Brager, H.R.; Matsumoto, W.Y.

    1986-03-01

    Nickel enriched in nickel 59 was extracted from the fragments of a fracture toughness specimen of Inconel 600 irradiated in the Engineering Test Reactor (ETR). The nickel contained 2.0% nickel 59. Three heats of austenitic steel doped with nickel-59 were prepared and inserted in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF). The experiment was single variable in helium effects because chemically identical alloys without nickel-59 were being irradiated side by side with the doped material. The alloys doped with nickel 59 produced 10 to 100 times more helium than the control alloys. The materials included ternary and quaternary alloys in the form of transmission electron microscope (TEM) discs and miniature tensile specimens. The helium to dpa ratio was in the range 5 to 35 and was nearly constant throughout the irradiation. The exposures ranged from 0.25 to 50 displacements per atom (dpa) over the duration of the experiment. The irradiation temperatures covered the range of 360 to 600 0 C

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhijie [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-03-31

    The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.

  17. Positron lifetime measurements on electron irradiated amorphous alloys

    International Nuclear Information System (INIS)

    Moser, P.; Hautojaervi, P.; Chamberod, A.; Yli-Kauppila, J.; Van Zurk, R.

    1981-08-01

    Great advance in understanding the nature of point defects in crystalline metals has been achieved by employing positron annihilation technique. Positrons detect vacancy-type defects and the lifetime value of trapped positrons gives information on the size of submicroscopic vacancy aglomerates and microvoids. In this paper it is shown that low-temperature electron irradiations can result in a considerable increase in the positron lifetimes in various amorphous alloys because of the formation of vacancy-like defects which, in addition of the pre-existing holes, are able to trap positrons. Studied amorphous alloys were Fe 80 B 20 , Pd 80 Si 20 , Cu 50 Ti 50 , and Fe 40 Ni 40 P 14 B 6 . Electron irradiations were performed with 3 MeV electrons at 20 K to doses around 10 19 e - /cm 2 . After annealing positron lifetime spectra were measured at 77 K

  18. Swelling in simple ferritic alloys irradiated to high fluence

    International Nuclear Information System (INIS)

    Gelles, D.S.; Meinecke, R.L.

    1984-01-01

    A series of Fe-Cr-C-Mo simple alloys has been measured for density change as a function of irradiation in EBR-II over the temperature range 400 to 650 0 C to fluences as high as 2.13 x 10 23 n/cm 2 (E > 0.1 MeV) or 105 dpa. The highest swelling was found in a Fe-12Cr binary alloy, 4.72 percent, after 1.87 x 10 23 n/cm 2 or 95 dpa at 425 0 C, which corresponds to a swelling rate of 0.06%/dpa. This peak swelling rate value can be used to define swelling predictions for commercial ferritic alloys to 40 MWy/m 2

  19. Irradiation testing of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-01-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 'microplates'. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U10Mo-0.05Sn, U2Mo, or U 3 Si 2 . These experiments will be discharged at peak fuel burnups of approximately 40 and 80 at.% U 235 . Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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 ∼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

  2. Neutron small angle scattering of irradiated aluminium-silicon alloys

    International Nuclear Information System (INIS)

    Kostorz, G.

    1976-01-01

    Technically pure aluminium and aluminium-silicon alloys (0.43, 0.83 and 1.2% Si, also containing 0.11 to 0.14 at. % Fe) were investigated by slow neutron small angle scattering after irradiation with fast neutrons at low temperatures. Different irradiation levels, ageing at room temperature and at 60/70 0 C had no measurable effect upon small angle scattering cross-sections. From the experimental precision upper limit for the amount of Si involved in clustering after irradiation can be given. The observed small angle scattering shows a strong dependence on scattering angles and is attributed to large precipitates of Al 12 Fe 3 Si. A surface layer on the as-received samples is identified as another source of low-intensity small angle scattering. (orig.) [de

  3. Factors affecting the grain growth of austenite in low alloy steel

    International Nuclear Information System (INIS)

    Parker, J.D.; Storer, S.M.

    1995-01-01

    The performance of steels is linked to the metallurgical transformations which occur during manufacture. Clearly then the optimization of a fabrication procedure must be based on fundamental relationships linking specific thermal treatments with transformation behaviour. Optimized manufacture of thick-section, multipass welds is therefore particularly complex since the thermal cycles associated with fusion welding result in the formation of heterogeneous microstructures. Moreover, these transformations will take place under rapid heating and cooling conditions so that standard data based on equilibrium behaviour may not be directly relevant. The present study is part of an integrated research programme aimed at establishing the basic microstructural relationships required to optimize the manufacture and performance of weldments. Work to date demonstrates that utilization of a computer controlled Gleeble simulation system allows a wider range of heating and cooling rates to be applied than is possible with traditional heat treatment techniques. Additional advantages of this system include precise control of time at peak temperature and uniform temperatures within a defined work zone. Results presented for a CrMoV creep resistant low alloy steel indicate that grain growth behaviour in the range 955-1390 C can be related to the time at peak temperature. The effect of this transformation behaviour on weldment behaviour is discussed. (orig.)

  4. Nucleation mechanisms of dynamic recrystallization in austenitic steel alloy 800H

    Energy Technology Data Exchange (ETDEWEB)

    Bruenger, E.; Wang, X.; Gottstein, G. [RWTH Aachen (Germany). Inst. fuer Metallkunde und Metallphysik

    1998-05-12

    Many metals and alloys with low and intermediate stacking fault energy undergo dynamic recrystallization (DRX). Due to the growing importance of hot deformation in metal forming there is an increasing interest in the understanding and modeling of microstructure evolution during DRX and its effect on flow behavior. However, despite extensive research in this field and numerous data on a variety of materials the physical understanding of DRX still remains very qualitative. Especially the nucleation of DRX lacks a detailed physical understanding and experimental evidence, due to the difficulties of investigating the micromechanisms of dynamic processes during high temperature deformation. The improved techniques of single grain orientation measurements by using EBSD (electron backscatter diffraction) in the SEM allow to measure the local orientation arrangement and thus identify the orientations of individual nuclei. The current report focuses on the examination of the substructure evolution during dynamic recrystallization with particular attention to the role of continuous subgrain rotation or instabilities of the subgrain structure near the grain boundary with regard to nucleation during DRX.

  5. 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.)

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

    International Nuclear Information System (INIS)

    Maziasz, P.J.; Braski, D.N.

    1984-01-01

    Swelling evaluation of PCA variants and 20%-cold-worked (N-Lot) type 316 stainless steel (CW 316) at 300 to 600 0 C was extended to 44 dpa. Swelling was negligible in all the steels at 300 0 C after approx. 44 dpa. At 500 to 600 0 C 25%-cold-worked PCA showed better void swelling resistance than type 316 at approx. 44 dpa. There was less swelling variation among alloys at 400 0 C, but again 25%-cold-worked PCA was the best

  7. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

  8. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    International Nuclear Information System (INIS)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

  9. Grain-boundary microchemistry and intergranular cracking of irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

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

    1993-01-01

    Constant-extension-rate tensile tests and grain-boundary analysis by Auger electron spectroscopy were conducted on high and commercial-purity (HP and CP) Type 304 stainless steel (SS) specimens from irradiated boiling-water reactor (BWR) components to identify the mechanisms of irradiation-assisted stress corrosion cracking (IASCC). Contrary to previous beliefs, susceptibility to intergranular fracture could not be correlated with radiation-induced segregation of impurities such as Si, P, C, or S, but a correlation was obtained with grain-boundary Cr concentration, indicating a role for Cr depletion. Detailed analysis of grain-boundary chemistry was conducted on BWR neutron absorber tubes that were fabricated from two similar heats of HP Type 304 SS of virtually identical bulk chemical composition but exhibiting a significant difference in susceptibility to IASCC after irradiation to ∼2 x 10 21 n/cm 2 (E > 1 MeV). Grain-boundary concentrations of Cr Ni, Si, P, S, and C of the cracking-resistant and -susceptible HP heats were virtually identical. However, grain boundaries of the cracking-resistant material contained less N and more B and Li than those of the cracking-susceptible material. This observation indicates that, besides the deleterious effect of grain-boundary Cr depletion, a synergism between grain-boundary segregation of N and B and transmutation to H and Li plays an important role in IASCC

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

    International Nuclear Information System (INIS)

    Audouard, A.

    1982-09-01

    Three Fe-B amorphous alloys (Fe 80 B 20 , Fe 27 Mo 2 B 20 and Fe 75 B 25 ) and the crystallized Fe 3 B alloy have been irradiated at the temperature of liquid hydrogen. Electron irradiation and irradiation by 10 B 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 235 U fission fragments induces some important structural modifications in the amorphous alloys [fr

  11. Recovery characteristics of neutron-irradiated V-Ti alloys

    International Nuclear Information System (INIS)

    Leguey, T.; Pareja, R.

    2000-01-01

    The recovery characteristics of neutron-irradiated pure V and V-Ti alloys with 1.0 and 4.5 at.% Ti have been investigated by positron annihilation spectroscopy. Microvoid formation during irradiation at 320 K is produced in pure V and V-1Ti but not in V-4.5Ti. The results are consistent with a model of swelling inhibition induced by vacancy trapping by solute Ti during irradiation. The temperature dependencies of the parameter S in the range 8-300 K indicate a large dislocation bias for vacancies and solute Ti. This dislocation bias prevents the microvoid nucleation in V-4.5Ti, and the microvoid growth in V-1Ti, when vacancies become mobile during post-irradiation annealing treatments. A characteristic increase of the positron lifetime is found during recovery induced by isochronal annealing. It is attributed to a vacancy accumulation into the lattice of Ti oxides precipitated during cooling down, or at their matrix/precipitate interfaces. These precipitates could be produced by the decomposition of metastable phases of Ti oxides formed during post-irradiation annealing above 1000 K

  12. A method for the calculation of retained austenite evolution during heat-treatment of low-alloy TRIP-assisted steels

    Energy Technology Data Exchange (ETDEWEB)

    Katsamas, A.I. [Dept. of Mechanical and Industrial Engineering, Univ. of Thessaly, Volos (Greece)

    2006-03-15

    Despite the critical effect of heat-treatment, and in particular of the isothermal bainitic treatment stage, on the amount and stability of retained austenite in the microstructure of low-alloy TRIP-assisted steels, determination of optimum heat-treatment conditions is still largely empirical and experiment-dependent. This work proposes a method by which it is possible to calculate the vol. fraction of retained austenite in the microstructure as a function of intercritical annealing temperature and isothermal bainitic treatment temperature and holding time. The method assumes diffusionless lengthening of bainitic ferrite ({alpha}{sub B}) plates in austenite ({gamma}), and subsequent thickness-wise C rejection from the {alpha}{sub B} plates to the adjacent {gamma} layers. The relative thickness of {alpha}{sub B} plates and adjacent {gamma} layers is determined by the T{sub o} line of the transforming system at any given bainitic transformation temperature. The C-concentration profiles in {gamma} are calculated with respect to a local time-scale, referring to any random section of any random {alpha}{sub B} plate. Determination of the variation of C-concentration profiles with local time in {gamma}, together with the use of a simple austenite-retention criterion, allows the calculation of vol. fraction retained austenite ({gamma}{sub R}) as a function of transformation temperature and local time. Transition from local (calculation) time to actual heat treatment time is performed by introducing a time-scale factor, which depends on transformation temperature and initial C-content of {alpha}{sub B}. The calculated behaviour of vol. fraction {gamma}{sub R} vs. bainitic holding time conforms to the well established, experimentally observed one: vol. fraction {gamma}{sub R} initially increases with holding time, reaches a maximum and decreases at longer holding times. According to calculated results, the decrease is attributed to the gradual homogenization of C inside the

  13. Imposed potential measurement to evaluate the pitting corrosion resistance and the galvanic behaviour of a highly alloyed austenitic stainless steel and its weldment in a LiBr solution at temperatures up to 150ºC

    OpenAIRE

    Blasco Tamarit, María Encarnación; García García, Dionisio Miguel; García Antón, José

    2011-01-01

    Pitting corrosion resistance and galvanic behaviour of Alloy 31, a highly alloyed austenitic stainless steel (UNS N08031), and its weldment were studied in a heavy brine LiBr solution 1080 g/l at different temperatures (75–150 °C) using electrochemical techniques. The Mixed Potential Theory was used to evaluate the galvanic corrosion between the base and welded metals. Cyclic potentiodynamic curves indicate that high temperatures make passivation and repassivation of pits difficult, because t...

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

    International Nuclear Information System (INIS)

    Ershova, O.V.; Shcherbakov, E.N.; Evseev, M.V.; Shihkalev, V.S.; Kozlov, A.V.; Garner, F.

    2007-01-01

    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 ∼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

  15. Neutron irradiation effects on magnetic properties of some Heusler alloys

    International Nuclear Information System (INIS)

    Onodera, Hideya; Shinohara, Takeshi; Yamamoto, Hisao; Watanabe, Hiroshi

    1975-01-01

    The neutron irradiation effects were studied with measurements of temperature dependence of magnetization in ordered and disordered Heusler alloys. The irradiation was carried out in JMTR with a total flux of fast neutrons of 10 20 nvt. Fully ordered Cu 2 MnIn, partially ordered Cu 2 MnAl and completely disordered Cu 2 MnSn were prepared with various temperature treatments. The magnetization-temperature curves of each specimen were measured before and after irradiation. In the irradiated Cu 2 MnIn, the disordering by the irradiation gave rise to a decrease of magnetization, and the temperature dependence of magnetization showed that the disordered region contained various regions with different degrees of disorder. For the distribution of the disordered region, the calculation based on the theory of temperature spike by Seitz and Koekler gave a feasible result that a disordered region comprised a central core with a radius of 5.4 A which was completely disordered and a periphery of 3.3 A thickness which was partially disordered. From the magnetization-temperature curves of Cu 2 MnAl, it was considered that the disordered regions induced by the irradiation had different properties from those induced by the heat treatment. The former were the localized and comprised regions corresponding to various degrees of disorder, while the latter spread spatially in a wide range with a certain degree of disorder. The ordering by enhanced diffusion occurred simultaneously to an extent comparable to the disordering, and so it played an important role in the magnetization in the partially disordered Cu 2 MnAl. In the disordered Cu 2 MnSn, however, the ordering effect was very small. It is supposed to be difficult for the A2 structure to transform into the L2 1 structure by the enhanced diffusion. (auth.)

  16. Irradiation response of rapidly solidified Path A type prime candidate alloys

    International Nuclear Information System (INIS)

    Imeson, E.; Tong, C.; Lee, M.; Vander Sande, J.B.; Harling, O.K.

    1981-01-01

    The objective of this study is to present a first assessment of the microstructural response to neutron irradiation shown by Path A alloys prepared by rapid solidification processing. To more fully demonstrate the potential of the method, alloys with increased titanium and carbon content have been used in addition to the Path A prime candidate alloy

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

    International Nuclear Information System (INIS)

    Audouard, Alain.

    1983-01-01

    Three iron-boron amorphous alloys and the crystalline Fe 3 B alloy have been irradiated at liquid hydrogen temperature. 2,4 MeV electron irradiation induces the creation of point defects in the amorphous alloys as well as in the crystalline Fe 3 B alloy. These point defects can be assimilated to iron ''Frenkel pairs''. They have been characterized by determining their intrinsic electrical resistivity and their formation volume. The displacement threshold energy of iron atoms has also been determined. 10 B fission fragments induce, in these amorphous alloys, displacement cascades which lead to stable vacancy rich zones. This irradiation also leads to a structural disorder in relation with the presence of defects. 235 U fission fragments irradiation modifies drastically the structure of the amorphous alloys. The results have been interpreted on the basis of the coexistence of two opposite processes which induce local disorder and crystallisation respectively [fr

  18. Post-irradiation mechanical properties of an AlMgSi alloy

    Energy Technology Data Exchange (ETDEWEB)

    Ismail, Z.H. [Atomic Energy Commission, Cairo (Egypt). Dept. of Metall.; Birt, B. [Atomic Energy Commission, Cairo (Egypt). Dept. of Metall.

    1995-03-01

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

  19. Austenite-to-ferrite transformation in low alloy steels during thermornechanically controlled process studied by in situ neutron diffraction

    Czech Academy of Sciences Publication Activity Database

    Xu, PG.; Tomota, Y.; Lukáš, Petr; Muránsky, Ondrej; Adachi, Y.

    2006-01-01

    Roč. 435, č. 5 (2006), s. 46-53 ISSN 0921-5093 R&D Projects: GA AV ČR IAA1048107 Institutional research plan: CEZ:AV0Z10480505 Keywords : austenite-to-ferrite tranformation * neutron diffraction * thermomechanically controlled prosess Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.490, year: 2006

  20. Changes in Mechanical Properties of SA508 Gr.4N Model Alloys with Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The mechanical properties and irradiation embrittlement behavior of SA508 Gr.4N low alloy steel were evaluated. The yield strength and tensile strength were increased with an increase in fluence level, but there is no drastic increase in strength. A significant increase in the transition temperature shifts from the Charpy impact test and fracture toughness test was not observed in SA508 Gr.4N model alloy. The overall irradiation embrittlement behavior of SA508 Gr.4N low alloy steel is almost similar to that of SA508 Gr.3 low alloy steel, and an increase in Ni content by a few percentage points in SA508 Gr.4N model alloys compared to SA508 Gr.3 low alloy steel did not result in an increased embrittlement of these alloys. The yield strength was increased with an increase in the neutron fluence level, and the amount of strength increase was comparable to commercial SA508 Gr.3 low alloy steel.

  1. A review of the effect of neutron irradiation on the deformation behaviour of copper and copper alloys

    International Nuclear Information System (INIS)

    Higgy, H.R.

    1976-08-01

    The basic mechanisms of irradiation hardening are described. The effects of neutron dose, alloying and pre-irradiation deformation on the deformation behaviour of neutron-irradiatied copper and its alloys are considered. The discrepancy in the reported data is discussed. Substitutional and interstitial additions are found to influence the rate of irradiation hardening, while pre-irradiation deformation has no influence. The deformation behaviour of copper is found to alter as a result of irradiation and alloying. (author)

  2. HIGH TEMPERATURE BRAZING ALLOY FOR JOINT Fe-Cr-Al MATERIALS AND AUSTENITIC AND FERRITIC STAINLESS STEELS

    Science.gov (United States)

    Cost, R.C.

    1958-07-15

    A new high temperature brazing alloy is described that is particularly suitable for brazing iron-chromiumaluminum alloys. It consists of approximately 20% Cr, 6% Al, 10% Si, and from 1.5 to 5% phosphorus, the balance being iron.

  3. Influence of value of the criterion of blunting on the selection of the brand of hard alloy and optimum cutting speed when turning austenitic steel

    Directory of Open Access Journals (Sweden)

    Lipatov Andrew A.

    2017-01-01

    Full Text Available A mechanisms of wear carbide tool when turning of austenitic steel 18-8 by cutters from firm hard-alloys of various groups (WC-Co, TiC-WC-Co, TiC-TaC-WC-Co were studied. During wear resistant tests it is established, that the prevalence of one of the two mechanisms of wear (of adhesion-fatigue or dif-fusional depends on the brand of hard alloy. It is shown, that in machining by titanium-containing carbide tool the intensity of the growth of the wear platform on the back surface of the tool as wear changes. This is due to the smooth transition from the predominance of adhesion-fatigue wear to prevalence of diffusional wear. Therefore, the intensity of wear should be considered as current, depending on the value wear platform, and value of the criterion of blunting is influence on the selection of the brand of hard-alloy and optimum cutting speed.

  4. PDS 1-5. Divertor heat sink materials pre- and post-neutron irradiation. Tensile and fatigue tests of brazed joints of molybdenum alloys and 316L stainless steel

    International Nuclear Information System (INIS)

    Lind, Anders.

    1994-01-01

    Tensile specimens from brazed joints of molybdenum alloys (TZM or Mo-5%Re) and Type 316L austenitic stainless steel tubes have been tested at ambient temperature and 127 degrees C before and after neutron irradiation at about 40 degrees C to approximately 0.2 dpa. The unirradiated specimens showed generally ductile behaviour, but the irradiated specimens were notch sensitive and failed in a brittle manner with zero elongation; in all cases the fracture occurred in the molybdenum alloy. The brittle behaviour is consistent with previously published data and results from the increase in strength (radiation hardening) and the associated increase in the ductile-brittle transition temperature (radiation embrittlement) induced in the body-centered-cubic (BCC) molybdenum alloys by irradiation to relatively low displacement doses. The same type of irradiated specimens were also used in fatigue tests. However, the results from the fatigue tests are too limited and complementary studies are needed. During exposure to water locally up to 25% of the wall thickness of the Mo-alloys has corroded away. These observations cast serious doubts on the viability of the molybdenum alloys for divertor applications in fusion systems. 8 refs, 29 figs

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. 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).

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

    International Nuclear Information System (INIS)

    Ansari, I.

    1985-04-01

    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 300 0 C under different tensile stresses. After irradiation transmission electron microscopic (TEM) investigations were made to observe the precipitate structure under irradiation for different experimental parameters. Moreover, the irradiation-induced changes in precipitate structure and changes of Al-concentrations in the matrix in Ni-13.1 at% Al alloys were studied by electrical resistivity measurements during irradiation. For comparison, the electrical resistivity of unirradiated specimens was also measured after thermal aging for different times. For correlation, TEM analysis was performed on irradiated and unirradiated aged specimens. Tensile tests on annealed and aged Ni-Al alloys were also done at various temperatures. (orig./RK)

  8. Positron annihilation in hydrogenated and electron-irradiated titanium alloys

    International Nuclear Information System (INIS)

    Mukashev, K.M.; Zaikin, Yu.A.

    2002-01-01

    that material properties were not completely recovered, probably due to residual point defects of radiation origin. It is obvious that the temperature of 600 deg. C was not sufficient for hydrogen extraction from titanium. These results are confirmed by data of previous studies where niobium and nickel hydrogenated after electron irradiation were studied. These data show that the shift in the recovery start exceeded 130 deg. C. Thus, interaction of vacancy-type defects with previously introduced hydrogen surrounding causes alterations in the efficient size of positron localization centers and shifts the first recovery stage to the region of higher temperature values. Generally, the results of this study demonstrate a significant role of hydrogen in alterations of the electron structure of damaged materials. They show the increasing hydrogen interaction with materials in presence of structural imperfections of deformational and radiation origin. Both hydrogen behavior in irradiated titanium alloys and the observed alterations in positron annihilation characteristics cannot be described in frames of a simple model based on the concept of proton interaction with a vacancy. Variety of radiation defects, such as complexes of point defects, dislocations, cracks, etc., should be taken into account. Application of positron annihilation methods provides important information on hydrogen interaction with lattice imperfections that can be a useful approach to the problem of hydrogen embrittlement of structural materials in the fields of ionizing radiation

  9. Neutron irradiation effects on the mechanical properties of thorium and thorium--carbon alloy

    International Nuclear Information System (INIS)

    Wang, S.C.P.

    1978-04-01

    The effects of neutron exposure to 3.0 x 10 18 neutrons/cm 2 on the mechanical properties of thorium and thorium-carbon alloy are described. Tensile measurements were done at six different test temperatures from 4 0 K to 503 0 K and at two strain rates. Thorium and thorium-carbon alloy are shown to display typical radiation hardening like other face-centered cubic metals. The yield drop phenomenon of the thorium-carbon alloy is unchanged after irradiation. The variation of shear stress and effective shear stress with test temperature was fitted to Seeger's and Fleischer's equations for irradiated and unirradiated thorium and thorium-carbon alloy. Neutron irradiation apparently contributes an athermal component to the yield strength. However, some thermal component is detected in the low temperature range. Strain-rate parameter is increased and activation volume is decreased slightly for both kinds of metal after irradiation

  10. Gas phase hydrogen permeation through ferritic iron, austenitic stainless steel and neutron irradiated austenitic stainless steel from near 3000K to 8730K

    International Nuclear Information System (INIS)

    Quick, N.R.

    1976-01-01

    Hydrogen permeation through iron was studied over the temperature range 300 to 873 0 K by an ultra high vacuum, monopole gas analyzer technique. Hydrogen gas input pressures were varied from 0.0043 to 0.62 atm and membrane thicknesses from 0.0165 to 0.243 cm. Volume diffusion control of the permeation process was demonstrated by the pressure and membrane thickness dependence of the steady state flux. The permeation coefficient, with an activation enthalpy found to be 8.1 +-.4 kcal/mole, was independent of both gas pressure and membrane thickness. At temperatures below approximately 600 0 K, the effective diffusivity increased with both increasing hydrogen gas pressure and increasing membrane thickness. The transition temperature from classical to anomalous behavior decreases with increasing thickness. Apparent activation enthalpies for diffusion were found to range from 1.6 to 8.2 kcal/mole with the lower values associated with thicker membranes. The permeation coefficient activation enthalpy was found to be 13.1 +- .4 kcal/mole while that for diffusivity was found to be 11.2 +- .45 kcal/mole. However, samples neutron irradiated at a fluence of 10 17 n/cm 2 showed anomalous effects in that both effective diffusivity and permeation were reduced in value

  11. The microstructure and hardness changes of neutron irradiated weld joint of vanadium alloy

    Science.gov (United States)

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

    2011-10-01

    Effects of neutron irradiation on YAG laser welded V-4Cr-4Ti alloy were irradiated in High Flux Isotope Reactor (HFIR). The samples were irradiated in Li environment at 723 K and 873 K up to the dose of 3.7 dpa. After the irradiation, the microstructure and Vickers hardness of the welded samples were compared of the base metal, which were simultaneously irradiated at the same irradiation cycle. At 723 K, very high density of dislocations was formed. But prominent Ti(CON) formation, which was commonly observed in He gas and vacuum environment condition, was not detected

  12. Correlative Microscopy of Alpha Prime Precipitation in Neutron-Irradiated Fe-Cr-Al Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States)

    2016-12-01

    Fe-Cr-Al alloys are currently being considered for accident tolerant light water reactor fuel cladding applications due to their superior high temperature oxidation and corrosion resistance compared to Zr-based alloys. This work represents the current state-of-the-art on both techniques for analysis of α' precipitate microstructures and the processes and mechanisms governing its formation in neutron-irradiated Fe-Cr-Al alloys.

  13. The Influence of Austenite Grain Size on the Mechanical Properties of Low-Alloy Steel with Boron

    Directory of Open Access Journals (Sweden)

    Beata Białobrzeska

    2017-01-01

    Full Text Available This study forms part of the current research on modern steel groups with higher resistance to abrasive wear. In order to reduce the intensity of wear processes, and also to minimize their impact, the immediate priority seems to be a search for a correlation between the chemical composition and structure of these materials and their properties. In this paper, the correlation between prior austenite grain size, martensite packets and the mechanical properties were researched. The growth of austenite grains is an important factor in the analysis of the microstructure, as the grain size has an effect on the kinetics of phase transformation. The microstructure, however, is closely related to the mechanical properties of the material such as yield strength, tensile strength, elongation and impact strength, as well as morphology of occurred fracture. During the study, the mechanical properties were tested and a tendency to brittle fracture was analysed. The studies show big differences of the analysed parameters depending on the applied heat treatment, which should provide guidance to users to specific applications of this type of steel.

  14. 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)

    Bellanger, G.

    1994-01-01

    The redox potential of 3 H 2 O, 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

  15. Microstructural development of tungsten and tungsten-rhenium alloys due to neutron irradiation in HFIR

    Science.gov (United States)

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

    2014-12-01

    The microstructural development of pure tungsten (W) and tungsten-rhenium (Re) alloys due to neutron irradiation in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, TN, USA, was investigated in this work. The irradiation conditions were ∼1 displacements per atom (dpa) at 500 and 800 °C. After the neutron irradiation, microstructural observations were performed using a transmission electron microscope (TEM). Large amounts of precipitates identified as sigma- and chi-phases were observed in not only the W-Re alloys but also in the pure W after the neutron irradiation. The precipitates observed in the pure W were coarse and larger than those in the W-Re alloys. This was considered to be caused by the transmutation products of W and Re, namely, Re and osmium (Os), respectively, under irradiation in the HFIR with a higher contents of thermal neutron flux.

  16. Solute segregation and void formation in ion-irradiated vanadium-base alloys

    International Nuclear Information System (INIS)

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

    1985-01-01

    The radiation-induced segregation of solute atoms in the V-15Cr-5Ti alloys was determined after either single- dual-, or helium implantation followed by single-ion irradiation at 725 0 C to radiation damage levels ranging from 103 to 169 dpa. Also, the effect of irradiation temperature (600-750 0 C) on the microstructure in the V-15Cr-5Ti alloy was determined after single-ion irradiation to 200 and 300 dpa. The solute segregation results for the single- and dual-ion irradiated alloy showed that the simultaneous production of irradiation damage and deposition of helium resulted in enhanced depletion of Cr solute and enrichment of Ti, C and S solute in the near-surface layers of irradiated specimens. The observations of the irradiation-damaged microstructures in V-15Cr-5Ti specimens showed an absence of voids for irradiations of the alloy at 600-750 0 C to 200 dpa and at 725 0 C to 300 dpa. The principle effect on the microstructure of these irradiations was to induce the formation of a high density of disc-like precipitates in the vicinity of grain boundaries and intrinsic precipitates and on the dislocation structure. 8 references, 4 figures

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

    International Nuclear Information System (INIS)

    Hoshiya, Taiji; Takada, Fumiki; Omi, Masao; Goto, Ichiro; Ando, Hiroei

    1992-01-01

    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 10 25 m -2 and subsequently annealed above 523 K, were examined by electrical resistance measurements and tensile tests. When irradiation was performed at 323 K, M s 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 M s 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 M s 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)

  18. A comparative study of the in vitro corrosion behavior and cytotoxicity of a superferritic stainless steel, a Ti-13Nb-13Zr alloy, and an austenitic stainless steel in Hank's solution.

    Science.gov (United States)

    Assis, S L; Rogero, S O; Antunes, R A; Padilha, A F; Costa, I

    2005-04-01

    In this study, the in vitro corrosion resistance of a superferritic stainless steel in naturally aerated Hank's solution at 37 degrees C has been determined to evaluate the steel for use as a biomaterial. The potentiodynamic polarization method and electrochemical impedance spectroscopy (EIS) were used to determine the corrosion resistance. The polarization results showed very low current densities at the corrosion potential and electrochemical behavior typical of passive metals. At potentials above 0.75 V (SCE), and up to that of the oxygen evolution reaction, the superferritic steel exhibited transpassive behavior followed by secondary passivation. The superferritic stainless steel exhibited high pitting resistance in Hank's solution. This steel did not reveal pits even after polarization to 3000 mV (SCE). The EIS results indicated high impedance values at low frequencies, supporting the results obtained from the polarization measurements. The results obtained for the superferritic steel have been compared with those of the Ti-13Nb-13Zr alloy and an austenitic stainless steel, as Ti alloys are well known for their high corrosion resistance and biocompatibility, and the austenitic stainless steel is widely used as an implant material. The cytotoxicity tests indicated that the superferritic steel, the austenitic steel, and the Ti-13Nb-13Zr alloy were not toxic. Based on corrosion resistance and cytotoxicity results, the superferritic stainless steel can be considered as a potential biomaterial. (c) 2005 Wiley Periodicals, Inc.

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

    International Nuclear Information System (INIS)

    Yamada, Hirokazu; Kawamura, Hiroshi

    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), Al 2 O 3 -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 10 24 n/m 2 (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)

  20. 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)

  1. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C

    DEFF Research Database (Denmark)

    Singh, B.N.; Stubbins, J.F.; Toft, P.

    2000-01-01

    The fatique behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risø witha flux of approx2.5 x 1017 n/m2s (E> 1 Me...... microscope. The present investigations demonstrated that the fatigue life decreases with increasingtemperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures...

  2. Fatigue performance of copper and copper alloys before and after irradiation with fission neutrons

    International Nuclear Information System (INIS)

    Singh, B.N.; Toft, P.; Stubbins, J.F.

    1997-05-01

    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 10 17 n/m 2 s (E > 1 MeV) to fluence levels of 1.5 - 2.5 x 10 24 n/m 2 s (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

  3. Temperature profiles of low-temperature alloy irradiated by pulsed ion beams

    International Nuclear Information System (INIS)

    Zhang, Guoliang; Wang, Boyu; Shi, Lei; Tan, Xiaohua; Xiang, Wei

    2013-01-01

    While alloy materials is irradiated by the high-intensity pulsed ion beams (HIPIB), the temperature distributions surrounding the primary heated regions used numerical analysis has been studied extensively over the past few years. Compared with the temperature distributions induced by HIPIB, few information is known about the temperature distributions on alloy materials used in practice as it is irradiated by the pulsed ion beams which possess characteristics of lower energy density and longer pulse width. The main reason is that the interaction between the alloy materials and the pulsed ion beams is only a few microseconds. It is difficult to detect temperature changes on alloy materials used in practice through traditional test. Ablation, melting, defects of microstructure on alloy materials are always used to validate the results of numerical analysis about the temperature distributions indirectly. In order to evaluate the temperature distributions directly, the dynamic thermal-dependent temperature behavior of the low-temperature alloy irradiated by the pulsed ion beams is investigated by experimental observation and finite element method (FEM) simulation in this paper. The temperature profiles generated from the interaction of μs-size between the alloy materials and the pulsed ion beams are evaluated by coupling characteristics of the low melting point and the pulsed ion beams. The FEM simulation results of the maximum temperature agree well with the experimental results on the surface of the low-temperature alloy. Results gained show that the maximum temperature on the surface of the low-temperature alloy irradiated by the pulsed ion beams can be applied to deduce the maximum temperature on alloy materials used in practice

  4. Dose dependence of irradiation hardening of neutron irradiated vanadium alloys by using temperature control rig in JMTR

    Directory of Open Access Journals (Sweden)

    Ken-ichi Fukumoto

    2016-12-01

    Full Text Available TEM observation and tensile test were examined for vanadium alloys irradiated in a temperature control rig in JMTR at 290°C with damage level ranged from 0.003 to 0.06dpa. With the increase of the neutron dose, irradiation hardening could be observed in all the vanadium alloys except for the V–5Nb alloy. In the case of pure vanadium, the relationship between irradiation hardening and neutron dose was described as Δσ ∝ (ϕt0.35-0.53. For V–5Cr alloy and V–4Cr–4Ti–0.1Si alloy, the dose dependence on irradiation hardening increase was shown as Δσ ∝ (ϕt0.8 and Δσ ∝ (ϕt0.8-1.0, respectively. From the TEM observation, the hardening source of radiation-induced defects was mainly determined to be dislocation loops for pure vanadium, loops with voids for V–5Cr and, loops and {100} precipitates for V–4Cr–4Ti–0.1Si and V–3Fe–4Ti–0.1Si alloys. From the strain rate dependence of 8% stress for V–4Cr–4Ti–0.1Si alloys tested at RT, the strain rate sensitivity, m=1/σ*(dσ/dln(dε/dt shows positive. Therefore, the dynamic interaction between interstitial impurities and dislocation is not strong in V–4Cr–4Ti alloys in the temperature range from RT to 290°C. A discrepancy of deformation mode of irradiated V–4Cr–4Ti–0.1Si alloys with 0.068dpa could be seen when the charpy impact test indicated the brittle behavior and the tensile test indicated the ductile behavior at room temperature. It can be explained by the difference of strain rate for the value of yield stress between tensile test and charpy test and the critical fracture stress.

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

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.

    1996-01-01

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

  6. Study of anti-laser irradiation performance of shot-peened 40CrNiMoA alloy steel

    International Nuclear Information System (INIS)

    Liu, Zhanwei; Wu, Ningning; Huang, Xianfu; Xie, Huimin; Lv, Xintao; He, Guang

    2012-01-01

    In this paper, shot-peening treatment was introduced to reinforce an alloy surface to protect it from laser irradiation, and experiments were carried out on 40CrNiMoA alloy steel. Macro-mechanical properties were studied and compared before and after both shot-peening and laser irradiation by conducting tensile and hardness measurements. Experimental results showed that the shot-peened alloy showed better mechanical properties after laser irradiation when compared to the alloy without shot-peening treatment. The enhanced ability of the shot-peened alloy for anti-laser irradiation was explained as due to the large residual compressive stress distributions over the shot-peening layer greatly reducing the thermal shock effect introduced by the laser. On the other hand, the growth of microstructures in specific shape absorbed the thermal energy during irradiation, giving a higher probability for the alloy to resist damage.

  7. The effect of neutron irradiation on the mechanical properties of precipitation hardened copper alloys

    International Nuclear Information System (INIS)

    Fabritsiev, S.A.; Pokrovsky, A.S.

    1997-01-01

    The effects of neutron irradiation on strength and ductility properties of precipitation hardened (PH) copper alloys are discussed. The analysis is based on the experimental study of radiation damage of PH alloys irradiated in the mixed spectrum reactor SM-2 to fluences of 3.7-5.5 x 10 25 n/m 2 (E>0.1 MeV), corresponding to NRT displacement dose levels of 2.6-3.8 dpa. At irradiation temperatures of 100-285 C the processes of radiation hardening and reduction in the uniform elongation are the major effects. Irradiation at temperatures higher than 300 C causes a dramatic softening and improvement in uniform elongation of the Cu-Cr-Zr and Cu-Cr-Zr-Mg alloys. The threshold softening temperature for the PH alloys is shown to be about 300 C at a dose of 4.5 x 10 25 n/m 2 (E>0.1 MeV). The effect of the irradiation dose and temperature on the shift of the threshold temperature of PH copper-alloys softening is also considered. (orig.)

  8. Microstructure and damage behavior of W-Cr alloy under He irradiation

    Science.gov (United States)

    Huang, Ke; Luo, Lai-Ma; Zan, Xiang; Xu, Qiu; Liu, Dong-Guang; Zhu, Xiao-Yong; Cheng, Ji-Gui; Wu, Yu-Cheng

    2018-04-01

    In this study, a large-power inductively coupled plasma source was designed to perform the continuous helium ion irradiations of W-Cr binary alloy (W-20 wt%Cr) under relevant conditions of the International Thermonuclear Experimental Reactor. Surface damages and microstructures of irradiated W-20Cr were observed by using scanning electron microscopy, energy-dispersive X-ray spectroscopy, and transmission electron microscopy. The addition of Cr dramatically enhanced the micro-hardness of the obtained bulk materials, and the interface between the W matrix and the second phase Cr-O is a semi-coherent interface. After irradiation, the doping of Cr element effectively reduces the damage of the W matrix during the irradiation process. The semi-coherent interface between the second phase and the W matrix improves the anti-irradiation performance of the W-20Cr alloy.

  9. Effects of iron irradiation om microstructure and properties of zirconium alloys: A review

    International Nuclear Information System (INIS)

    Yan, Chun Guang; Wang, Yanli; Wang, Xitao; Wang, Rong Shan; Bai, Guang Hai

    2015-01-01

    Zirconium alloys are widely used in nuclear reactors as structural materials. During the operation, they are exposed to fast neutrons. Ion irradiation is used to simulate the damage introduced by neutron irradiation. In this article, we briefly review the neutron irradiation damage of zirconium alloys, then summarize the effect of ion irradiation on microstructural evolution, mechanical and corrosion properties, and their relationships. The microstructure components consist of dislocation loops, second phase precipitates, and gas bubbles. The microstructure parameters are also included such as domain size and microstrain determined by X-ray diffraction and the S-parameter determined by positron annihilation. Understanding the relationships of microstructure and properties is necessary for developing new advanced materials with higher irradiation tolerance.

  10. 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.

  11. A phase field model for segregation and precipitation induced by irradiation in alloys

    International Nuclear Information System (INIS)

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

    2015-01-01

    A phase field model is introduced to model the evolution of multicomponent alloys under irradiation, including radiation-induced segregation and precipitation. The thermodynamic and kinetic components of this model are derived using a mean-field model. The mobility coefficient and the contribution of chemical heterogeneity to free energy are rescaled by the cell size used in the phase field model, yielding microstructural evolutions that are independent of the cell size. A new treatment is proposed for point defect clusters, using a mixed discrete-continuous approach to capture the stochastic character of defect cluster production in displacement cascades, while retaining the efficient modeling of the fate of these clusters using diffusion equations. The model is tested on unary and binary alloy systems using two-dimensional simulations. In a unary system, the evolution of point defects under irradiation is studied in the presence of defect clusters, either pre-existing ones or those created by irradiation, and compared with rate theory calculations. Binary alloys with zero and positive heats of mixing are then studied to investigate the effect of point defect clustering on radiation-induced segregation and precipitation in undersaturated solid solutions. Lastly, irradiation conditions and alloy parameters leading to irradiation-induced homogeneous precipitation are investigated. The results are discussed in the context of experimental results reported for Ni–Si and Al–Zn undersaturated solid solutions subjected to irradiation. (paper)

  12. Neutron diffraction analysis of Cr–Ni–Mo–Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    International Nuclear Information System (INIS)

    Voronin, V.I.; Valiev, E.Z.; Berger, I.F.; Goschitskii, B.N.; Proskurnina, N.V.; Sagaradze, V.V.; Kataeva, N.F.

    2015-01-01

    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

  13. Neutron diffraction analysis of Cr-Ni-Mo-Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    Science.gov (United States)

    Voronin, V. I.; Valiev, E. Z.; Berger, I. F.; Goschitskii, B. N.; Proskurnina, N. V.; Sagaradze, V. V.; Kataeva, N. F.

    2015-04-01

    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.

  14. Effect of helium on swelling and microstructural evolution in ion-irradiated V-15Cr-5Ti alloy

    International Nuclear Information System (INIS)

    Loomis, B.A.; Kestel, B.J.; Gerber, S.B.; Ayrault, G.

    1986-03-01

    An investigation was made on the effects of implanted helium on the swelling and microstructural evolution that results from energetic single- and dual-ion irradiation of the V-15Cr-5Ti alloy. Single-ion irradiations were utilized for a simulated production of the irradiation damage that might be expected from neutron irradiation of the alloy in a reactor with a fast neutron energy spectrum (E > 0.1 MeV). Dual-ion irradiations were utilized for a simulated production of the simultaneous creation of helium atoms and irradiation damage in the alloy in the MFR environment. Experimental results are also presented on the radiation-induced segregation of the constituent atoms in the single- and dual-ion irradiated alloy

  15. Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

    International Nuclear Information System (INIS)

    Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

    1997-01-01

    This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

  16. Microstructures of neutron-irradiated Fe-12Cr-XMn (X=15-30) ternary alloys

    International Nuclear Information System (INIS)

    Miyahara, K.; Hosoi, Y.; Garner, F.A.

    1992-01-01

    The objective of this effort is to determine the factors which control the stability of irradiated alloys proposed for reduced activation applications. The Fe-Cr-Mn alloy system is being studied as an alternative to the Fe-Cr-Ni system because of the need to reduce long-term radioactivation in fusion-power devices. In this study, four Fe-12Cr-XMn (X =15, 20, 25, 30 wt%) alloys were irradiated in the Fast Flux Test Facility to 20 dpa at 643K and 40 dpa at 679, 793, and 873K to investigate the influence of manganese content on void swelling and phase stability. The results confirm and expand the results of earlier studies that indicate that the Fe-Cr-Mn system is relatively unstable compared to that of the Fe-Cr-Ni system, with alpha and sigma phases forming as a consequence of thermal aging or high temperature irradiation

  17. Void formation in NiTi shape memory alloys by medium-voltage electron irradiation

    International Nuclear Information System (INIS)

    Schlossmacher, P.; Stober, T.

    1995-01-01

    In-situ electron irradiation experiments of NiTi shape memory alloys, using high-voltage transmission electron microscopes, result in amorphization of the intermetallic compound. In all of these experiments high-voltages more than 1.0 MeV had to be applied in order to induce the crystalline-to-amorphous transformation. To their knowledge no irradiation effects of medium-voltage electrons of e.g. 0.5 MeV have been reported in the literature. In this contribution, the authors describe void formation in two different NiTi shape memory alloys, resulting from in-situ electron irradiation, using a 300 kV electron beam in a transmission electron microscope. First evidence is presented that void formation is correlated with the total oxygen content of the alloys

  18. Electron irradiation-induced nanocrystallization of amorphous Fe85B15 alloy: Evidence for athermal nature

    International Nuclear Information System (INIS)

    Qin, W.; Nagase, T.; Umakoshi, Y.

    2009-01-01

    Nanocrystallization of amorphous alloys induced by electronic energy deposition has been frequently reported in recent years. In this paper, the crystallization of amorphous Fe 85 B 15 alloy was performed by electron irradiation with 2 MeV electrons up to a flux of 4.0 x 10 24 m -2 s -1 . It was found that at 298 K, nanocrystalline Fe-B intermetallic phases formed prior to α-Fe phase, while at 463 K, only the α-Fe phase was observed. This phenomenon cannot be interpreted in terms of the electron-beam heating, but may be attributed to the irradiation-induced increases in the short-range order and atomic diffusivity. Theoretical analysis also showed that the maximum-temperature rise driven by beam heating is much lower than that required for thermal crystallization. Our work offers strong evidence that the irradiation-induced crystallization in amorphous alloys is not a thermal activation process

  19. 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 * magneto elasticity 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

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

    Science.gov (United States)

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

    2015-07-01

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

  1. Microstructure and phase transformations in the ODS alloys irradiated by swift heavy ions

    International Nuclear Information System (INIS)

    Zlotski, S.V.; Anishchik, V.M; Skuratov, V.A.; O’Connell, J.; Neethling, J.H.

    2015-01-01

    Microstructure of KP4 ODS alloy irradiated with 700 MeV bismuth ions at 300 K has been studied using high resolution transmission electron microscopy. No latent tracks have been observed in Y 4 Al 2 O 9 particles in KP4 irradiated with Bi ions. Small oxides (~ 5 nm) in KP4 alloy remain crystalline at Bi ion fluence 1.5*10 13 cm -2 , while subsurface regions in large (~ 20 nm) particles faced to the beam entrance became amorphous. (authors)

  2. Defect microstructure in copper alloys irradiated with 750 MeV protons

    DEFF Research Database (Denmark)

    Zinkle, S.J.; Horsewell, A.; Singh, B.N.

    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...... significant effect on the total density of small defect clusters, but they did cause a significant decrease in the fraction of defect clusters resolvable as SFT to similar to 20 to 25%. In addition, the dislocation loop density (> 5 nm diameter) was more than an order of magnitude higher in the alloys...

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

    International Nuclear Information System (INIS)

    Idrees, Y.; Yao, Z.; Sattari, M.; Daymond, M.R.

    2012-01-01

    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 100 o C-400 o C. 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)

  4. 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)

    Shahrabi, T.

    2004-01-01

    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 NaNO 3 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)

  5. Study by transmission electron microscopy of ion and electron irradiation defects in zirconium and four alloys

    International Nuclear Information System (INIS)

    Hellio, C.

    1988-01-01

    In order to understand better zirconium growth under irradiation and to estimate the influence of alloying elements, we have performed crystallographic and kinetic analysis of the dislocation loops induced during 500 keV Zr + ion or 1 MeV electron irradiation in pure Zr and four of its alloys: Zr/1760 wt ppm 0. Zr/1 % Nb/430 ppm 0. Zr/1 % Nb/1800 ppm 0 and Zircaloy- 4 . The irradiations were realized between 400 and 700 0 C. The Burgers vectors of the observed loops are a/3 ; the loops lie preferentially on (1010) planes and are aligned along directions parallel to the (0001) plane. The loops induced by 1 MeV electron irradiation and of size larger than 10 nm are of interstitial type. The kinetic analysis shows that oxygen strongly decreases the loop groth speed, the apparent vacancy migration energy increasing from 0.72 eV for pure Zr to 1.76 eV for Zr/1750 ppm 0. The alloying elements (Sn, Nb, 0) increase the loop density. Observation of the microstructural behaviour shows that loops shrink and aggregate during annealing treatments after ion irradiation. This annealing is easier in pure zirconium than in its alloys [fr

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

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-12-01

    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

  7. 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.

  8. Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

    Science.gov (United States)

    Long, Fei

    Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion

  9. Processing of Refractory Metal Alloys for JOYO Irradiations

    International Nuclear Information System (INIS)

    RF Luther; ME Petrichek

    2006-01-01

    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

  10. 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.

  11. Effect of 50-keV proton irradiation on the magnetism of a Fe66Ni34 Invar alloy

    Science.gov (United States)

    Matsushita, M.; Matsushima, Y.; Uruga, T.; Ishigami, R.; Iwase, A.

    2013-05-01

    The magnetism of Fe-Ni Invar alloys is very sensitive to the lattice constant, stress, and the number of nearest-neighbor Fe-Fe atomic pairs. Ion irradiation is a useful tool to alter the local atomic structure of a given material. Therefore, the effects of low-energy and light-ion irradiation on the magnetism of a Fe66Ni34 Invar alloy were investigated in this study. The Fe66Ni34 Invar alloy was irradiated with 50-keV protons at a fluence of 1×1015 ions/cm2 at room temperature. The Curie temperature was found to increase from 465 K (before irradiation) to 535 K (after irradiation). The X-ray absorption analysis of the fine structure of the alloy revealed that irradiation had no effect on the atomic structures surrounding Fe and Ni.

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

    International Nuclear Information System (INIS)

    Luo, Dian; Tang, Guangze; Ma, Xinxin; Gu, Le; Sun, Mingren; Wang, Liqin

    2015-01-01

    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 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 2 ) and multi-number of irradiation (≥50 times) contributed to high quality of alloyed layers after irradiation

  13. Effect of irradiation on the critical currents of alloy and compound superconductors

    International Nuclear Information System (INIS)

    Sekula, S.T.

    1977-06-01

    The effects of energetic-particle irradiation on the critical-current density J/sub c/(H) of several superconducting compounds and Nb-Ti alloys have been examined by a number of workers. The irradiations used in the investigations include electrons, fast neutrons, ions, and fission fragments. The results of these studies are reviewed and summarized. In the alloys, changes in J/sub c/(H) on irradiation depend on the metallurgical history of the material and indicate that radiation defects modify the strength of the interaction between the fluxoid array and the sample microstructure. Radiation defects in alloys can also affect J/sub c/(H) through small decreases in T/sub c/, the transition temperature and rho, the normal-state resistivity. Irradiations of A15 compounds up to moderate fluences (dependent on the type and energy of irradiating particle) lead to decreases in T/sub c/ of approximately 1 0 K and increases in J/sub c/(H) with dose for most of the samples investigated. This result can be qualitatively understood as resulting from radiation-induced changes in rho and the pinning force acting on the fluxoids. At higher dose levels, significant depressions of T/sub c/ and possibly gamma, the electronic specific heat coefficient, lead to drastic reductions in J/sub c/(H). The effect of various energetic particles and irradiation temperature on changes in J/sub c/(H) are discussed

  14. Study of microstructural evolutions of the 6061-T6 aluminium alloy under irradiation

    International Nuclear Information System (INIS)

    Flament, Camille

    2015-01-01

    The 6061-T6 Aluminium alloy, whose microstructure contains Al(Fe,Mn,Cr)Si dispersoids and hardening needle-shaped β'' precipitates (Mg, Si), has been chosen as the structural material for the core vessel of the Material Testing Jules Horowitz Nuclear Reactor. Because it will be submitted to high neutron flux at a temperature around 50 C, it is necessary to study microstructural evolutions induced by irradiation and especially the stability of the second phase particles. In this work, an analytical study by in-situ and ex-situ electron and ion irradiations has been performed, as well as a study under neutron irradiation. The precipitate characterization by Transmission Electron Microscopy demonstrates that Al(Fe,Mn,Cr)Si dispersoids are driven under irradiation towards their equilibrium configuration, consisting of a core/shell structure, enhanced by irradiation, with a (Fe, Mn) enriched core surrounded by a Cr-enriched shell. In contrast, the (Mg,Si) β'' precipitates are destabilized by irradiation. They dissolve under ion irradiation in favor of a new precipitation of (Mg,Si,Cu,Cr,Al) rich clusters resulting in an increase of the alloy's hardness. β'' precipitates tend towards a transformation to cubic precipitates under neutron irradiation. (author) [fr

  15. TEM characterization of irradiated microstructure of Fe-9%Cr ODS and ferritic-martensitic alloys

    Science.gov (United States)

    Swenson, M. J.; Wharry, J. P.

    2018-04-01

    The objective of this study is to evaluate the effects of irradiation dose and dose rate on defect cluster (i.e. dislocation loops and voids) evolution in a model Fe-9%Cr oxide dispersion strengthened steel and commercial ferritic-martensitic steels HCM12A and HT9. Complimentary irradiations using Fe2+ ions, protons, or neutrons to doses ranging from 1 to 100 displacements per atom (dpa) at 500 °C are conducted on each alloy. The irradiated microstructures are characterized using transmission electron microscopy (TEM). Dislocation loops exhibit limited growth after 1 dpa upon Fe2+ and proton irradiation, while any voids observed are small and sparse. The average size and number density of loops are statistically invariant between Fe2+, proton, and neutron irradiated specimens at otherwise fixed irradiation conditions of ∼3 dpa, 500 °C. Therefore, we conclude that higher dose rate charged particle irradiations can reproduce the neutron irradiated loop microstructure with temperature shift governed by the invariance theory; this temperature shift is ∼0 °C for the high sink strength alloys studied herein.

  16. Irradiation creep and deformation under flux of austenitic stainless steels 304 and 316; Fluage d irradiation et deformation sous flux des aciers inoxydables austenitiques 304 et 316

    Energy Technology Data Exchange (ETDEWEB)

    Garnier, J.; Dubuisson, P. [CEA DEN-DANS/DMN/SRMA, Saclay 91191 Gif-sur-Yvette (France); Delnondedieu, M.; Massoud, J.P. [EDF R et D, MMC, Site des Renardieres 77818 Moret sur Loing (France); Brechet, Y. [LTPCM, BP75, 38402 St Martin d Heres (France)

    2006-07-01

    The materials constituting the PWR reactors vessels internals are submitted to a neutron flux, at a temperature between 280 and 380 C, and at mechanical solicitations. On account of the C. Pokor works, the irradiation effects are now well known; the following step is the combined study of the irradiation and the mechanical solicitation. In order to understand the mechanisms which induce the microstructural changes, irradiations have been carried out in the following experimental reactors: Osiris at 330 C until 10 dpa and BOR-60 at 330 C beyond 100 dpa. Two tests types have been studied: creep tests and deformation tests under flux. Transmission electronic microscopy analyses have allowed to quantify these microstructural changes, particularly the density and the size of the Frank loops. A behaviour law, developed by J. Besson (ENSMP) and S. Leclerq (EDF), integrating irradiation creep and plasticity of the material after irradiation, allows to describe the change of stress during the reactors tests. (O.M.)

  17. Silver and palladium alloy nanoparticle catalysts: reductive coupling of nitrobenzene through light irradiation.

    Science.gov (United States)

    Peiris, Sunari; Sarina, Sarina; Han, Chenhui; Xiao, Qi; Zhu, Huai-Yong

    2017-08-15

    Silver-palladium (Ag-Pd) alloy nanoparticles strongly absorb visible light and exhibit significantly higher photocatalytic activity compared to both pure palladium (Pd) and silver (Ag) nanoparticles. Photocatalysts of Ag-Pd alloy nanoparticles on ZrO 2 and Al 2 O 3 supports are developed to catalyze the nitroaromatic coupling to the corresponding azo compounds under visible light irradiation. Ag-Pd alloy NP/ZrO 2 exhibited the highest photocatalytic activity for nitrobenzene coupling to azobenzene (yield of ∼80% in 3 hours). The photocatalytic efficiency could be optimized by altering the Ag : Pd ratio of the alloy nanoparticles, irradiation light intensity, temperature and wavelength. The rate of the reaction depends on the population and energy of the excited electrons, which can be improved by increasing the light intensity or by using a shorter wavelength. The knowledge developed in this study may inspire further studies on Ag alloy photocatalysts and organic syntheses using Ag-Pd nanoparticle catalysts driven under visible light Irradiation.

  18. EFFECTS OF IRRADIATION ON THERMAL CONDUCTIVITY OF ALLOY 690 AT LOW NEUTRON FLUENCE

    Directory of Open Access Journals (Sweden)

    WOO SEOG RYU

    2013-04-01

    Full Text Available Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7−28 × 1019n/cm2 (E>0.1MeV at 250°C, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at 250°C did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1% at fluences of (0.7∼28 × 1019n/cm2 (E>0.1MeV.

  19. Effect of impurities on mechanical properties of vanadium alloys under liquid-lithium environment during neutron irradiation at HFIR

    Science.gov (United States)

    Fukumoto, K.; Kuroyanagi, Y.; Kuroiwa, H.; Narui, M.; Matsui, H.

    2011-10-01

    Vanadium alloys, including the highly purified V-4Cr-4Ti alloy, were irradiated in liquid lithium up to a damage level of 3.7 dpa in the HFIR at 425 °C and 598 °C. Neutron irradiation caused an increase of the ductile-brittle transition temperature (DBTT) and irradiation hardening was observed. Adding titanium to the V-Cr alloys was effective for increasing irradiation hardening at 425 °C. For highly purified (Zr-treated) V-4Cr-4Ti alloys the irradiation hardening was significantly reduced at both 425 °C and 598 °C. However, microstructural observations after the irradiation experiments showed that there was no significant difference in microstructure between the original and the highly purified specimens. It is suggested that the reduction of irradiation hardening in the highly purified V-4Cr-4Ti alloys was caused by the configuration and distribution of interstitial impurities in the neutron-irradiated specimen matrix. Controlling the impurities in V-4Cr-4Ti alloys has a very important effect for improving their mechanical properties that take place under neutron irradiation at around 400 °C.

  20. Correlation between intrinsic hardness and defect structures of ion irradiated Fe alloys

    International Nuclear Information System (INIS)

    Shin, C.; Jin, H. H.; Kwon, J.

    2008-01-01

    Evolution of micro structures and mechanical properties during an in-service irradiation is one of the key issues to be addressed in nuclear materials. Ion irradiation is an effective method to study these irradiation effects thanks to an ease in handling post-irradiated specimens. But the characteristics of an ion irradiation pose a certain difficulty in evaluating irradiation effects. For example, ion irradiated region extends only a few hundred nano-meters from the surface of a sample and the depth profile of an irradiation damage level is quite heterogeneous. Thus it requires special care to quantify the changes in properties after an ion irradiation. We measured changes in a hardness by using a nano-indentation combined with a continuous stiffness measurement (CSM technique. Although the SM technique allows for a continuous measurement of hardness along penetration depth of an indenter; it is difficult to obtain an intrinsic hardness of an irradiation hardened region because one is measuring hardness of a hard layer located on a soft matrix. Thus we modeled the nano-indentation test by using a finite element method. We can extract the intrinsic hardness and the yield stress of an irradiation hardened region by using a so-called inverse method. We investigated the irradiation effects on Fe-Cr binary alloy by using the methods mentioned above. TEM analysis revealed that an irradiation forms dislocation loops with Burgers vector of and 1/2 . These loops varied in size and density with the Cr content and dose level. We discuss in detail a correlation between the measured irradiation-induced changes in the surface hardness and an irradiation induced defect. (authors)

  1. 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.

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

    DEFF Research Database (Denmark)

    Barlow, P.; Leffers, Torben

    1977-01-01

    of 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......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...... 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...

  3. 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 ...

  4. Shear Punch Testing on ATR Irradiated MA956 FeCrAl Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-13

    The shear punch testing of irradiated and control MA956 (FeCrAl) Alloy from the NSUF-ATR-UCSB irradiation is presented. This is the first data taken on a new shear punch fixture design to test three 1.5mm punches from each 8mm x 0.5mm Disc Multipurpose Coupon (DMC). Samples were irradiated to 6.1dpa at a temperature of 315°C and 6.2 dpa at 400°C.

  5. Effects of energetic ion irradiation on the magnetism of Fe–Ni Invar alloy

    Energy Technology Data Exchange (ETDEWEB)

    Matsushita, M., E-mail: matsushita@eng.ehime-u.ac.jp [Graduate School of Science and Engineering, Ehime University, 3-Bunkyocho, Matsuyama (Japan); Akamatsu, S. [Graduate School of Science and Engineering, Ehime University, 3-Bunkyocho, Matsuyama (Japan); Matsushima, Y. [Graduate School of Natural Science and Technology, Okayama University, Tsushima-naka Kitaku, Okayama (Japan); Iwase, A. [Department of Materials Science, Osaka Prefecture University, Gakuencho, Sakai (Japan)

    2013-11-01

    Highlights: •16-MeV Au{sup 3+} ions were irradiated to Fe{sub 66}Ni{sub 34} alloy. •Magnetic properties of Fe{sub 66}Ni{sub 34} were changed by the irradiation. •T{sub c} of a part of sample increases due to the irradiation. •FCC structure is stable before and after irradiation. -- Abstract: The magnetic properties of Fe–Ni Invar alloys are significantly affected by ion irradiation. Au{sup 3+} with the energy of 16 MeV irradiation effects on the magnetism of Fe{sub 66}Ni{sub 34} have been reported in this paper. Considering from the temperature variations of AC susceptibility of irradiated Fe{sub 66}Ni{sub 34}, Curie temperature of a part of sample increase with increasing incident ion fluence, and the magnetization of irradiated Fe{sub 66}Ni{sub 34} is also increase. The FCC structure of Fe{sub 66}Ni{sub 34} is not changed by ion irradiation; however peaks become broader with increasing ion fluence. It means that lattice fluctuations are generated owing to ion irradiation. However it cannot be considered that lattice fluctuations observed X-ray diffraction measurements are enough to increase the Curie temperature observed in AC susceptibility measurements. Then, we suggest as the considerable origin of increasing T{sub C}, atomic mixing effects owing to the ion irradiation. It might change the chemical ordering reported in the diffused scattering, such as Fe–Fe coupling.

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

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1981-01-01

    A series of (Fe, Ni)/sub 3/V long-range-ordered alloys were irradiated with neutrons in the Oak Ridge Research Reactor (ORR) and with 4 MeV Ni ions at temperatures above 250/sup 0/C. 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.

  7. U.S. Contribution 1994 Summary Report Task T12: Compatibility and irradiation testing of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L. [comp.

    1995-03-01

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication.

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

    Directory of Open Access Journals (Sweden)

    Buxiang Zheng

    2014-02-01

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

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

    International Nuclear Information System (INIS)

    Onimus, F.

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  10. Accumulation of dislocation loops in the α phase of Zr Excel alloy under heavy ion irradiation

    Science.gov (United States)

    Yu, Hongbing; Yao, Zhongwen; Idrees, Yasir; Zhang, He K.; Kirk, Mark A.; Daymond, Mark R.

    2017-08-01

    In-situ heavy ion irradiations were performed on the high Sn content Zr alloy 'Excel', measuring type dislocation loop accumulation up to irradiation damage doses of 10 dpa at a range of temperatures. The high content of Sn, which diffuses slowly, and the thin foil geometry of the sample provide a unique opportunity to study an extreme case where displacement cascades dominate the loop formation and evolution. The dynamic observation of dislocation loop evolution under irradiation at 200 °C reveals that type dislocation loops can form at very low dose (0.0025 dpa). The size of the dislocation loops increases slightly with irradiation damage dose. The mechanism controlling loop growth in this study is different from that in neutron irradiation; in this study, larger dislocation loops can condense directly from the interaction of displacement cascades and the high concentration of point defects in the matrix. The size of the dislocation loop is dependent on the point defect concentration in the matrix. A negative correlation between the irradiation temperature and the dislocation loop size was observed. A comparison between cascade dominated loop evolution (this study), diffusion dominated loop evolution (electron irradiation) and neutron irradiation suggests that heavy ion irradiation alone may not be enough to accurately reproduce neutron irradiation induced loop structures. An alternative method is proposed in this paper. The effects of Sn on the displacement cascades, defect yield, and the diffusion behavior of point defects are established.

  11. Irradiation effect on the precipitation in Fe-Cr model alloys with around 15% of chromium

    International Nuclear Information System (INIS)

    Jaquet, Virginie

    2000-01-01

    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 10 19 per cm 3 . 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) [fr

  12. Post irradiation fracture properties of precipitation-strengthened alloy D21

    International Nuclear Information System (INIS)

    Huang, F.H.

    1986-03-01

    The precipitation strengthened alloys have the potential for use in fuel cladding and duct applications for liquid metal reactors due to their high strength and low swelling rate. Unfortunately, these high strength alloys tend to exhibit poor fracture toughness, and the effects of neutron irradiation on the fracture properties of the material are of concern. Compact tension specimens of alloy D21 were irradiated in the Experimental Breeder Reactor II to a fluence of 2.7 x 10 22 n/cm 2 (E > 0.1 MeV) at 425, 500, 550 and 600 0 C. Fracture toughness tests on these specimens wre performed using electric potential techniques at temperatures ranging from 205 to 425 C. The material exhibited low postirradiation fracture toughness which increased with either increasing test or irradiation temperature. The tearing modulus, however, increased with increasing irradiation temperature but decreased with increasing test temperature. Results wre analyzed using the J-integral approach. The fracture toughness of irradiated D21 was evaluated essentially following the procedure recommended in ASTM Test Method E813. It was found that the data elimination limits illustrated in E813 were too large for the specimens tested, although the thickness criterion was satisfied. The precautions needed to determine J/sub 1c/ based on a reduced data qualification range were disussed

  13. Internal friction in Al alloys after neutron irradiation at low temperature

    International Nuclear Information System (INIS)

    Takamura, S.; Kobiyama, M.

    1985-01-01

    Internal friction and elastic modulus of dilute Al alloys have been measured after fast neutron irradiation at about 5 K. The internal friction spectra in Al-Pb, Al-Si, Al-Zn, Al-Ag, Al-Sn and Al-In are very similar. This result suggests that the configuration of the interstitial-solute atom complex in these alloys is very similar. In Al-Mg, the main complexes have the configuration with nearly symmetry, but its internal friction spectrum is different from that of the above-mentioned alloys. The internal friction spectra and their annealing behavior in Al-Be, Al-Mn, Al-Fe and Al-Cu demonstrate that the configuration of their interstitial-solute atom complex seems to be different from each other and the main complex in these alloys is immobile until stage III. (author)

  14. Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-01-01

    In order to characterize the microstructural evolution of 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 through small angle neutron scattering and atom probe experiments. 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 such as 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

  15. Ductility and microstructure of precipitation-strengthened alloys irradiated in HFIR

    International Nuclear Information System (INIS)

    Yang, W.J.S.; Hamilton, M.L.

    1983-08-01

    Six γ' and γ'/γ'' strengthened Ni-base alloys have shown near-zero ductility after irradiation at 300 to 600 0 C in HFIR to a peak exposure of 9 dpa. Microstructural examination of the irradiated specimens showed that the loss of ductility in these alloys arises from the simultaneous existence of a strong matrix and weak grain boundaries. The strong matrix is attributed to the irradiation-induced γ' and γ'/γ'' precipitates, the faulted loops and a high density of fine helium bubbles. The weak grain boundaries are attributed to the formation of an unfavorable precipitate, such as eta-plates, recrystallized grains, a thin layer of γ' and helium bubbles

  16. Minaturized disk bend tests of neutron-irradiated path A type alloys

    International Nuclear Information System (INIS)

    Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

    1983-01-01

    Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

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

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, B.N.

    1998-01-01

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

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

    Science.gov (United States)

    Eldrup, M.; Singh, B. N.

    1998-10-01

    The electrical conductivity of three different types of copper alloys, viz. CuNiBe, CuCrZr and Cu-Al 2O 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 (⩽55% of that of pure Cu), and Cu-Al 2O 3 the highest (75-90% of pure Cu).

  19. Solidification cracking in austenitic stainless steel welds

    Indian Academy of Sciences (India)

    Solidification cracking is a significant problem during the welding of austenitic stainless steels, particularly in fully austenitic and stabilized compositions. Hot cracking in stainless steel welds is caused by low-melting eutectics containing impurities such as S, P and alloy elements such as Ti, Nb. The WRC-92 diagram can be ...

  20. The response of dispersion-strengthened copper alloys to high fluence neutron irradiation at 415 degrees C

    International Nuclear Information System (INIS)

    Edwards, D.J.; Newkirk, J.W.; Garner, F.A.; Hamilton, M.L.; Nadkarni, A.; Samal, P.

    1993-01-01

    Various oxide-dispersion-strengthened copper alloys have been irradiated to 150 dpa at 415 degrees C in the Fast Flux Test Facility (FFTF). The Al 2 O 3 -strengthened GlidCop TM alloys, followed closely by a HfO 2 -strengthened alloy, displayed the best swelling resistance, electrical conductivity, and tensile properties. The conductivity of the HfO 2 -strengthened alloy reached a plateau at the higher levels of irradiation, instead of exhibiting the steady decrease in conductivity observed in the other alloys. A high initial oxygen content results in significantly higher swelling for a series of castable oxide-dispersion-strengthened alloys, while a Cr 2 O 3 -strengthened alloy showed poor resistance to radiation

  1. XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

    International Nuclear Information System (INIS)

    Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

    1996-01-01

    Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

  2. Dislocation loop formation in model FeCrAl alloys after neutron irradiation below 1 dpa

    Science.gov (United States)

    Field, Kevin G.; Briggs, Samuel A.; Sridharan, Kumar; Yamamoto, Yukinori; Howard, Richard H.

    2017-11-01

    FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited details exist on dislocation loop formation and growth for this class of alloys under neutron irradiation. Four model FeCrAl alloys with chromium contents ranging from 10.01 to 17.51 wt % and aluminum contents of 4.78 to 2.93 wt % were neutron irradiated to doses of 0.3-0.8 displacements per atom (dpa) at temperatures of 335-355 °C. On-zone STEM imaging revealed a mixed population of black dots and larger dislocation loops with either a / 2 〈 111 〉 or a 〈 100 〉 Burgers vectors. Weak composition dependencies were observed and varied depending on whether the defect size, number density, or ratio of defect types was of interest. Results were found to mirror those of previous studies on FeCrAl and FeCr alloys irradiated under similar conditions, although distinct differences exist.

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

    International Nuclear Information System (INIS)

    Motta, A.T.; Faldowski, J.A.; Okamoto, P.R.

    1996-01-01

    The High Voltage Electron Microscope (HVEM)/Tandem facility at Argonne National Laboratory has been used to conduct detailed studies of the phase stability and microstructural evolution in zirconium alloys and compounds under ion and electron irradiation. Detailed kinetic studies of the crystalline-to-amorphous transformation of the intermetallic compounds Zr 3 (Fe 1-x Ni x ), Zr(Fe 1-x ,Cr x ) 2 , Zr 3 Fe, and Zr 1.5 Nb 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 ω-phase in Zr-20%Nb under electron and Ar ion irradiation was investigated as well as the β-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

  4. Semiquantitative activation analysis in metallic alloys submitted to irregular irradiation

    International Nuclear Information System (INIS)

    Veissid, N.; Lucki, G.

    1979-01-01

    An analytic semiquantitative method using neutron activation was developed to determine the impurities and verify the composition of metallic alloys. By the radioactive transformation law, the number of atoms of each element present in the sample is determined measuring the activity in a multichannel. Two samples were analysed: a) Sample of nominal compositions FeNiCr (49,95-49,95 - 0,1% at). b) Sample of nominal composition NiCr (80,20% at). (Author) [pt

  5. Microstructure analysis of magnesium alloy melted by laser irradiation

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. 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.

  7. Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

    International Nuclear Information System (INIS)

    Verlet, Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

  9. 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.

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

    International Nuclear Information System (INIS)

    Onimus, Fabien

    2015-01-01

    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

  11. 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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  13. 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.

  14. A brief review of cavity swelling and hardening in irradiated copper and copper alloys

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1990-01-01

    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 ∼10 24 n/m 2 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

  15. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm 3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm 3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  16. Irradiation effects in alloys Hg1-x CdxTe

    International Nuclear Information System (INIS)

    Favre, J.; Konczykowski, M.; Ecole Polytechnique, 91 - Palaiseau; Blanchard, C.

    1989-01-01

    The Hall coefficient R H and resistivity ρ were measured in situ during low temperature irradiations by electrons and ions. It has been shown that excess free electrons were provided by created mercury defects. The position of the associated level is determined for various mercury contents. In CdTe, it lies in the gap; otherwise, it is degenerated with the conduction band. The analysis of the rate of the defect creation versus energy of impinging electrons and mercury content indicates that the energy threshold and charge are independent of the composition. The effects of C, O, Xe ions irradiations are similar to those of electrons. The stability of created defects strongly depends on mercury content [fr

  17. Internal friction study of neutron-irradiation effects on an amorphous Cu40Ti60 alloy

    International Nuclear Information System (INIS)

    Dong, Y.; Wu, G.; Xiao, K.; Li, X.; He, Y.

    1988-01-01

    Effects of neutron irradiation on the structure of an amorphous Cu 40 Ti 60 alloy have been studied by internal friction measurements. After irradiation, the position of the first internal friction peak remains almost unchanged and the shoulder position shifts towards a higher temperature by about 5 K, which indicates that the Cu 40 Ti 60 glass becomes more stable. These results are finally discussed based on the concept of changes of chemical short-range ordering and geometrical short-range ordering due to radiation damage

  18. Subthreshold displacement damage in copper--aluminum alloys during electron irradiation

    International Nuclear Information System (INIS)

    Drosd, R.; Kosel, T.; Washburn, J.

    1976-12-01

    During electron irradiation at low energies which results in a negligible damage rate in a pure material, lighter solute atoms are displaced, which may in turn indirectly displace solvent atoms by a focussed replacement collision or an interstitial diffusion jump. The extent to which lighter solute atoms contribute to the subthreshold damage rate has been examined by irradiating copper--aluminum alloys at high temperatures in a high voltage electron microscope. The damage rate, as measured by monitoring the growth rate of dislocation loops, at 300 kV was found to increase linearly with the aluminum concentration

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

    International Nuclear Information System (INIS)

    1986-02-01

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

  20. Impact of β- radiolysis and transient products on irradiation-enhanced corrosion of zirconium alloys

    International Nuclear Information System (INIS)

    Lemaignan, C.

    1992-01-01

    An analysis has been undertaken of the various cases of local enhancement of the corrosion rate of zirconium alloys under irradiation. It is observed that in most cases a strong emission of energetic β - is present leading to a local energy desorption rate higher than the core average. This suggests that the local transient radiolytic oxidising species produced in the coolant by the β - particles could contribute to corrosion enhancement, by increasing the local corrosion potential. This process is applicable to the local enhanced corrosion found in front of stainless steels structural parts, due to the contribution of Mn, in front of Pt inserts and Cu-rich cruds. It explains also the irradiation corrosion enhancement of Cu-rich Zr alloys. Enhanced corrosion around neutron absorbing material is explained similarly by pair production from conversion of high energy capture photons in the cladding, leading to energetic electrons. The same process was found to be active with other highly ionising species like α in Ni-rich alloys and fission products in homogeneous reactors. This mechanism, applicable for an explanation of localised irradiation-enhanced corrosion, is proposed to be extended to the reactor core, where the general enhancement of Zr-alloy corrosion under irradiation would be due to the general radiolysis. It suggests that care should be taken to avoid any source of β - emission or other ionising species in the reactor core that could give an increase of energy deposition rate for radiolysis. Also the corrosion testing conditions for the materials to be used in reactors have to be relevant to the radiolytic environments found in the reactor cores. (orig.)

  1. Controlling radiation induced segregation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Ahmedabadi, Parag M.; Kain, Vivekanand

    2011-01-01

    In-core components of austenitic stainless steels in light water reactors (LWRs) are susceptible to irradiation assisted stress corrosion cracking (IASCC) in high temperature and high pressure oxygenated water at temperature around 300 deg C . Though, the exact mechanism for IASCC is not fully understood, radiation-induced segregation (RIS) is considered to be a part of a complex process that leads to IASCC. Therefore, controlling RIS in austenitic stainless steels may lead to improvement in resistance to IASCC. RIS is non-equilibrium segregation/depletion of alloying elements in austenitic stainless steels at LWR operating temperatures. RIS occurs due to adsorption of point defects at grain boundaries and leads to segregation of Si and P and depletion of Cr at grain boundaries. Thus by controlling point defect flux towards grain boundaries, the extent of RIS at grain boundaries can be controlled. An extensive study was carried out to simulate and control RIS in austenitic stainless steels using proton irradiation at 300 deg C . The primary aim of this study was to reduce point defect flux towards grain boundaries. Various approaches viz. grain boundary engineering, addition of oversized alloying element, residual strain within matrix and presence of fine precipitates within the grains and at grain boundaries were employed to control RIS in austenitic stainless steels. A novel approach involving combination of electrochemical technique followed by atomic force microscopic (AFM) examination has been used to examine the nature and the extent of RIS. Type 304, 316 and 347 stainless steels were irradiated at 300 deg C (in FOTIA and PELLETRON) in the range of 0.2 to 1.0 dpa using proton beam. The results obtained so far have indicated that a small amount of pre-strain within the grains is very effective in reducing the flux of point defects towards grain boundaries and reducing the extent of RIS at grain boundaries. The presence of NbC precipitates within the grains is

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

    Science.gov (United States)

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

    2016-04-01

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

  3. Numerical atomic scale simulations of the microstructural evolution of ferritic alloys under irradiation

    International Nuclear Information System (INIS)

    Vincent, E.

    2006-12-01

    In this work, we have developed a model of point defect (vacancies and interstitials) diffusion whose aim is to simulate by kinetic Monte Carlo (KMC) the formation of solute rich clusters observed experimentally in irradiated FeCuNiMnSi model alloys and in pressure vessel steels. Electronic structure calculations have been used to characterize the interactions between point defects and the different solute atoms. Each of these solute atoms establishes an attractive bond with the vacancy. As for Mn, which is the element which has the weakest bond with the vacancy, it establishes more favourable bonds with interstitials. Binding energies, migration energies as well as other atomic scale properties, determined by ab initio calculations, have led to a parameter set for the KMC code. Firstly, these parameters have been optimised on thermal ageing experiments realised on the FeCu binary alloy and on complex alloys, described in the literature. The vacancy diffusion thermal annealing simulations show that when a vacancy is available, all the solutes migrate and form clusters, in agreement with the observed experimental tendencies. Secondly, to simulate the microstructural evolution under irradiation, we have introduced interstitials in the KMC code. Their presence leads to a more efficient transport of Mn. The first simulations of electron and neutron irradiations show that the model results are globally qualitatively coherent with the experimentally observed tendencies. (author)

  4. Effect of heavy ion irradiation on thermodynamically equilibrium Zr-Excel alloy

    Science.gov (United States)

    Yu, Hongbing; Liang, Jianlie; Yao, Zhongwen; Kirk, Mark A.; Daymond, Mark R.

    2017-05-01

    The thermodynamically equilibrium state was achieved in a Zr-Sn-Nb-Mo alloy by long-term annealing at an intermediate temperature. The fcc intermetallic Zr(Mo, Nb)2 enriched with Fe was observed at the equilibrium state. In-situ 1 MeV Kr2+ heavy ion irradiation was performed in a TEM to study the stability of the intermetallic particles under irradiation and the effects of the intermetallic particle on the evolution of type dislocation loops at different temperatures from 80 to 550 °C. Chemi-STEM elemental maps were made at the same particles before and after irradiation up to 10 dpa. It was found that no elemental redistribution occurs at 200 °C and below. Selective depletion of Fe was observed from some precipitates under irradiation at higher temperatures. No change in the morphology of particles and no evidence showing a crystalline to amorphous transformation were observed at all irradiation temperatures. The formation of type dislocation loops was observed under irradiation at 80 and 200 °C, but not at 450 and 550 °C. The loops were non-uniformly distributed; a localized high density of type dislocation loops were observed near the second phase particles; we suggest that loop nucleation is favored as a result of the stress induced by the particles, rather than by elemental redistribution. The stability of the second phase particles and the formation of the type loops under heavy ion irradiation are discussed.

  5. Effects of energetic ion irradiation on the magnetism of Fe-Ni Invar alloy

    Science.gov (United States)

    Matsushita, M.; Akamatsu, S.; Matsushima, Y.; Iwase, A.

    2013-11-01

    The magnetic properties of Fe-Ni Invar alloys are significantly affected by ion irradiation. Au3+ with the energy of 16 MeV irradiation effects on the magnetism of Fe66Ni34 have been reported in this paper. Considering from the temperature variations of AC susceptibility of irradiated Fe66Ni34, Curie temperature of a part of sample increase with increasing incident ion fluence, and the magnetization of irradiated Fe66Ni34 is also increase. The FCC structure of Fe66Ni34 is not changed by ion irradiation; however peaks become broader with increasing ion fluence. It means that lattice fluctuations are generated owing to ion irradiation. However it cannot be considered that lattice fluctuations observed X-ray diffraction measurements are enough to increase the Curie temperature observed in AC susceptibility measurements. Then, we suggest as the considerable origin of increasing TC, atomic mixing effects owing to the ion irradiation. It might change the chemical ordering reported in the diffused scattering, such as Fe-Fe coupling.

  6. Spectral and raw quasi in-situ energy dispersive X-ray data captured via a TEM analysis of an ODS austenitic stainless steel sample under 1 MeV Kr2+ high temperature irradiation.

    Science.gov (United States)

    Brooks, Adam J; Yao, Zhongwen

    2017-10-01

    The data presented in this article is related to the research experiment, titled: ' Quasi in-situ energy dispersive X-ray spectroscopy observation of matrix and solute interactions on Y-Ti-O oxide particles in an austenitic stainless steel under 1 MeV Kr 2+ high temperature irradiation' (Brooks et al., 2017) [1]. Quasi in-situ analysis during 1 MeV Kr 2+ 520 °C irradiation allowed the same microstructural area to be observed using a transmission electron microscope (TEM), on an oxide dispersion strengthened (ODS) austenitic stainless steel sample. The data presented contains two sets of energy dispersive X-ray spectroscopy (EDX) data collected before and after irradiation to 1.5 displacements-per-atom (~1.25×10 -3  dpa/s with 7.5×10 14  ions cm -2 ). The vendor software used to process and output the data is the Bruker Esprit v1.9 suite. The data includes the spectral (counts vs. keV energy) of the quasi in-situ scanned region (512×512 pixels at 56k magnification), along with the EDX scanning parameters. The.raw files from the Bruker Esprit v1.9 output are additionally included along with the.rpl data information files. Furthermore included are the two quasi in-situ HAADF images for visual comparison of the regions before and after irradiation. This in-situ experiment is deemed ' quasi' due to the thin foil irradiation taking place at an external TEM facility. We present this data for critical and/or extended analysis from the scientific community, with applications applying to: experimental data correlation, confirmation of results, and as computer based modeling inputs.

  7. Alternative Zr alloys with irradiation resistant precipitates for high burnup BWR application

    International Nuclear Information System (INIS)

    Garzarolli, F.; Ruhmann, H.; Van Swan, L.

    2002-01-01

    In the core of BWRs, the second-phase particles (SPP) of Zircaloy-2 and Zircaloy-4, the Zr(FeCr) 2 and the Zr 2 (FeNi) phase, release Fe and dissolve. The degree of dissolution depends on initial size and fluence. These SPP, however, are important for the corrosion behavior of Zircaloy. Zircaloy shows an increase of corrosion at a certain burnup, depending on the initial SPP size and fast neutron fluence. Only Zr alloys with irradiation resistant SPP avoid this type of increased corrosion completely. Two types of irradiation resistant materials were considered. One is a Zr-Sn-Fe alloy containing the Zr 3 Fe phase, which is irradiation resistant under BWR conditions. The other material is a Zr-Sn-Nb alloy containing the irradiation resistant β-Nb phase. In-BWR tests have shown that a Sn content of >0.8% is mandatory to minimize the nodular corrosion. Two prototypes of irradiation resistant alloys, Zr1.3Sn0.25-0.3 Fe and Zr1Sn2-3Nb, were irradiated in a BWR for 1372 days to a fast fluence of 9 x 10 21 n/cm 2 (E > 1 MeV). These irradiation tests showed that Zr1.3Sn0.25-0.3 Fe has a little lower resistance against nodular corrosion than optimized LTP (Low Temperature Process) Zircaloy-2/4 and revealed that Zr1Sn2-3Nb is superior to LTP Zircaloy-2/4 with respect to nodular and shadow corrosion resistance. The BWR corrosion resistance of Zr1Sn2-3Nb depends on heat treatment. The lowest corrosion was observed with material fabricated completely in the α-range, but also material manufactured in the lower (α+β)-range exhibits low corrosion. Material fabricated in the upper (α+β)-range showed a somewhat higher corrosion, a corrosion behavior similar to LTP Zircaloy-2/4. As far as final annealing is concerned, a long time annealing at 540 deg C is superior to a standard recrystallization treatment (e.g., at 580 deg C), which still leads to a corrosion behavior that is better than stress relieved Zr1Sn2-3Nb. Zr1Sn2-3Nb is resistant to shadow corrosion, when fabricated

  8. Irradiation effects on Cr de-mixing in FeCr alloys

    International Nuclear Information System (INIS)

    Tissot, Olivier

    2016-01-01

    Owing to their good thermal properties and excellent swelling resistance, Ferritic-Martensitic (F-M) alloys and ODS steels are potential candidates as structural material and for cladding of future reactors (GEN IV). However, alloys containing more than 10 at.% Cr, which are corrosion resistant, are prone to embrittlement due mainly to α' precipitation. Study of FeCr alloys, model alloys of F-M and ODS steels, is a key point in the understanding of mechanism which are involved by irradiation. The main objective of this study is to identify and quantify the irradiation effects on Cr de-mixing. In a first approach, study of the α - α' decomposition under thermal ageing have been carried out with APT, SANS, and MS. This experiments allow to establish a referent kinetics. an agreement between SANS and APT measurements have been found. Electrons irradiations have been realized between 250 C and 400 C at different doses. α' precipitation have been observed since the first dose (0.012 dpa). The comparison of results with neutron data have shown the efficacy of electron irradiation in α' precipitation. It have also allowed us to determine equilibrium composition of the miscibility gap at 300 C. Ions irradiation with different damage rates (10 -3 and 10 -5 dpa.s -1 ) have been conducted to understand the absence of α' phase reported in literature under this irradiation type. For the first time, APT characterization have revealed α' after ions irradiation at low damage rate. The in depth analyses have shown that injected interstitials strongly reduce α' precipitation. In fact, these interstitials lead to the formation of dislocations loops or could recombine with vacancies and thus reduce the number of vacancy available for diffusion. At higher damage rate (10 -3 dpa.s -1 ), no precipitation have been observed. It has been shown that it could be explain by ballistic dissolution of α' nucleus which are in formation. (author

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

    International Nuclear Information System (INIS)

    Petrenko, P.V.; Grabovskij, Yu.E.; Gritskevich, A.L.; Kulish, N.P.; Mel'nikova, N.A.

    2003-01-01

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

  10. Vacancy enhancement of diffusion after quenching and during irradiation in silver-zinc alloys

    International Nuclear Information System (INIS)

    Schuele, W.

    1980-01-01

    Quenching and annealing experiments were performed on silver-zinc alloys with 8.14 and 30 at %Zn. From the changes of the electrical resistivity due to an increase of the degree of short-range order, the activation energy of self-diffusion was determined to be Qsub(SD) = 1.60 and Qsub(SD) = 1.38 eV for both alloys, respectively. For the migration energy of vacancies, a value Esub(V)sup(M) = 0.64 eV was found for the alloy with 8.14 at %Zn. Evidence is given that the vacancy migration energy Esub(V)sup(M) of the alloys with 30 at %Zn is smaller than 0.60 eV in agreement with data given by Berry and Orehotsky. The results of measurements of radiation-enhanced diffusion obtained by a Russian and a French group, are reinterpreted. It follows that the increase of the degree of order during irradiation is obtained only be vacancy enhancement of diffusion and that the migration activation energy of self-interstitials is Esub(I)sup(M) approximately 0.46 eV and Esub(I)sup(M) approximately 0.41 eV for the alloys with 8.14 and 30 at %Zn, respectively. (author)

  11. Investigation of irradiation strengthening of b.c.c. metals and their alloys. Progress report, January 1976--October 1976

    International Nuclear Information System (INIS)

    1976-01-01

    Research on irradiation of bcc metals and alloys is reported. Data and information are presented in appendixes on low temperature neutron irradiation of Nb, effects of tritium on the yield stress of Nb, multiple dislocation motion, dislocation group motion, dislocation kinetics, and computer simulation of dislocation motion

  12. Damage rates in neutron irradiated FeCo and FeCo2V ordered and disordered alloys

    International Nuclear Information System (INIS)

    Riviere, J.P.; Dinhut, J.F.

    1979-01-01

    Ordered and disordered samples of FeCo and FeCo2V alloys have been irradiated at liquid hydrogen temperature with fission neutrons up to an integrated dose of about 7.2 x 10 17 n/cm 2 (E > 1 MeV). During the irradiation, the resistivity increases continuously due to point defect production. (author)

  13. Influence of irradiation on mechanical properties of Si-Ge alloys

    Energy Technology Data Exchange (ETDEWEB)

    Sichinava, Avtandil; Bokuchava, Guram; Chubinidze, Giorgi; Archuadze, Giorgi [Ilia Vekua Sukhumi Institute of Physics and Technology, Tbilisi (Georgia); Gapishvili, Nodar [Ilia Vekua Sukhumi Institute of Physics and Technology, Tbilisi (Georgia); Georgian Technical University, Tbilisi (Georgia)

    2017-07-15

    Impact of various irradiation (Ar and He ions, high energy electrons) on microhardness and indentation of monocrystalline Si{sub 0,98}Ge{sub 0,02} alloy is studied. Samples of Si and SiGe alloy are obtained by Czochralski (CZ) method in the [111] direction in the atmosphere of high purity Ar. High energy electron irradiation with fluence of ∝10{sup 12} cm{sup -2} is conducted at the Clinac 2100iX. Ar and He ion implantation is performed on modernized ''VEZUVI-3M'' plant. It is shown that for all types of irradiation the microhardness and indentation modulus versus load are characterized by reverse indentation size effect (ISE). With the increase of fluences of Ar and He ions, the maximum value of the effect increases. At high values of loading force impact on the indenter the mechanical characteristics slowly decrease. Impact of isochronous thermal annealing on mechanical properties of high energy electron irradiated samples is studied. Non-monotonic changes of microhardness and indentation modulus are revealed in the temperature range of 200-260 C. It is proposed that such changes are caused by radiation defects transformation. (copyright 2017 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

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

    International Nuclear Information System (INIS)

    Jin, Hyung Ha; Kwon, Sang Chul; Kwon, Jun Hyun

    2011-01-01

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

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

    Science.gov (United States)

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

    2014-02-01

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

  16. Atomic rearrangements in ordered fcc alloys during neutron irradiation

    International Nuclear Information System (INIS)

    Kirk, M.A.; Blewitt, T.H.

    1978-01-01

    Three sets of experiments performed at Argonne National Laboratory over the past few years are described. These experiments deal with atomic rearrangements in the ordered alloys Ni 3 Mn and Cu 3 Au during fast and thermal neutron bombardment. The unique magnetic properties of ordered Ni 3 Mn are utilized to investigate radiation damage production mechanisms at low temperature (5 K) where defect migration is not possible and only disordering is observed. In the case of thermal neutron bombardment, the average recoil energy is about 450 eV and significant disordering due to [110] replacement collision sequences is observed. For fast neutron bombardment where typical recoil energies are 20 keV, significant random disordering is observed but no evidence for sizable replacement sequences is found. The bombardment of ordered Cu 3 Au by fast and thermal neutrons at higher temperature (approx. 150 0 C) is studied by electrical resistance techniques. Both ordering and disordering are observed and related to the number of migrating vacancies escaping from the high energy collision cascade

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

    International Nuclear Information System (INIS)

    Ribis, J.

    2007-12-01

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

  18. Correlative Microscopy of alpha' Precipitation in Neutron-Irradiated Fe-Cr-Al Alloys

    Science.gov (United States)

    Briggs, Samuel A.

    Fe-Cr-Al alloys are currently being considered for accident tolerant light water reactor fuel cladding applications due to their superior high temperature oxidation and corrosion resistance compared to Zr-based alloys. However, precipitation of the Cr-rich alpha' phase during exposure to LWR operational environments can result in application-limiting hardening and embrittlement. To study this effect, four Fe-Cr-Al model alloys with compositions between 10-18 at.% Cr and 5.8-9.3 at.% Al have been neutron-irradiated in the High Flux Isotope Reactor at a target temperature of 320°C to nominal damage doses of up to 7 dpa in order to emulate typical LWR exposure conditions. A correlative microscopy approach involving atom probe tomography, small-angle neutron scattering, and scanning transmission electron microscopy coupled with energy dispersive x-ray spectroscopy was employed to study the resulting precipitate microstructure. This approach necessitated the development of various analysis techniques to allow for cross-comparison between experimental techniques, including a novel method for correcting for trajectory aberration artifacts in atom probe data sets through phase density comparison. Successful correlation of results from these techniques allows for the individual limitations of each to be overcome and enables the detailed microstructural information gleaned from highly localized atom probe tomography analyses to be extrapolated to bulk alloy behavior. Precipitation response was found to increase with Cr content, while Al additions appeared to partially destabilized the alpha' phase, resulting in precipitate compositions with reduced Cr content compared to binary Fe-Cr systems. Observed precipitate evolution with radiation dose indicates a diffusion-limited coarsening mechanism that is similar to what is observed in the thermally aged system. This work represents the current state-of-the-art on both techniques for analysis of alpha' precipitate

  19. Environmental effects on irradiation creep behavior of highly purified V-4Cr-4Ti alloys (NIFS-Heats) irradiated by neutrons

    OpenAIRE

    FUKUMOTO, K; MARUI, M; MATSUI, H; NAGASAKA, T; MUROGA, T; LI, M; HOELZER, D.T.; ZINKLE, S.J.

    2009-01-01

    In order to investigate the effect of the environment on the irradiation creep properties ofhighly purified V-4Cr-4Ti alloys, neutron irradiation experiments with sodium-enclosed irradiationcapsules in Joyo and lithium-enclosed irradiation capsules in HFIR-17J were carried out usingpressurized creep tubes (PCTs).It was found that the creep strain rate exhibited a linear relationship with the effective stressup to 150 MPa at 458°C and 598°C in the Joyo irradiation experiments. For HFIR-17J irr...

  20. Laser-Irradiation-Induced Melting and Reduction Reaction for the Formation of Pt-Based Bimetallic Alloy Particles in Liquids.

    Science.gov (United States)

    Han, Yechuang; Wu, Shouliang; Dai, Enmei; Ye, Yixing; Liu, Jun; Tian, Zhenfei; Cai, Yunyu; Zhu, Xiaoguang; Liang, Changhao

    2017-05-05

    Laser melting in liquids (LML) is one of the most effective methods to prepare bimetallic alloys; however, despite being an ongoing focus of research, the process involved in the formation of such species remains ambiguous. In this paper, we prepared two types of Pt-based bimetallic alloys by LML, including Pt-Au alloys and Pt-iron group metal (iM=Fe/Co/Ni) alloys, and investigated the corresponding mechanisms of alloying process. Detailed component and structural characterizations indicate that laser irradiation induced a quite rapid formation process (not exceeding 10 s) of Pt-Au alloy nanospheres, and the crystalline structures of Pt-Au alloys is determined by the monometallic constituents with higher content. For Pt-iM alloys, we provide direct evidence to support the conclusion that FeO x /CoO x /NiO x colloids can be reduced to elementary Fe/Co/Ni particles by ethanol molecules during laser irradiation, which then react with Pt colloids to form Pt-iM sub-microspheres. These results demonstrate that LML provides an optional route to prepare Pt-based bimetallic alloy particles with tunable size, components, and crystalline phase, which should have promising applications in biological and catalysis studies. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. Probing the electronic structure of Ni–Mn–In–Si based Heusler alloys thin films using magneto-optical spectra in martensitic and austenitic phases

    Energy Technology Data Exchange (ETDEWEB)

    Novikov, A. [Department of Physics, Lomonosov Moscow State University, Moscow 119991 (Russian Federation); Sokolov, A., E-mail: asokol@unlserve.unl.edu [Department of Physics and Astronomy, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Gan’shina, E.A. [Department of Physics, Lomonosov Moscow State University, Moscow 119991 (Russian Federation); Quetz, Abdiel; Dubenko, I.S. [Department of Physics, Southern Illinois University, Carbondale, IL 62901 (United States); Stadler, S. [Department of Physics and Astronomy, Louisiana State University, Baton Rouge, LA 70803 (United States); Ali, N. [Department of Physics, Southern Illinois University, Carbondale, IL 62901 (United States); Titov, I.S.; Rodionov, I.D. [Department of Physics, Lomonosov Moscow State University, Moscow 119991 (Russian Federation); Lähderanta, E. [Lappeenranta University of Technology, 53851 (Finland); Zhukov, A. [Dpto. de Física de Materiales, Fac. Químicas, UPV/EHU, 20018 San Sebastian (Spain); IKERBASQUE, Basque Foundation for Science, 48011 Bilbao (Spain); Granovsky, A.B. [Department of Physics, Lomonosov Moscow State University, Moscow 119991 (Russian Federation); Sabirianov, R. [Department of Physics, University of Nebraska at Omaha, Omaha, NE 68182 (United States)

    2017-06-15

    Highlights: • Magneto-optical properties of NiMnIn thin films with a magnetostructural transition. • Comparative analysis of magnetic properties in martensitic and austenite phases. • DFT calculations of the MO Kerr effect and site-resolved DOS agree with experiment. • The electronic structure does not change significantly with Martensitic transition. - Abstract: Thin films of Ni{sub 52}Mn{sub 35−x}In{sub 11+x}Si{sub 2} were fabricated by magnetron sputtering on MgO (0 0 1) single crystal substrates. Magnetization as function of temperature for Ni{sub 52}Mn{sub 35}In{sub 11}Si{sub 2} exhibits features consistent with a magnetostructural transition (MST) from an austenitic phase to a martensitic phase, similar to the bulk material. We observed that the martensitic transformation is externally sensitive to small changes in chemical composition and stoichiometry. It has been found that thin films of Ni{sub 52}Mn{sub 34−x}In{sub 11+x}Si{sub 2} with x = 0 and 1 undergo a temperature-induced MST or remain in a stable austenitic phase, respectively. Comparison of magneto-optical transverse Kerr effect spectra obtained at 0.5–4.0 eV in the 35–300 K temperature interval reveal insignificant differences between the martensitic and austenite phases. We found that the field and temperature dependencies of the transverse Kerr effect are quite different from the magnetization behavior, which is attributed to magnetic inhomogeneity across the films. To elucidate the effects of magnetostructural phase transitions on the electronic properties, we performed density functional calculations of the magneto-optical Kerr effect.

  2. The influence of microstructure on blistering and bubble formation by He ion irradiation in Al alloys

    International Nuclear Information System (INIS)

    Soria, S.R.; Tolley, A.; Sánchez, E.A.

    2015-01-01

    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. - Highlights: • In Al and Al–4Cu, He bubbles were formed, but no bubbles were observed in Al-5.6Cu-0.5Si-0.5Ge. • Bubble formation was enhanced at incoherent matrix/precipitate interfaces in Al–4Cu. • The bubble size was insensitive to displacement rate in pure Al. • In Al and Al-5.6Cu-0.5Si-0.5Ge blistering was observed, which was more severe in the alloy. • Blistering effects were enhanced by increasing the displacement rate in Al and Al–4Cu.

  3. Microstructure and tensile properties of heavily irradiated 5052-0 aluminum alloy

    International Nuclear Information System (INIS)

    Farrell, K.

    1980-01-01

    During neutron irradiation of an aluminum 2.2% magnesium solid solution alloy in the High Flux Isotope Reactor to fast and thermal fluences > 10 27 neutrons (n)m 2 at 328 0 K (0.35 T/sub m/) about seven percent insoluble, transmutant silicon was produced. Some of this silicon reacted with the dissolved magnesium to form a fine precipitate of Mg 2 Si. A tight dislocation structure was also created. The alloy showed good resistance to cavity formation. Tension tests at 323, 373, and 423 0 K (0.35, 0.40, and 0.45 T/sub m/) showed pronounced irradiation-induced strengthening and an associated marked loss in ductility. These changes were greater than in magnesium-free aluminum and in alloys containing preexisting, thermally-aged Mg 2 Si precipitate. Increasing the thermal-to-fast flux ratio from 1.7 to 2.1 caused further strengthening beyond that expected from a simple increase in silicon level

  4. Experimental investigations on thermo mechanical behaviour of aluminium alloys subjected to tensile loading and laser irradiation

    Science.gov (United States)

    Jelani, Mohsan; Li, Zewen; Shen, Zhonghua; Sardar, Maryam; Tabassum, Aasma

    2017-05-01

    The present work reports the investigation of the thermal and mechanical behaviour of aluminium alloys under the combined action of tensile loading and laser irradiations. The two types of aluminium alloys (Al-1060 and Al-6061) are used for the experiments. The continuous wave Ytterbium fibre laser (wavelength 1080 nm) was employed as irradiation source, while tensile loading was provided by tensile testing machine. The effects of various pre-loading and laser power densities on the failure time, temperature distribution and on deformation behaviour of aluminium alloys are analysed. The experimental results represents the significant reduction in failure time and temperature for higher laser powers and for high load values, which implies that preloading may contribute a significant role in the failure of the material at elevated temperature. The reason and characterization of material failure by tensile and laser loading are explored in detail. A comparative behaviour of under tested materials is also investigated. This work suggests that, studies considering only combined loading are not enough to fully understand the mechanical behaviour of under tested materials. For complete characterization, one must consider the effect of heating as well as loading rate.

  5. Green and Facile Synthesis of Pd-Pt Alloy Nanoparticles by Laser Irradiation of Aqueous Solution.

    Science.gov (United States)

    Nakamura, Takahiro; Sato, Shunichi

    2015-01-01

    Solid-solution palladium-platinum (Pd-Pt) alloy nanoparticles (NPs) with fully tunable compositions were directly fabricated through high-intensity laser irradiation of an aqueous solution of palladium and platinum ions without using any reducing agents or thermal processes. Transmission electron microscopy (TEM) observations showed that nanometer-sized particles were fabricated by laser irradiation of mixed aqueous solutions of palladium and platinum ions with different feeding ratios. The crystalline nature of the NPs was precisely characterized by X-ray diffraction (XRD). Despite the fact that, for the bulk systems, a pair of XRD peak was detected between the palladium and platinum peaks because of the large miscibility gap in the Pd-Pt binary phase diagram, only a single XRD peak was seen for the Pd-Pt NPs fabricated in the present study. Moreover, the peak position shifted from that of pure palladium towards platinum with increasing fraction of platinum ions in solution. Consequently, the interplanar spacings of the alloy NPs agreed well with the estimated values obtained from Vegard's law. These observations strongly indicate the formation of solid-solution Pd-Pt alloy NPs with fully tunable compositions. This technique is not only a "green" (environmentally-friendly) and facile process, but is also widely applicable to other binary and ternary systems.

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

    International Nuclear Information System (INIS)

    Drouet, Julie

    2014-01-01

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

  7. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  8. Adaptation of fuel code for light water reactor with austenitic steel rod cladding

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira; Giovedi, Claudia

    2015-01-01

    Light water reactors were used with steel as nuclear fuel cladding from 1960 to 1980. The high performance proved that the use of low-carbon alloys could substitute the current zirconium alloys. Stainless steel is an alternative that can be used as cladding. The zirconium alloys replaced the steel. However, significant experiences in-pile occurred, in commercial units such as Haddam Neck, Indian Point, and Yankee experiences. Stainless Steel Types 347 and 348 can be used as cladding. An advantage of using Stainless Steel was evident in Fukushima when a large number of hydrogens was produced at high temperatures. The steel cladding does not eliminate the problem of accumulating free hydrogen, which can lead to a risk of explosion. In a boiling water reactor, environments easily exist for the attack of intergranular corrosion. The Stainless Steel alloys, Types 321, 347, and 348, are stabilized against attack by the addition of titanium, niobium, or tantalum. The steel Type 348 is composed of niobium, tantalum, and cobalt. Titanium preserves type 321, and niobium additions stabilize type 347. In recent years, research has increased on studying the effects of irradiation by fast neutrons. The impact of radiation includes changes in flow rate limits, deformation, and ductility. The irradiation can convert crystalline lattices into an amorphous structure. New proposals are emerging that suggest using a silicon carbide-based fuel rod cladding or iron-chromium-aluminum alloys. These materials can substitute the classic zirconium alloys. Once the steel Type 348 was chosen, the thermal and mechanical properties were coded in a library of functions. The fuel performance codes contain all features. A comparative analysis of the steel and zirconium alloys was made. The results demonstrate that the austenitic steel alloys are the viable candidates for substituting the zirconium alloys. (author)

  9. On the ductile-to-brittle transition behavior of martensitic alloys neutron irradiated to 26 dpa

    International Nuclear Information System (INIS)

    Hu, W.L.; Gelles, D.S.

    1987-01-01

    Charpy impact tests were conducted on specimens made of HT-9 and 9Cr-1Mo in various heat treatment conditions which were irradiated in EBR-II to 26 dpa at 390 to 500 0 C. The results are compared with previous results on specimens irradiated to 13 dpa. HT-9 base metal irradiated at low temperatures showed a small additional increase in ductile brittle transition temperature and a decrease in upper shelf energy from 13 to 26 dpa. No fluence effect was observed in 9Cr-1Mo base metal. The 9Cr-1Mo weldment showed degraded DBTT but improved USE response compared to base metal, contrary to previous findings on HT-9. Significant differences were observed in HT-9 base metal between mill annealed material and normalized and tempered material. The highest DBTT for HT-9 alloys was 50 0 C higher than for the worst case in 9Cr-1Mo alloys. Fractography and hardness measurements were also obtained. Significant differences in fracture appearance were observed in different product forms, although no dependence on fluence was observed. Failure was controlled by the preirradiation microstructure

  10. The effect of irradiation and sputtering on the near-surface composition of dilute alloys

    International Nuclear Information System (INIS)

    Marwick, A.D.; Piller, R.C.

    1978-07-01

    A dilute nickel alloy has been irradiated with 75 keV Ni + ions at temperatures between -8.5 0 C and 550 0 C. Redistribution of solute atoms (Al, Mn, Ti, Cr) has been observed at all temperatures, and is ascribed to the action of point defect fluxes in inducing corresponding fluxes of solute atoms. The solute depth profiles were measured by simultaneous sputtering and SIMS. At temperatures above 350 0 C solute atoms migrate into a peak of concentration at 200 A depth, and are depleted at the surface. At lower temperatures, solute atoms migrate out of the damage region, and are enriched near the surface. The effects of these changes on the sputtering of the dilute components of the alloy are discussed. (author)

  11. Modification of the grain boundary microstructure of the austenitic PCA stainless steel to improve helium embrittlement resistance

    International Nuclear Information System (INIS)

    Maziasz, P.J.; Braski, D.N.

    1986-01-01

    Grain boundary MC precipitation was produced by a modified thermal-mechanical pretreatment in 25% cold worked (CW) austenitic prime candidate alloy (PCA) stainless steel prior to HFIR irradiation. Postirradiation tensile results and fracture analysis showed that the modified material (B3) resisted helium embrittlement better than either solution annealed (SA) or 25% CW PCA irradiated at 500 to 600 0 C to approx.21 dpa and 1370 at. ppM He. PCA SA and 25% CW were not embrittled at 300 to 400 0 C. Grain boundary MC survives in PCA-B3 during HFIR irradiation at 500 0 C but dissolves at 600 0 C; it does not form in either SA or 25% CW PCA during similar irradiation. The grain boundary MC appears to play an important role in the helium embrittlement resistance of PCA-B3

  12. Mechanical Properties and Microstructure of Neutron Irradiated Cold-worked Al-1050 and Al-6063 Alloys

    International Nuclear Information System (INIS)

    Munitz, A.; Cotler, A; Talianker, M.

    1998-01-01

    The impact of neutron irradiation on the internal microstructure, mechanical properties and fracture morphology of cold-worked Al-1050 and Al-6063 alloys was studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens consisting of 50 mm long and 6 mm wide gauge sections, were punched out from Al-1050 and Al-6063 23% cold-worked tubes. They were exposed to prolonged neutron irradiation of up to 4.5x10 25 and 8x10 25 thermal neutrons/m 2 (E -3 s -1 . In general, the uniform and total elongation, the yield stress, and the ultimate tensile strength increase as functions of fluence. However, for Al-1050 a decrease in the ultimate tensile strength and yield stress was observed up to a fluence of 1x10 25 thermal neutrons/m 2 which then increase with thermal neutrons fluence. Metallographic examination and fractography for Al-6063 revealed a decrease in the local area reduction of the final fracture necking. This reduction is accompanied by a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. In contrast, for Al-1050, fracture morphology remains ductile transgranular shear rupture and the final local area reduction remains almost constant No voids could be observed in either alloy up to the maximum fluence. The dislocation density of cold-worked Al was found to decrease with the thermal neutron fluence. Prolonged annealing of unirradiated cold-worked Al-6063 at 52 degree led to similar results. Thus, it appears that, under our irradiation conditions, whereby the temperature encompassing the samples increases the exposure to this thermal field is the major factor influencing the mechanical properties and microstructure of aluminum alloys

  13. Evolution Law of Helium Bubbles in Hastelloy N Alloy on Post-Irradiation Annealing Conditions

    Directory of Open Access Journals (Sweden)

    Jie Gao

    2016-10-01

    Full Text Available This work reports on the evolution law of helium bubbles in Hastelloy N alloy on post-irradiation annealing conditions. After helium ion irradiation at room temperature and subsequent annealing at 600 °C (1 h, the transmission electron microscopy (TEM micrograph indicates the presence of helium bubbles with size of 2 nm in the depth range of 0–300 nm. As for the sample further annealed at 850 °C (5 h, on one hand, a “Denuded Zone” (0–38 nm with rare helium bubbles forms due to the decreased helium concentration. On the other hand, the “Ripening Zone” (38–108 nm and “Coalescence Zone” (108–350 nm with huge differences in size and separation of helium bubbles, caused by different coarsening rates, are observed. The mechanisms of “Ostwald ripening” and “migration and coalescence”, experimentally proved in this work, may explain these observations.

  14. Neutron irradiation effects on grain-refined W and W-alloys

    International Nuclear Information System (INIS)

    Hasegawa, A.; Fukuda, M.; Tanno, T.; Nogami, S.; Yabuuchi, K.; Tanaka, T.; Muroga, T.

    2014-10-01

    Microstructural data of neutron irradiated Tungsten (W) such as size and number density of voids and precipitates obtained by W up to 1.5dpa irradiation in the temperature range of 400-800degC were compiled quantitatively. Nucleation and growth process of these defects were clarified and a qualitative prediction of the damage structure development and hardening of W in fusion reactor environments were made taking into account the solid transmutation effects for the first time. To improve recrystallization behavior and low temperature embrittlement, grain refined-W alloys were fabricated by K- or La-doping method. Rhenium addition to the grain refining process was also examined to improve mechanical properties. Characterizations of unirradiated materials were performed. (author)

  15. Effects of fluence and fluence rate of proton irradiation upon magnetism in Fe65Ni35 Invar alloy

    Science.gov (United States)

    Matsushita, Masafumi; Wada, Hideki; Matsushima, Yasushi

    2015-11-01

    Curie temperature, TC, of the Fe-Ni Invar alloys increase due to irradiation with electron and some kinds of ions. In this study, proton irradiation effects upon magnetism in an Fe65Ni35 alloy have been investigated. It is found that the increment of TC, ∆TC, increases with increasing fluence. The magnetic hysteresis curve of the alloy was found to be unaffected by irradiation. Comparing ∆TC and the calculated energy transfer from the ions to the sample, it seemed that ∆TC was found to be related to the number of vacancies formed in nuclear collision events. In addition, ∆TC was influenced by the fluence rate, i.e., the deposited energy per unit time.

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

    Energy Technology Data Exchange (ETDEWEB)

    Korchuganova, Olesya A., E-mail: KorchuganovaOA@gmail.com [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 115409, Moscow (Russian Federation); State Scientific Center of the Russian Federation, Institute for Theoretical and Experimental Physics of National Research Centre “Kurchatov Institute”, 117218, Moscow (Russian Federation); Thuvander, Mattias [Chalmers University of Technology, SE-412 96, Göteborg (Sweden); Aleev, Andrey A.; Rogozhkin, Sergey V. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 115409, Moscow (Russian Federation); State Scientific Center of the Russian Federation, Institute for Theoretical and Experimental Physics of National Research Centre “Kurchatov Institute”, 117218, Moscow (Russian Federation); Boll, Torben [Chalmers University of Technology, SE-412 96, Göteborg (Sweden); Kulevoy, Timur V. [State Scientific Center of the Russian Federation, Institute for Theoretical and Experimental Physics of National Research Centre “Kurchatov Institute”, 117218, Moscow (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 115409, Moscow (Russian Federation)

    2016-08-15

    Nanostructure evolution during ion irradiation of two thermally aged binary Fee22Cr alloys has been investigated using atom probe tomography. Specimens aged at 500 °C for 50 and 200 h were irradiated by 5.6 MeV Fe ions at room temperature up to fluences of 0.3 × 10{sup 15} ions/cm{sup 2} and 1 × 10{sup 15} ions/cm{sup 2}. The effect of irradiation on the material nanostructure was examined at a depth of 1 μm from the irradiated surface. The analysis of Cr radial concentration functions reveals that dense α′-phase precipitates in the 200 h aged alloy become diffuse and thereby larger when subjected to irradiation. On the other hand, less Cr-enriched precipitates in the alloy aged for 50 h are less affected. The CreCr pair correlation function analysis shows that matrix inhomogeneity decreases under irradiation. Irradiation leads to a decrease in the number density of diffuse clusters, whereas in the case of well-developed precipitates it remains unchanged.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    Nanostructure evolution during ion irradiation of two thermally aged binary Fee22Cr alloys has been investigated using atom probe tomography. Specimens aged at 500 °C for 50 and 200 h were irradiated by 5.6 MeV Fe ions at room temperature up to fluences of 0.3 × 10 15 ions/cm 2 and 1 × 10 15 ions/cm 2 . The effect of irradiation on the material nanostructure was examined at a depth of 1 μm from the irradiated surface. The analysis of Cr radial concentration functions reveals that dense α′-phase precipitates in the 200 h aged alloy become diffuse and thereby larger when subjected to irradiation. On the other hand, less Cr-enriched precipitates in the alloy aged for 50 h are less affected. The CreCr pair correlation function analysis shows that matrix inhomogeneity decreases under irradiation. Irradiation leads to a decrease in the number density of diffuse clusters, whereas in the case of well-developed precipitates it remains unchanged.

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

    Science.gov (United States)

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

    2016-08-01

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

  19. Evaluation of irradiation effects of near-infrared free-electron-laser of silver alloy for dental application.

    Science.gov (United States)

    Kuwada-Kusunose, Takao; Kusunose, Alisa; Wakami, Masanobu; Takebayashi, Chikako; Goto, Haruhiko; Aida, Masahiro; Sakai, Takeshi; Nakao, Keisuke; Nogami, Kyoko; Inagaki, Manabu; Hayakawa, Ken; Suzuki, Kunihiro; Sakae, Toshiro

    2017-08-01

    In the application of lasers in dentistry, there is a delicate balance between the benefits gained from laser treatment and the heat-related damage arising from laser irradiation. Hence, it is necessary to understand the different processes associated with the irradiation of lasers on dental materials. To obtain insight for the development of a safe and general-purpose laser for dentistry, the present study examines the physical effects associated with the irradiation of a near-infrared free-electron laser (FEL) on the surface of a commonly used silver dental alloy. The irradiation experiments using a 2900-nm FEL confirmed the formation of a pit in the dental alloy. The pit was formed with one macro-pulse of FEL irradiation, therefore, suggesting the possibility of efficient material processing with an FEL. Additionally, there was only a slight increase in the silver alloy temperature (less than 0.9 °C) despite the long duration of FEL irradiation, thus inferring that fixed prostheses in the oral cavity can be processed by FEL without thermal damage to the surrounding tissue. These results indicate that dental hard tissues and dental materials in the oral cavity can be safely and efficiently processed by the irradiation of a laser, which has the high repetition rate of a femtosecond laser pulse with a wavelength around 2900 nm.

  20. Effects of cavitation on damage calculations in ion-irradiated P7 alloy

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Farrens, S.N.; Kulcinski, G.L.

    1985-01-01

    The purpose of this study is to investigate the effect of voids on the depth-dependent damage energy in ion-irradiated metals. Corrections to the dose at the swelling peak will be used to obtain the swelling rate of ion-irradiated 316-type stainless steels. Samples of the P7 alloy were ion-irradiated to four fluence levels up to a peak dose level of 100 dpa at 650 0 C. The depth-dependent void parameters extracted in cross section were used to model the effect of voids on the depth-dependent damage produced during 14 MeV nickel ion irradiation. An increase in the range of damage produced from the original foil surface for the target containing voids was modeled as a first-order correction to the damage profile. A second-order effect, void straggling, was shown to cause a time-dependent decrease in the damage rate at the peak swelling depth. Corrections applied to the dose at the peak swelling depth yield swelling rates approaching 0.7%/dpa

  1. Tensile and electrical properties of copper alloys irradiated in a fission reactor

    International Nuclear Information System (INIS)

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.; Rowcliffe, A.F.

    1996-01-01

    Postirradiation electrical sensitivity and tensile measurements have been completed on pure copper and copper alloy sheet tensile specimens irradiated in the SM-2 reactor to doses of ∼0.5 to 5 dpa and temperatures between ∼80 and 400 degrees C. Considerable radiation hardening and accompanying embrittlement was observed in all of the specimens at irradiation temperature below 200 degrees C. The radiation-induced electrical conductivity degradation consisted of two main components: solid transmutation effects and radiation damage (defect cluster and particle dissolution) effects. The radiation damage component was nearly constant for the doses in this study, with a value of ∼1.2nΩm for pure copper and ∼1.6nΩm for dispersion strengthened copper irradiated at ∼100 degrees C. The solid transmutation component was proportional to the thermal neutron flux, and became larger than the radiation damage component for fluences larger than ∼5 10 24 n.m 2 . The radiation hardening and electrical conductivity degradation decreased with increasing irradiation temperature, and became negligible for temperatures above ∼300 degrees C

  2. Influence of irradiation with energy-rich particles on the hardness of the Fe-Cr alloy

    International Nuclear Information System (INIS)

    Heintze, Cornelia

    2013-01-01

    Ferritic/martensitic and oxide dispersion strengthened ferritic/martensitic steels are candidate structural materials for components exposed to high neutron fluxes in future nuclear applications like fusion and generation IV fission reactors. The ductilebrittle transition and its shift to higher temperatures which is predominantly caused by irradiation hardening are main concerns for these materials. In the present work, the irradiation behaviour of binary Fe-Cr model alloys, which represent a simplified model for ferritic/martensitic steels, is studied. To this end irradiation with iron ions is used in order to simulate the neutron-induced damage. Due to the limited penetration depth characterization methods suitable for thin layers have to be applied. In the present case, nanohardness testing and transmission electron microscopy (TEM) are employed. The results, including the irradiation-induced hardness change of the layer as a function of chromium content, fluence and irradiation temperature and, for selected cases, quantitative TEM analyses, were exploited to identify irradiation-induced dislocation loops as one source of irradiation hardening. Additional results of small-angle neutron scattering experiments on neutron-irradiated specimens of the same alloys show that nm-scaled α'-phase precipitates also significantly contribute to the irradiation-induced hardness increase. An Orowan model is used to estimate the obstacle strengths posed to dislocation glide by these lattice defects. The topic is stepwise extended to more complex situations with respect to the irradiation conditions and the materials. Considering simultaneous and sequential irradiations with iron- and helium-ions it is shown that the effect of helium on irradiation hardening depends on the chronological order of the irradiations and that the simultaneous introduction of helium in fusion-relevant concentrations amplifies irradiation hardening based on a synergistic effect. There is no

  3. On the position of local levels of defects in proton-irradiated Pb1-xSnxTe alloys

    International Nuclear Information System (INIS)

    Brandt, N.B.; Gas'kov, A.M.; Ladygin, E.A.; Skipetrov, E.P.; Khorosh, A.G.

    1989-01-01

    Effect of fast proton irradiation (T≅300 K, E=200 keV, F≤2x10 14 cm -2 ) on electrophysical properties of thin layers p-Pb 1-x Sn x Te (0.17 ≤x≤ 0.26) is investigated. Saturation of radiation flux dependences of hole density due to occurrence of a resonance level under irradiation, which is near the ceiling of the valence band of alloys, and due to stabilization of the Fermi level with the resonance level is detected. Possibility of coordination of novadays data on the position of the levels of radiation defects in alloys Pb 1-x Sn x Te is discussed

  4. Evaluation of austenitic stainless steels for transpassive corrosion by metal purification technology. Synergistic effect of Si and P on intergranular corrosion of Fe-18Cr-14Ni alloys

    International Nuclear Information System (INIS)

    Mayuzumi, Masami; Ohta, Joji; Kako, Kenji; Kawakami, Eishi

    2001-01-01

    The synergistic effect of Si, Mn, C, P, and S on the transpassive corrosion of HP18Cr-14Ni alloys was studied in 13N nitric acid. The specimens were fabricated using a cold crucible method in a high-vacuum chamber to reduce contamination. The additions of Si<1% and Mn<2% had no effect on the corrosion behavior of HP18Cr-14Ni alloys, and the addition of Si<1% also had no effect on the corrosion behavior of HP18Cr-14Ni-1Mn alloys, although 1% Si induced intergranular corrosion in both the alloys. Thus, HP18Cr-14Ni-1Mn-0.5Si alloys were selected to evaluate the effects of C, P and S (100 ppm each). The addition of P, and the co-addition of C, P, and S to HP18Cr-14Ni-1Mn-0.5Si induced intergranular corrosion of the same degree in the solution annealed condition. This result suggests the synergistic effect of Si and P to induce intergranular corrosion, since the single addition of Si or P to this level did not lead to intergranular corrosion of HP18Cr-14Ni alloys. HP18Cr-14Ni-1Mn-0.5Si alloys containing C, P, and S at the 100 ppm level each showed superior corrosion resistance compared to a commercial Type 304L in 13N nitric acid. (author)

  5. Final report on characterization of physical and mechanical properties of copper and copper alloys before and after irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Tähtinen, S.

    2002-01-01

    The present report summarizes and highlights the main results of the work carried out during the last 5-6 years on effects of neutron irradiation on physical and mechanical properties of copper and copper alloys. The work was an European contribution toITER Research and Development programme...... the suitability of a copper alloy for its use in the first wall and divertor components of ITER. It is pointed out that the present work has managed onlyto identify some of the critical problems and limitations of the copper alloys for their employment in the hostile environment of 14 MeV neutrons. A considerable...

  6. Radiation-Induced α' Phase Formation on Dislocation Loops in Fe-Cr Alloys During Electron Irradiation

    OpenAIRE

    Wakai, E.; Hishinuma, A.; Kato, Y.; Yano, H.; Takaki, S.; Abiko, K.

    1995-01-01

    Radiation-induced precipitates on dislocation loops in low and high purity Fe-9, -18 and -50 % Cr alloys were examined under electron irradiation in a high voltage electron microscope operated at 1 MV. Two types of dislocation loops on {100} planes with a Burgers vectors and on {111} planes with a /2 are formed in high purity Fe-Cr alloys. However, only a type loops are formed in low purity alloys, i.e. where carbon concentration is greater than about 60 wt.ppm. The growth rate of the loops...

  7. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Onimus, F

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  8. Performance assessment of MOX fuel with Alloy D9 cladding and wrapper irradiated in FBTR

    International Nuclear Information System (INIS)

    Joseph, Jojo; Ramachandran, Divakar; Venkiteswaran, C. N.; Karthik, V.; Johny, T.; Rao, B. P. C.; Jayakumar, T.

    2015-01-01

    A test fuel sub-assembly (FSA) with 37 fuel pins consisting of annular MOX fuel pellets encapsulated in Alloy D9 cladding and wrapper simulating the fuel design of the Prototype Fast Breeder Reactor (PFBR) was irradiated to a peak burn-up of 112 GWd/t in the FBTR and subjected to post-irradiation examination (PIE) at the Radiometallurgy Laboratory (RML) of IGCAR. The investigations consisted of non-destructive and destructive tests designed to evaluate the performance of the fuel and structural materials. Moderate fuel swelling and fission gas release of around 85% was observed. Non-destructive tests gave indications of changes in the central hole dimensions of the fuel pellet and Fuel-Clad Chemical Interaction (FCCI) and these were confirmed by metallographic sections. Initiation of FCMI was also revealed at the core top locations. The Alloy D9 cladding and wrapper have performed satisfactorily with respect to swelling resistance and residual mechanical properties up to the maximum displacement damage of about 60 dpa that has been attained in the FBTR. The results give confidence to operate PFBR with the designated fuel design. (author)

  9. Identification of ultra-fine Ti-rich precipitates in V-Cr-Ti alloys irradiated below 300 deg. C by using positron CDB technique

    International Nuclear Information System (INIS)

    Fukumoto, Ken-ichi; Matsui, Hideki; Ohkubo, Hideaki; Tang, Zheng; Nagai, Yasuyoshi; Hasegawa, Masayuki

    2008-01-01

    Irradiation-induced Ti-rich precipitates in V-Ti and V-4Cr-4Ti alloys are studied by TEM and positron annihilation methods (positron lifetime, and coincidence Doppler broadening (CDB)). The characteristics of small defect clusters formed in V alloys containing Ti at irradiation temperatures below 300 deg. C have not been identified by TEM techniques. Strong interaction between vacancy and Ti solute atoms for irradiated V alloys containing Ti at irradiation temperatures from 220 to 350 deg. C are observed by positron lifetime measurement. The vacancy-multi Ti solute complexes in V-alloys containing Ti are definitely identified by using CDB measurement. It is suggested that ultra-fine Ti-rich precipitates or Ti segregation at periphery of dislocation loops are formed in V alloys containing Ti at irradiation temperatures below 300 deg. C

  10. Microstructure and tensile properties of heavily irradiated 5052-0 aluminum alloy

    Science.gov (United States)

    Farrell, K.

    1981-03-01

    During neutron irradiation of an Al-2.2% Mg solid solution alloy in the High Flux Isotope Reactor to fast and thermal fluences μ10 27neutrons( n)/ m2at 328 K (0.35 Tm) about seven percent insoluble, transmutant silicon was produced. Some of this silicon reacted with the dissolved magnesium to form a fine precipitate of Mg 2Si. A tight dislocation structure was also created. The alloy showed good resistance to cavity formation. These microstructural features are responsible for pronounced strengthening and an associated marked loss in ductility as revealed by tensile tests at 323, 373, and 423 K (0.35, 0.40 and 0.45 Tm). These changes were greater than in magnesium-free aluminum and in alloys containing preexisting, thermally-aged Mg 2Si precipitate. Increasing the thermal-to-fast flux ratio from 1.7 to 2.1 caused further strengthening beyond that expected from a simple increase in silicon level.

  11. Microstructural evolutions of zirconium alloys under irradiation. Link with the irradiation growth phenomenon

    International Nuclear Information System (INIS)

    Simonot, C.

    1995-01-01

    This study deals with the irradiation-induced growth and microstructural evolutions of Zircaloy-4 type materials (ZrSn 1.2-1.7 Fe 0.18-0.24 Cr 0.07-0.13 O 0.09-0.15 ), used as cladding and guide-tubes in PWR's fuel assemblies. The main objective was to obtain a better understanding of the growth acceleration which may occur at high doses for the recrystallized metallurgical state. The elongation values of stress-free tubes irradiated at 400 deg in experimental reactors give clear indication of accelerated growth after a critical dose. Microstructural investigations reveal some large basal plane dislocation loops with vacancy character, which is an unexpected defect configuration for an hexagonal material with a c/a ratio less than the ideal value. In addition, a significant redistribution of iron and chromium solute elements comes from the dissolution of the initial Zr(Fe,Cr) 2 phases. In a guide-tube irradiated to high dose at 320 deg in a power reactor, a large density of these c-component loops was also observed in coincidence with a large iron re-solution due to the progressive partial amorphization of Laves phases. By contrast, as long as a negligible amount of iron is available in the matrix (start of progressive amorphization at 350 deg or complete amorphization without any chemical change at 280 deg, only prism plane loops with interstitial and vacancy character are observed and the steady-state growth rate is low. A mechanism taking into account the Diffusional Anisotropy Difference of the radiation induced point defects seems to be the most suitable to explain the correlations between microstructural evolutions and growth rates. However it does not allow to predict the dose necessary for the formation of the basal plane loops responsible for the growth acceleration. Several factors (dose, temperature, metallurgical state) are required and the iron re-solution is likely to play a major role to stabilize such defects, by a decrease in stacking fault energy

  12. Characterization of ion-irradiated ODS Fe–Cr alloys by doppler broadening spectroscopy using a positron beam

    Energy Technology Data Exchange (ETDEWEB)

    Parente, P.; Leguey, T. [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain); Castro, V. de, E-mail: vanessa.decastro@uc3m.es [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain); Gigl, T.; Reiner, M.; Hugenschmidt, C. [FRM II and Physics Department, Technische Universität München, 85747 Garching (Germany); Pareja, R. [Departamento de Física and IAAB, Universidad Carlos III de Madrid, 28911 Leganés (Spain)

    2015-09-15

    The damage profile of oxide dispersion strengthened steels after single-, or simultaneous triple-ion irradiation at different conditions has been characterized using a low energy positron beam in order to provide information on microstructural changes induced by irradiation. Doppler broadening and coincident Doppler broadening measurements of the positron annihilation line have been performed on different Fe–Cr–(W,Ti) alloys reinforced with Y{sub 2}O{sub 3}, to identify the nature and stability of irradiation-induced open-volume defects and their possible association with the oxide nanoparticles. It was found that irradiation induced vacancy clusters are associated with Cr atoms. The results are of highest interest for modeling the damage induced by 14 MeV neutrons in reduced activation Fe–Cr alloys relevant for fusion devices.

  13. Effect of irradiation on corrosion of low-activation austenite Cr-Mn steel in technological liquid mediums of nuclear power plant

    International Nuclear Information System (INIS)

    Demina, E.V.; Prusakova, M.D.; Vinogradova, N.A.; Orlova, G.D.; Nechaev, A.F.; Doil'nitsyn, V.A.

    2008-01-01

    Effect of γ-radiation on corrosion rate in cold-worked and annealed low-activation austenite 12Cr-20Mn steel has been studied. Corrosion tests were carried out in water solutions which simulate the coolant medium in the primary coolant circuit of WWER power reactor and in the circuit of multiple forced circulation of RBMK-1000 reactor as well as an aquatic environment in cooling pond for spent fuel. The worst radiation effect was observed in the cooling pond environment where the value of corrosion rate is increased by tens or hundreds times

  14. The effects of thermal-neutron irradiation on platinum and dilute platinum-gold alloys

    International Nuclear Information System (INIS)

    Piani, C.S.B.

    1978-12-01

    The effect of varying defect concentrations on the recovery spectrum of thermal-neutron-irradiated pure platinum after isochronal anneals was investigated. The dose-independence of substages I(A), I(B) and I(C), and the dose dependence of substage I(D) and I(E), were observed to be in agreement with electron-irradiated studies. The 120 K substage in pure platinum was shown not to be due to interstitial-interstitial reactions, but could possibly be accounted for in terms of detrapping of interstitials from impurities or intrinsic immobile defects. The 360 K stage was shown to shift and was suppressed with increasing defect concentration. The possible conversion of the crowdion to a dumbbell near 160 K in Stage ll in platinum, as predicted by the two-interstitial model, was investigated by consideration of the initial slopes of the production curves between 80 K and 300 K. A minimum in these slopes was observed near 160 K and could be interpreted as due to the conversion of the highly mobile crowdion to an immobile dumbbell at this temperature. The influence of varying gold concentrations on the recovery spectrum of platinum was investigated in dilute platinum-gold alloys. The characteristics of several additional substages in Stage ll, due to the gold alloying were comparable to the results of electron-irradiation experiments. The observations made with regard to the impurity (gold) dependence of these substages could be interpreted in terms of the concentrations of the interstitials, vacancies and impurities present in the material. The interpretation of these substages was found to be consistent, if the recovery spectrum was investigated as a function of defect concentration [af

  15. Pitting corrosion resistant austenite stainless steel

    Science.gov (United States)

    van Rooyen, D.; Bandy, R.

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

  16. Microstructural changes of Y-doped V-4Cr-4Ti alloys after ion and neutron irradiation

    Directory of Open Access Journals (Sweden)

    H. Watanabe

    2016-12-01

    Full Text Available High-purity Y-doped V-4Cr-4Ti alloys (0.1–0.2wt. % Y, manufactured by the National Institute for Fusion Science (NIFS, were used for this study. Heavy-ion and fission-neutron irradiation was carried out at temperatures 673–873K. During the ion irradiation at 873K, the microstructure was controlled by the formation of Ti(C,O,N precipitates lying on the (100 plane. Y addition effectively suppressed the growth of Ti(C,O,N precipitates, especially at lower dose irradiation to up to 4 dpa. However, at higher dose levels (12.0 dpa, the number density was almost at the same levels irrespective of the presence of Y. After neutron irradiation at 873K, fine titanium oxides were also observed in all V alloys. However, smaller oxide sizes were observed in the Y-doped samples under the same irradiation conditions. The detailed analysis of EDS showed that the center of the Ti(C,O,N precipitates was mainly enriched by nitrogen. The results showed that the contribution of not only oxygen atoms picked up from the irradiation environment but also nitrogen atoms is essential to understand the microstructural evolution of V-4Cr-4Ti-Y alloys.

  17. Characterization of ion irradiation effects on the microstructure, hardness, deformation and crack initiation behavior of austenitic stainless steel:Heavy ions vs protons

    Science.gov (United States)

    Gupta, J.; Hure, J.; Tanguy, B.; Laffont, L.; Lafont, M.-C.; Andrieu, E.

    2018-04-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is a complex phenomenon of degradation which can have a significant influence on maintenance time and cost of core internals of a Pressurized Water Reactor (PWR). Hence, it is an issue of concern, especially in the context of lifetime extension of PWRs. Proton irradiation is generally used as a representative alternative of neutron irradiation to improve the current understanding of the mechanisms involved in IASCC. This study assesses the possibility of using heavy ions irradiation to evaluate IASCC mechanisms by comparing the irradiation induced modifications (in microstructure and mechanical properties) and cracking susceptibility of SA 304 L after both type of irradiations: Fe irradiation at 450 °C and proton irradiation at 350 °C. Irradiation-induced defects are characterized and quantified along with nano-hardness measurements, showing a correlation between irradiation hardening and density of Frank loops that is well captured by Orowan's formula. Both irradiations (iron and proton) increase the susceptibility of SA 304 L to intergranular cracking on subjection to Constant Extension Rate Tensile tests (CERT) in simulated nominal PWR primary water environment at 340 °C. For these conditions, cracking susceptibility is found to be quantitatively similar for both irradiations, despite significant differences in hardening and degree of localization.

  18. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends

  19. The effect of cold work, heat treatment, and composition on the austenite to R-phase transformation temperature of Ni-Ti shape memory alloys

    International Nuclear Information System (INIS)

    Thoma, P.E.; Angst, D.R.; Schachner, K.D.

    1995-01-01

    The influence of cold work (CW) and heat treatment (HT) on the austenite to R-Phase (A→R) transformation temperature (TT) of a near equiatomic and three other Ti rich NiTi SMA is examined. For the SMA having a near equiatomic composition, the A→R TT increases with increasing CW at low HT temperatures. For the SMA having the maximum possible Ti content, the A→R TT decreases with increasing CW at low HT temperatures. For all compositions, the A→R TT is not sensitive to CW at high HT temperatures. At a Ti content slightly below the maximum possible, the A→R TT is relatively insensitive to CW and HT temperature. For all of the SMA investigated, the A→R TT increases with increasing Ti content for a specific CW and HT temperature, and this effect is greatest at low CW and at high HT temperatures. (orig.)

  20. Metastable phase formation in ion-irradiated nickel-aluminum alloys

    International Nuclear Information System (INIS)

    Eridon, J.M.

    1986-01-01

    Phase transformations induced by ion beam mixing of nickel-aluminum alloys with 500-keV krypton ions were investigated over a range of temperatures (80 K to 300K), composition (NiAl 3 , NiAl, Ni 1 Al), initial structures (both nickel-aluminum layers and ordered intermetallic compounds), and doses (ranging from 2 x 10 14 cm -2 to 5 x 10 16 cm -2 ). Samples were formed by alternate evaporation of layers of nickel and aluminum in high vacuum onto copper grids. These samples were check for purity with energy dispersive-x-ray spectroscopy, electron energy-loss spectroscopy, and Rutherford backscattering spectrometry. A portion of these samples was annealed to form the intermetallic compounds appropriate to the given composition. Irradiations were performed at both room temperature (300 K) and 80 K using the 2-MV ion accelerator at Argonne National Laboratory. Phase transformations were observed during both in-situ irradiations in the High Voltage Electron Microscopy at Argonne and also in subsequent electron-diffraction analysis of an array of samples irradiated in a target chamber. Metastable phases formed include disordered crystalline structures at composition s of 25% and 50% aluminum, an amorphous structure at 75% aluminum, and a hexagonal closed-packed structure formed at 25% aluminum. These metastable states were all converted to the stable intermetallic compounds through annealing treatments

  1. Prediction model for oxide thickness on aluminum alloy cladding during irradiation

    International Nuclear Information System (INIS)

    Kim, Yeon Soo; Hofman, G.L.; Hanan, N.A.; Snelgrove, J.L.

    2003-01-01

    An empirical model predicting the oxide film thickness on aluminum alloy cladding during irradiation has been developed as a function of irradiation time, temperature, heat flux, pH, and coolant flow rate. The existing models in the literature are neither consistent among themselves nor fit the measured data very well. They also lack versatility for various reactor situations such as a pH other than 5, high coolant flow rates, and fuel life longer than ∼1200 hrs. Particularly, they were not intended for use in irradiation situations. The newly developed model is applicable to these in-reactor situations as well as ex-reactor tests, and has a more accurate prediction capability. The new model demonstrated with consistent predictions to the measured data of UMUS and SIMONE fuel tests performed in the HFR, Petten, tests results from the ORR, and IRIS tests from the OSIRIS and to the data from the out-of-pile tests available in the literature as well. (author)

  2. Effect of irradiation temperature on crystallization of {alpha}-Fe induced by He irradiations in Fe{sub 80}B{sub 20} amorphous alloy

    Energy Technology Data Exchange (ETDEWEB)

    San-noo, Toshimasa; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Hayashi, Nobuyuki; Sakamoto, Isao

    1997-03-01

    Since amorphous alloys are generally highly resistant to irradiation and their critical radiation dose is an order of magnitude higher for Fe-B amorphous alloy than Mo-methods, these alloys are expected to become applicable as for fusion reactor materials. The authors investigated {alpha}-Fe crystallization in an amorphous alloy, Fe{sub 80}B{sub 20} using internal conversion electron Moessbauer spectroscopy. The amount of {alpha}-Fe component was found to increase by raising the He-irradiation dose. The target part was modified to enable He ion radiation at a lower temperature (below 400 K) by cooling with Peltier element. Fe{sub 80}B{sub 20} amorphous alloy was cooled to keep the temperature at 300 K and exposed to 40 keV He ion at 1-3 x 10{sup 8} ions/cm{sup 2}. The amount of {alpha}-Fe crystal in each sample was determined. The crystal formation was not observed for He ion radiation below 2 x 10{sup 18} ions/cm{sup 2}, but that at 3 x 10{sup 8} ions/ cm{sup 2} produced a new phase ({delta} +0.40 mm/sec, {Delta} = 0.89 mm/sec). The decrease in the radiation temperature from 430 to 300 K resulted to extremely repress the production of {alpha}-Fe crystal, suggesting that the crystallization induced by He-radiation cascade is highly depending on the radiation temperature. (M.N.)

  3. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kohnert, Aaron A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dasgupta, Dwaipayan [Univ. of Tennessee, Knoxville, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-23

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

  4. In-situ synchrotron diffraction and digital image correlation technique for characterizations of retained austenite stability in low-alloyed transformation induced plasticity steel

    International Nuclear Information System (INIS)

    Brauser, S.; Kromm, A.; Kannengiesser, Th.; Rethmeier, M.

    2010-01-01

    Direct measurement and quantification of phase transformation in a low-alloyed transformation induced plasticity steels depending on the tensile load as well as determination of the real true stress and true strain values were carried out in-situ using high energy synchrotron radiation. Digital image correlation technique was used to quantify more precisely the true strain values. The aim of the work was to obtain a better understanding of the phase transformation of commercial low-alloyed transformation induced plasticity steel depending on the true strain and true stress values.

  5. Effect of irradiation in a spallation neutron environment on tensile properties and microstructure of aluminum alloys 5052 and 6061

    Energy Technology Data Exchange (ETDEWEB)

    Dunlap, J.A.; Stubbins, J.F. [Univ. of Illinois, Urbana, IL (United States); Borden, M.J.; Sommer, W.F. [Los Alamos National Lab., NM (United States)

    1996-12-31

    The Accelerator Production of Tritium (APT) and the Accelerator Transmutation of Waste (ATW) programs require structural materials which retain good mechanical properties when exposed in a spallation neutron irradiation environment. One group of materials likely to withstand the environment anticipated for these systems is the aluminum alloy series. To characterize this class of materials in a prototypical irradiation environment, AL5052 (Al-2.7Mg) and Al6061 (Al-1.1Mg-0.5Si) in hardened and annealed conditions were irradiated to a fluence of 4.2 {times} 10{sup 20} neutrons/cm{sup 2} at {approximately} 100 C in a spallation neutron source. Following irradiation, tensile tests and post-test examinations were performed to determine the influence of irradiation and test temperature on mechanical properties and fracture mode. It was found that, the properties of these two aluminum alloys were not significantly affected by the irradiation exposure conditions examined here. Thus these materials may be acceptable as structural materials for APT and ATW applications. This conclusion is based on limited mechanical properties testing, supported by other information in the literature on the performance of these materials in other irradiation environments.

  6. Effects of neutron irradiation at 4500C and 16 dpa on the properties of various commercial copper alloys

    International Nuclear Information System (INIS)

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

    1985-01-01

    High-purity copper and eight copper alloys were irradiated to approx.16 dpa at approx.450 0 C in the MOTA experiment in FFTF. These alloys were also examined after aging at 400 0 C for 1000 hours. The radiation-induced changes in the electrical conductivity, tensile properties, and density were measured and compared to those of the aged materials. The changes in conductivity can be either positive or negative depending on the alloy. Changes in tensile properties of most, but not all, of the alloys seem to be primarily dependent on thermal effects rather than the effect of atomic displacements. Radiation at 450 0 C induced changes in density varying from 0.66% densification to 16.6% swelling. The latter occurred in Cu-O.1% Ag and implies a swelling rate of at least 1%/dpa. 6 references, 3 figures, 2 tables

  7. High-temperature strength of Nb-1%Zr alloy for irradiation-capsules inner-shell

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Nakata, Hirokatsu; Tanaka, Mitsuo; Fukaya, Kiyoshi.

    1978-04-01

    Coated fuel particles in capsules will be irradiated at about 1600 0 C in JMTR. Nb-1%Zr alloy was chosen for inner shell material of the capsules because of its sufficient strength at 1000 0 C and low induced radioactivity. Nb-1%Zr ingot produced by electron beam melting was formed into seamless tubes by hollowing and swaging, followed by annealing. Creep test in helium flow and tensile test in vacuum were made to examine mechanical strength of the Nb-1%Zr tubes at 1000 0 C. Following are the results; 1) 0.2% yield stress at 1000 0 C is about 6 kg/mm 2 . 2) 3000 hr creep rupture stress at 1000 0 C is about 6 kg/mm 2 . (auth.)

  8. Corrosion of zirconium alloys in nuclear reactors: A model for irradiation induced enhancement by local radiolysis in the porous oxide

    International Nuclear Information System (INIS)

    Lemaignan, C.; Salot, R.

    1997-01-01

    An analysis has been undertaken of the various cases of local enhancement of corrosion rate of zirconium alloys under irradiation. It is observed that in most cases a strong emission of energetic β - is present leading to a local energy deposition rate higher than the core average. This suggests that the local transient radiolytic oxidizing species produced in the coolant by the β - particles could contribute to corrosion enhancement, by increasing the local corrosion potential. This process is applicable to the local enhanced corrosion found in front of stainless steels structural parts, due to the contribution of Mn, and in front of Pt inserts or Cu-rich cruds. It explains also the irradiation corrosion enhancement of Cu-Zr alloys. Enhanced corrosion around neutron absorbing material is explained similarly by pair production from conversion of high energy capture photons in the cladding, leading to energetic electrons. The same process was found to be active with other highly ionizing species like α from Ni-rich alloys and fission products in homogeneous reactors. Due to the changes induced by the irradiation intensity on the concentration of the radiolytic species, the coolant chemistry, that controls the boundary conditions for oxide growth, has to be analyzed with respect to the local value of the energy deposition rate. An analysis has been undertaken which shows that, in a porous media, the water is exposed to a higher intensity than bulk water. This leads to a higher concentration of oxidizing radiolytic species at the root of the cracks of the porous oxide, and increases the corrosion rate under irradiation. This mechanism, deduced from the explanation proposed for localized irradiation enhanced corrosion, can be extended to the whole reactor core, where the general enhancement of Zr alloys corrosion under irradiation could be attributed to the general radiolysis in the porous zirconia. (author). 18 refs, 3 figs, 3 tabs

  9. Phase instability of alloys caused by transmutation effects during neutron irradiation

    International Nuclear Information System (INIS)

    Platov, Yu.M.; Pletnev, M.N.

    1994-01-01

    A theory of the phase changes in a two-phase binary A-B alloy in the coarsening condition caused by burnout of solute B due to nuclear reactions is presented. It is shown that this burnout process introduces diffusion redistribution of solute between second phase precipitates and solid solution. The burnout induced solute flux away from second phase precipitates to solid solution maintaining the concentration of element B in the vicinity to its solubility limit and stimulates, thus, the second phase particle dissolution. This occurs in addition to a process decreasing their sizes as a result of direct burnout of atoms B in the precipitates. In the framework of the theory developed here, analytical expressions describing time evolution of the precipitate size distributions, changes of mean radius and number density of the precipitates, and second phase dissolution times are obtained. On the basis of these results and numerical calculations for aluminium-scandium alloy, it is shown that the burnout processes can induce essential phase changes, and thus cause significant changes of the properties of irradiated materials at high neutron fluences. ((orig.))

  10. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hee Joon; Edwards, Dan J.; Kurtz, Richard J.; Yamamoto, Takuya; Wu, Yuan; Odette, G. Robert

    2017-02-01

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 59Ni(nth, 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  11. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    Science.gov (United States)

    Jung, Hee Joon; Edwards, Dan J.; Kurtz, Richard J.; Yamamoto, Takuya; Wu, Yuan; Odette, G. Robert

    2017-02-01

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 58Ni(nth,γ) 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

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

    Energy Technology Data Exchange (ETDEWEB)

    Koegler, R., E-mail: r.koegler@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Bautzner Landstrasse 400, 01328 Dresden (Germany); Anwand, W. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Bautzner Landstrasse 400, 01328 Dresden (Germany); Richter, A. [Department of Engineering, Technical University of Applied Sciences Wildau, Bahnhofstrasse 1, 15745 Wildau (Germany); Butterling, M.; Ou, Xin; Wagner, A. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Bautzner Landstrasse 400, 01328 Dresden (Germany); Chen, C.-L. [Department of Materials Science and Engineering, I-Shou University, Kaohsiung 840, Taiwan (China)

    2012-08-15

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

  13. Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

    2011-01-01

    A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

  14. Mechanical properties testing of several 800 MeV proton irradiated BCC metals and alloys

    International Nuclear Information System (INIS)

    Brown, R.D.; Wechsler, M.S.; Tschalaer, C.

    1986-01-01

    A spallation neutron source for the 600-MeV proton accelerator facility at the Swiss Institute for Nuclear Research (SIN) consists of a vertical cylinder filled with molten Pb-Bi. The proton beam enters the cylinder, passing upward through a window in contact with the Pb-Bi eutectic liquid. Investigations are underway at the 800-MeV proton accelerator at LAMPF to test the performance of candidate SIN window materials. Based on considerations of chemical compatibility with molten Pb-Bi, as well as radiation damage mechanisms, Fe, Ta, Fe-2.25Cr-1Mo, and Fe-12Cr-1Mo (Ht-9) were chosen as candidate materials. Sheet tensile samples were sealed inside capsules containing Pb-Bi and were proton-irradiated at LAMPF to two fluences, 4.8 and 54 x 10 23 p/m 2 . The beam current was approximately equal to the 1 mA anticipated for the upgraded SIN accelerator. Yield and ultimate strengths increased upon irradiation in all materials, while the ductility decreased. The pure metals, Ta and Fe, exhibited the greatest radiation hardening and embrittlement. The HT-9 alloy showed the smallest changes in strength and ductility

  15. Evaluation of a Belt-Cast Austenitic Steel Alloy from Salzgitter Mannesmann Forschung: Effect of Hardness on the Ballistic Resistance against Two 0.30-cal. Projectile Types

    Science.gov (United States)

    2017-08-01

    manufacturing technology and requisite alloy compositions for the production of high-strength and high-ductility security steels. Security steels...company was still experiencing fabrication issues that prevent the material from being fabricated in thicker sections. Therefore, currently, only 5-mm... production , an extended period of negotiations was required to obtain representative samples of this experimental steel from Salzgitter. As a result, 5

  16. Enhancement in anomalous Hall resistivity of Co/Pd multilayer and CoPd alloy by Ga+ ion irradiation

    KAUST Repository

    Guo, Zaibing

    2014-02-01

    In this paper, we report the effect of Ga+ ion irradiation on anomalous Hall effect (AHE) and longitudinal resistivity (ρxx) in [Co(3 Å)/Pd(5 Å)]80 multilayer and Co 42Pd58 alloy. 4- and 2-fold increases in anomalous Hall resistivity (ρAH) in the Co/Pd multilayer and CoPd alloy have been observed after irradiations at doses of 2.4 × 1015 and 3.3×10 15 ions/cm2, respectively. Skew scattering and side jump contributions to AHE have been analyzed based on the scaling relationship ρAH = aρxx + bρ2xx. For the Co/Pd multilayer, AHE is mainly affected by ion irradiation-induced interface diffusion and defects. For the CoPd alloy, the increase in doses above 1.5 × 1015 ions/cm2 induces a sign change in skew scattering, followed by the skew scattering contribution to AHE overwhelming the side jump contribution, this phenomenon should be attributed to irradiation-induced defects and modifications in chemical ordering. © Copyright EPLA, 2014.

  17. Electron microscopy of some irradiated solids

    International Nuclear Information System (INIS)

    Housseau, N.

    1985-01-01

    Are studied defects induced by ''in situ'' irradiation in a High Tension Electron Microscope (1MV) in various materials such as: metals and usual alloys for nuclear reactors, low dimensional conductors, organic conductors. The migration energy for interstitial in stainless steels is 0.8 eV for a synthetic austenitic steel and 2.0 eV for an industrial titanium steel. Irradiation damage of low dimensional compounds, TaS 2 and TaS 3 are studied as a function of temperature and irradiation dose. Modifications of surstructure diffraction spots and the progressive disappearance of ordered phases of Charge Density Wave are observed. Electron diffraction pattern of crystalline organic compouds very sensitive to irradiation are obtained at 7K. Classification according to decreaseing sensitivity is Qsub(n)-(TCNQ) 2 → TM-TSF-DM-TCNQ → TTF-TCNQ → TTT 2 I 3 [fr

  18. Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 degrees C

    DEFF Research Database (Denmark)

    Singh, B.N.; Edwards, D.J.; Eldrup, Morten Mostgaard

    2001-01-01

    Screening experiments were carried out to determine the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile...... results are described and their salient features discussed. The most significant effect of neutron irradiation is a severe loss of ductility in the case of CuNiBe alloys. (C) 2001 Elsevier Science B.V. All rights reserved....

  19. Hydrogen induced ductility losses in austenitic stainless steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, J.A.; West, A.J.

    1978-06-01

    The effect of hydrogen on the tensile behavior of austenitic stainless steel welds was studied in two AISI 300 series alloys and two nitrogen strengthened alloys. The microstructure of these welds typically contained several percent ferrite in an austenite matrix. Hydrogen was found to reduce the ductility of all welds; however, the severity of ductility loss decreased with increasing stacking fault energy, as observed in previous studies on wrought material. In the lowest stacking fault energy welds, 304L and 308L, hydrogen changed the fracture mode from simple rupture to a mixed mode of ductile and brittle fracture associated with the austenite ferrite interface. Higher stacking fault energy welds, 309S and 22-13-5, showed smaller losses in ductility. In these materials hydrogen assisted the ductile rupture process by aiding void growth and coalescence, without changing the fracture mode. Varying the amount of ferrite from approximately one to 10 percent had no significant effect on performance in hydrogen.

  20. Evaluation of ferritic alloy Fe-2 1/4Cr-1Mo after neutron irradiation: Microstructural development

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1986-10-01

    As part of a program to provide a data base on the bainitic alloy Fe-2-1/4-1Mo for fusion energy applications, microstructural examinations are reported for nine specimen conditions for 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510 0 C. Void swelling is found following irradiation at 400 0 C to 480 0 C. Concurrently dislocation structure and precipitation developed. Peak void swelling, void density, dislocation density and precipitate number density formed at the lowest temperature, approximately 400 0 C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior

  1. A study on the influence of trace elements (C, S, B, Al, N) on the hot ductility of the high purity austenitic alloy Fe-Ni 36% (INVAR)

    Energy Technology Data Exchange (ETDEWEB)

    Simonetta-Perrot, M.T.

    1994-11-01

    In order to study the damage mechanisms leading to the ductility decrease of the Invar alloy at 600 C, a high-purity Fe-Ni 36% sample has been doped with trace elements with the purpose of identifying the role of sulfur, sulfur with Al N or B N precipitates and sulfur with boron, on the ductility, the failure modes, the intergranular damage and the plastic deformation mechanisms prior to failure. A new AES segregation quantification method has been used to study the kinetics and thermodynamics of intergranular and surface segregations and determine the relation between sulfur segregation and grain joint fragility. refs., figs., tabs.

  2. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  3. Bactericidal and antimicrobial effects of pure titanium and titanium alloy treated with short-term, low-energy UV irradiation.

    Science.gov (United States)

    Itabashi, T; Narita, K; Ono, A; Wada, K; Tanaka, T; Kumagai, G; Yamauchi, R; Nakane, A; Ishibashi, Y

    2017-02-01

    The surface of pure titanium (Ti) shows decreased histocompatibility over time; this phenomenon is known as biological ageing. UV irradiation enables the reversal of biological ageing through photofunctionalisation, a physicochemical alteration of the titanium surface. Ti implants are sterilised by UV irradiation in dental surgery. However, orthopaedic biomaterials are usually composed of the alloy Ti6Al4V, for which the antibacterial effects of UV irradiation are unconfirmed. Here we evaluated the bactericidal and antimicrobial effects of treating Ti and Ti6Al4V with UV irradiation of a lower and briefer dose than previously reported, for applications in implant surgery. Ti and Ti6Al4V disks were prepared. To evaluate the bactericidal effect of UV irradiation, Staphylococcus aureus 834 suspension was seeded onto the disks, which were then exposed to UV light for 15 minutes at a dose of 9 J/cm 2 . To evaluate the antimicrobial activity of UV irradiation, bacterial suspensions were seeded onto the disks 0, 0.5, one, six, 24 and 48 hours, and three and seven days after UV irradiation as described above. In both experiments, the bacteria were then harvested, cultured, and the number of colonies were counted. No colonies were observed when UV irradiation was performed after the bacteria were added to the disks. When the bacteria were seeded after UV irradiation, the amount of surviving bacteria on the Ti and Ti6Al4V disks decreased at 0 hours and then gradually increased. However, the antimicrobial activity was maintained for seven days after UV irradiation. Antimicrobial activity was induced for seven days after UV irradiation on both types of disk. Irradiated Ti6Al4V and Ti had similar antimicrobial properties. Cite this article: T. Itabashi, K. Narita, A. Ono, K. Wada, T. Tanaka, G. Kumagai, R. Yamauchi, A. Nakane, Y. Ishibashi. Bactericidal and antimicrobial effects of pure titanium and titanium alloy treated with short-term, low-energy UV irradiation. Bone Joint

  4. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    International Nuclear Information System (INIS)

    Ernst, Frank

    2016-01-01

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute - carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress-corrosion cracking (hydrogen-induced embrittlement), and - potentially - radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non-treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

  5. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, Frank [Case Western Reserve Univ., Cleveland, OH (United States)

    2016-10-13

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

  6. Carbon Concentration of Austenite

    Directory of Open Access Journals (Sweden)

    Z. Ławrynowicz

    2007-07-01

    Full Text Available The investigation was carried out to examine the influence of temperature and times of austempering process on the maximum extend towhich the bainite reaction can proceed and the carbon content in retained austenite. It should be noted that a small percentage change in theaustenite carbon content can have a significant effect on the subsequent austempering reaction changing the volume fraction of the phasespresent and hence, the resulting mechanical properties. Specimens were prepared from an unalloyed ductile cast iron, austenitised at 950oCfor 60 minutes and austempered by the conventional single-step austempering process at four temperatures between BS and MS, eg., 250,300, 350 and 400oC. The samples were austempered at these temperatures for 15, 30, 60, 120 and 240 minutes and finally quenched toambient temperature. Volume fractions of retained austenite and carbon concentration in the residual austenite have been observed byusing X-ray diffraction. Additionally, carbon concentration in the residual austenite was calculated using volume fraction data of austeniteand a model developed by Bhadeshia based on the McLellan and Dunn quasi-chemical thermodynamic model. The comparison ofexperimental data with the T0, T0' and Ae3' phase boundaries suggests the likely mechanism of bainite reaction in cast iron is displacive rather than diffusional. The carbon concentration in retained austenite demonstrates that at the end of bainite reaction the microstructure must consist of not only ausferrite but additionally precipitated carbides.

  7. Characteristics of surface modified Ti-6Al-4V alloy by a series of YAG laser irradiation

    Science.gov (United States)

    Zeng, Xian; Wang, Wenqin; Yamaguchi, Tomiko; Nishio, Kazumasa

    2018-01-01

    In this study, a double-layer Ti (C, N) film was successfully prepared on Ti-6Al-4V alloy by a series of YAG laser irradiation in nitrogen atmosphere, aiming at improving the wear resistance. The effects of laser irradiation pass upon surface chemical composition, microstructures and hardness were investigated. The results showed that the surface chemicals were independent from laser irradiation pass, which the up layer of film was a mixture of TiN and TiC0.3N0.7, and the down layer was nitrogen-rich α-Ti. Both the surface roughness and hardness increased as raising the irradiation passes. However, surface deformation and cracks happened in the case above 3 passes' irradiation. The wear resistance of laser modified sample by 3 passes was improved approximately by 37 times compared to the as received substrate. Moreover, the cytotoxic V ion released from laser modified sample was less than that of as received Ti-6Al-4V alloy in SBF, suggesting the potentiality of a new try to modify the sliding part of Ti-based hard tissue implants in future biomedical application.

  8. Investigation of growth in post electron irradiated Al-11.8 at% Zn alloy by small angle neutron scattering

    International Nuclear Information System (INIS)

    Baig, M.R.

    1996-01-01

    The samples of Al-11.8 at% Zn alloy were irradiated with 2 Mev electrons at a flux of 10 4 electrons/cm 2 -s for different periods of time. The small angle neutron scattering (SANS) experiments were performed on unirradiated and irradiated samples under identical conditions. The results were obtained in the form of differential scattering cross section, versus momentum transfer vector Q(A 0-1 ). In general, results indicate an initial large drop in the magnitude of peak scattering cross section, increase in the precipitate size and reduction in the number density of the precipitates. More significant changes in these parameters have been noticed as the dose is increased. However, no such changes have been observed in the repeated measurements of post irradiated room temperature aged samples. We thus conclude that the changes induced by irradiation are found to be stable, and the small angle neutron scattering results are reproducible. It is thus concluded that there is an absence of growth in post irradiated room temperature aged Al-Zn alloy. (Author)

  9. Effects of bonding bakeout thermal cycles on pre- and post irradiation microstructures, physical, and mechanical properties of copper alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

    At present, dispersion strengthened (DS) copper is being considered as the primary candidate material for the ITER first wall and divertor components. Recently, it was agreed among the ITER parties that a backup alloy should be selected from the two well known precipitation hardened copper alloys, CuCrZr and CuNiBe. It was therefore decided to carry out screening experiments to simulate the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties, and electrical resistivity of CuCrZr and CuNiBe alloys. On the basis of the results of these experiments, one of the two alloys will be selected as a backup material. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime ageing, and bonding thermal cycle followed by reageing and the reactor bakeout treatment at 623K for 100 hours. Tensile specimens of the DS copper were also given the heat treatment corresponding to the bonding thermal cycle. A number of these heat treated specimens of CuCrZr, CuNiBe, and DS copper were neutron irradiated at 523K to a dose level of {approx}0.3 dpa (NRT) in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens with the various heat treatments were tensile tested at 532K. The dislocation, precipitate and void microstructures and electrical resistivity of these specimens were also determined. Results of these investigations will be reported and discussed in terms of thermal and irradiation stability of precipitates and irradiation-induced precipitation and recovery of dislocation microstructure. Results show that the bonding and bakeout thermal cycles are not likely to have any serious deleterious effects on the performance of these alloys. The CuNiBe alloys were found to be susceptible to radiation-induced embrittlement, however, the exact mechanism is not yet known. It is thought that radiation-induced precipitation and segregation of the beryllium may be responsible.

  10. Cyclic deformation behaviour of austenitic steels at ambient and ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Abstract. The aim of the present investigation is to characterise cyclic deforma- tion behaviour and plasticity-induced martensite formation of metastable austenitic stainless steels at ambient and elevated temperatures, taking into account the influ- ence of the alloying elements titanium and niobium. Titanium and niobium are.

  11. Factors which determine the swelling rate of austenitic stainless steels

    International Nuclear Information System (INIS)

    Garner, F.A.; Wolfer, W.G.

    1983-01-01

    Once void nucleation subsides, the swelling rate of many austenitic alloys becomes rather insensitive to variables that control the transient regime of swelling. Models are presented which describe the roles of nickel, chromium and silicon in void nucleation. The relative insensitivity of steady-state swelling to temperature, displacement rate and composition is also discussed

  12. Damage development and hardening in 14 MeV neutron irradiation of copper alloys at 250C

    International Nuclear Information System (INIS)

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

    1981-07-01

    Copper and copper alloyed with five atom percent of either aluminum, nickel or manganese were irradiated at 25 0 with 14 MeV neutrons to fluences up to 7.5 x 10 17 n/cm 2 (0.003 dpa). The radiation-induced microstructure of these materials was characterized by the coupled use of electron microscopy and microhardness. The irradiation-induced microhardness changes were found to be independent of alloy identity and the magnitude of the solute-induced hardening. It appears that at least 70% of the defect clusters are smaller than resolvable by microscopy (approx. 1 nm). The point defects at 25 0 C which survive recombination and aggregate in either visible or invisible clusters constitute at least 8% of those produced in the cascades

  13. Electron irradiation effect on short-range ordering in Cu-Al and Ag-Al alloys

    International Nuclear Information System (INIS)

    Kulish, N.P.; Mel'nikova, N.A.; Petrenko, P.V.; Ryabishchuk, A.L.; Tatarov, A.A.

    1990-01-01

    Method of X-ray diffuse scattering is used to study short-range order variation in Cu-Al and Ag-Al alloys under radiation effect and the following heat treatment. Irradiation was carried out at -40 deg C by 1.6 MeV electrons, fluence of 5x10 7 cm -2 and 0.5 MeV gamma-rays, the dose being 10 7 pH

  14. The use of staining metallograplic reagents in the optical analysis as contribution to the microscopical study of irradiated austenitic canning materials

    International Nuclear Information System (INIS)

    Heylen-Geladi, M.

    1975-09-01

    Investigation by interference film microscopy is rather difficult to realize in post-irradiation studies because of the lack of suitable facilities. To enable comparison to be made between structural compounds of unirradiated and irradiated samples, tests on staining metallographic reagents were performed and the respective results compared with those obtained by interference film microscopy. Two solutions are particularly indicated to fulfil the condition of reliability for these canning materials; one is based on the reduction of selenic acid to selenium, the other on the decomposition of a metabisulphite in an acid medium. (author)

  15. Effect of austenitization conditions on kinetics of isothermal transformation of austenite of structural steels

    International Nuclear Information System (INIS)

    Konopleva, E.V.; Bayazitov, V.M.; Abramov, O.V.; Kozlova, A.G.

    1987-01-01

    Effect of austenization of kinetics of pearlite and bainite transformations for steels with different carbon content differing by alloying character and degree has been investigated. Austenization temperature increase is shown to leads to retardation of ferrite-pearlite transformation in low- and medium-carbon alloyed steels. Step-like holding in the region of austenite stable state (850, 950 deg) after high-temperature heating (1100 deg C) increases the rate of transformation partially recovering its kinetics and decomposition velocity after low-temperature heating in steels alloyed advantageously with carbide-forming elements (08Kh2G2F, 30Kh3) and does not affect kinetics in the 35Kh, 30KhGSN2A, 45N5 steels. Increase of heating temperature and growth of an austenite grain cause considerable acceleration of bainite transformation, increase of the temperaure of bainite transformation beginning and increase of the transformation amplitude in the 08Kh2G2F, 30Kh3 steels and affect weakly kinetics in steels with mixed alloying (30KhGSN2A) or low-alloy one (35Kh). The bainite transformation rate in the 45N5 steelite does not depend on austenization. The effect of additional acceleration of bainite transformation as a result holding after high-temperature heating in those steels, where activation of transformation occurs with increase of heating temperature

  16. Positron Annihilation Spectroscopy and Small Angle Neutron Scattering Characterization of Nanostructural Features in Irradiated Fe-Cu-Mn Alloys

    International Nuclear Information System (INIS)

    Wirth, B D; Asoka-Kumar, P; Howell, R H; Odette, G R; Sterne, P A

    2001-01-01

    Radiation embrittlement of nuclear reactor pressure vessel steels results from a high number density of nanometer sized Cu-Mn-Ni rich precipitates (CRPs) and sub-nanometer matrix features, thought to be vacancy-solute cluster complexes (VSC). However, questions exist regarding both the composition of the precipitates and the defect character and composition of the matrix features. We present results of positron annihilation spectroscopy (PAS) and small angle neutron scattering (SANS) characterization of irradiated and thermally aged Fe-Cu and Fe-Cu-Mn alloys. These complementary techniques provide insight into the composition and character of both types of nanoscale features. The SANS measurements indicate populations of CRPs and VSCs in both alloys. The CRPs are coarser in the Fe-Cu alloy and the number densities of CRP and VSC increase with the addition of Mn. The PAS involved measuring both the positron lifetimes and the Doppler broadened annihilation spectra in the high momentum region to provide elemental sensitivity at the annihilation site. The spectra in Fe-Cu-Mn specimens thermally aged to peak hardness at 450 C and irradiated at 288 C are nearly identical to elemental Cu. Positron lifetime and spectrum measurements in Fe-Cu specimens irradiated at 288 C clearly show the existence of long lifetime (∼500 ps) open volume defects, which also contain Cu. Thus the SANS and PAS provide a self-consistent picture of nanostructures composed of CRPs and VSCs and tend to discount high Fe concentrations in the CRPs

  17. Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

    Science.gov (United States)

    Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

    2017-11-01

    Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

  18. Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under gamma-rays irradiation

    International Nuclear Information System (INIS)

    Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

    2014-01-01

    Seawater was injected into reactor cores of Units 1, 2, and 3 in the Fukushima Daiichi nuclear power station as an urgent coolant. It is considered that the injected seawater causes corrosion of steels of the reactor pressure vessel and primary containment vessel. To investigate the effects of gamma-rays irradiation on weight loss in carbon steel and low-alloy steel, corrosion tests were performed in diluted seawater at 50°C under gamma-rays irradiation. Specimens were irradiated with dose rates of 4.4 kGy/h and 0.2 kGy/h. To evaluate the effects of hydrazine (N 2 H 4 ) on the reduction of oxygen and hydrogen peroxide, N 2 H 4 was added to the diluted seawater. In the diluted seawater without N 2 H 4 , weight loss in the steels irradiated with 0.2 kGy/h was similar to that in the unirradiated steels, and weight loss in the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those in the unirradiated steels. Weight loss in the steels irradiated in the diluted seawater containing N 2 H 4 was similar to that in the diluted seawater without N 2 H 4 . When N 2 was introduced into the gas phase in the flasks during gamma-rays irradiation, weight loss in the steels decreased. (author)

  19. Shallow-Land Buriable PCA-type austenitic stainless steel for fusion application

    International Nuclear Information System (INIS)

    Zucchetti, M.

    1991-01-01

    Neutron-induced activity in the PCA (Primary Candidate Alloy) austenitic stainless steel is examined, when used for first-wall components in a DEMO fusion reactor. Some low-activity definitions, based on different waste management and disposal concepts, are introduced. Activity in the PCA is so high that any recycling of the irradiated material can be excluded. Disposal of PCA radioactive wastes in Shallow-Land Buriable (SLB) is prevented as well. Mo, Nb and some impurity elements have to be removed or limited, in order to reduce the radioactivity of the PCA. Possible low-activity versions of the PCA are introduced (PCA-la); they meet the requirements for SLB and may also be recycled under certain conditions. (author)

  20. Alloy development for irradiation performance in fusion reactors. Annual report, September 1978-September 1979

    International Nuclear Information System (INIS)

    1979-12-01

    This report is the first annual report of research activities directed toward the development of improved performance alloys for such severe environments as the fusion reactor fist wall. Major project efforts are directed toward definition of alloy performance requirements, alloy design, alloy production and alloy performance evaluation. Rapid solidification from the melt is being used to manipulate alloy microstructure and to produce the desired design properties. Integrated testing and modeling procedures have been developed to minimize testing requirements. Progress during the first project year and future plans are summarized in this annual report

  1. Internal friction of Fe-B alloys neutron irradiated at low temperature

    International Nuclear Information System (INIS)

    Kitajima, Kazunori; Futagami, Koji; Abe, Hironobu; Yoshida, Hiroyuki.

    1975-01-01

    Measurements were made on the internal friction of Fe-B alloys irradiated by neutron at 16 0 K to the dose of 3x10 16 nvt (>1 MeV) and 6x10 17 nvt (thermal). Boron was used to enhance the production of defects by the nuclear transformation B 10 (n,α)Li 7 . Relaxation peaks were found in specimens containing dispersed fine precipitates of NbB 2 in range of B 500--7200 wt ppm and Nb 2000--30000 wt ppm. The most prominent peak is the one with the peak temperature of 169 0 K at the frequency of 264 c/sec. Activation energy determined from the peak shift is 0.28+-0.01 eV, which is nearly equal to that of migration of self-interstitial reported on pure iron. However activation energy of the decay of peaks by annealing is about 0.7 eV. Interpretation was presented that the peak may be attributed to re-orientation of self-interstitials loosely bound to a boron atom. (auth.)

  2. Defects and related phenomena in electron irradiated ordered or disordered Fe-Co and Fe-Co-V alloys

    International Nuclear Information System (INIS)

    Riviere, J.P.; Dinhut, J.F.; Desarmot, G.

    1983-01-01

    Two B 2 type alloys Fe 50 at.%-Co 50 at.% and Fe 49 at.%-Co 49 at.%-V 2 at.% either in the ordered or the disordered state have been irradiated with 2.5 MeV electrons at liquid hydrogen temperature. The recovery of the resistivity damage was studied during subsequent isochronal annealing up to 700 K. The resistivity damage rates for both initially disordered Fe-Co and Fe-Co-V alloys are interpreted in terms of point defect production. The intrinsic resistivities rhosub(F) of Frenkel pairs and the effective recombination volumes V 0 are determined. In the Fe-Co ordered alloy point defect production superimposed with a disordering process can account for the resistivity damage. The effective displacement rate causing disordering is determined, indicating that replacement collisions are the dominant disordering mechanism. A calculation of the average number of replacements along directions per Frenkel pair is proposed. During the recovery of the radiation induced resistivity three main stages are observed in both ordered and disordered alloys. The particular resistivity behavior of the Fe-Co-V alloy complicates the interpretation of production and recovery data. (author)

  3. The effect of the initial microstructure in terms of sink strength on the ion-irradiation-induced hardening of ODS alloys studied by nanoindentation

    Science.gov (United States)

    Duan, Binghuang; Heintze, Cornelia; Bergner, Frank; Ulbricht, Andreas; Akhmadaliev, Shavkat; Oñorbe, Elvira; de Carlan, Yann; Wang, Tieshan

    2017-11-01

    Oxide dispersion strengthened (ODS) Fe-Cr alloys are promising candidates for structural components in nuclear energy production. The small grain size, high dislocation density and the presence of particle matrix interfaces may contribute to the improved irradiation resistance of this class of alloys by providing sinks and/or traps for irradiation-induced point defects. The extent to which these effects impede hardening is still a matter of debate. To address this problem, a set of alloys of different grain size, dislocation density and oxide particle distribution were selected. In this study, three-step Fe-ion irradiation at both 300 °C and 500 °C up to 10 dpa was used to introduce damage in five different materials including three 9Cr-ODS alloys, one 14Cr-ODS alloy and one 14Cr-non-ODS alloy. Electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), small angle neutron scattering (SANS), and nanoindentation testing were applied, the latter before and after irradiation. Significant hardening occurred for all materials and temperatures, but it is distinctly lower in the 14Cr alloys and also tends to be lower at the higher temperature. The possible contribution of Cr-rich α‧-phase particles is addressed. The impact of grain size, dislocation density and particle distribution is demonstrated in terms of an empirical trend between total sink strength and hardening.

  4. The Kinetics of Dislocation Loop Formation in Ferritic Alloys Through the Aggregation of Irradiation Induced Defects

    Science.gov (United States)

    Kohnert, Aaron Anthony

    The mechanical properties of materials are often degraded over time by exposure to irradiation environments, a phenomenon that has hindered the development of multiple nuclear reactor design concepts. Such property changes are the result of microstructural changes induced by the collision of high energy particles with the atoms in a material. The lattice defects generated in these recoil events migrate and interact to form extended damage structures. This study has used theoretical models based on the mean field chemical reaction rate theory to analyze the aggregation of isolated lattice defects into larger microstructural features that are responsible for long term property changes, focusing on the development of black dot damage in ferritic iron based alloys. The purpose of such endeavors is two-fold. Primarily, such models explain and quantify the processes through which these microstructures form. Additionally, models provide insight into the behavior and properties of the point defects and defect clusters which drive general microstructural evolution processes. The modeling effort presented in this work has focused on physical fidelity, drawing from a variety of sources of information to characterize the unobservable defect generation and agglomeration processes that give rise to the observable features reported in experimental data. As such, the models are based not solely on isolated point defect creation, as is the case with many older rate theory approaches, but instead on realistic estimates of the defect cluster population produced in high energy cascade damage events. Experimental assessments of the microstructural changes evident in transmission electron microscopy studies provide a means to measure the efficacy of the kinetic models. Using common assumptions of the mobility of defect clusters generated in cascade damage conditions, an unphysically high density of damage features develops at the temperatures of interest with a temperature dependence

  5. Effects of neutron irradiation on microstructure and deformation behaviour of mono- and polycrystalline molybdenum and its alloys

    DEFF Research Database (Denmark)

    Singh, B.N.; Evans, J.H.; Horsewell, A.

    1998-01-01

    The influence of neutron irradiation on microstructural evolution and mechanical properties of mono- and polycrystalline molybdenum and its alloys has been investigated. Tensile specimens and 3 mm diameter discs of monocrystals of pure molybdenum and Mo-5%Re were irradiated with fission neutrons...... specimens were tensile tested at 295 K. Post-irradiation microstructures were quantitatively characterized using a transmission electron microscope (TEM). Fracture surfaces were examined in a scanning electron microscope (SEM). The results of tensile testing as well as of transmission and scanning...... microscopy are presented and discussed in terms of intracascade clustering of self-interstitial atoms and the role of one-dimensional glide of these clusters in controlling microstructural evolution and the resulting mechanical properties....

  6. Anomalously large deformation of 12Cr18Ni10Ti austenitic steel irradiated to 55 dpa at 310 deg. C in the BN-350 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gusev, M.N. [Institute of Nuclear Physics, Almaty (Kazakhstan)], E-mail: gusev.maxim@inp.kz; Maksimkin, O.P.; Osipov, I.S. [Institute of Nuclear Physics, Almaty (Kazakhstan); Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

    2009-04-30

    Whereas most previous irradiation studies conducted at lower neutron exposures in the range 100-400 deg. C have consistently produced strengthening and strongly reduced ductility in stainless steels, it now appears possible that higher exposures may lead to a reversal in ductility loss for some steels. A new radiation-induced phenomenon has been observed in 12Cr18Ni10Ti stainless steel irradiated to 55 dpa. It involves a 'moving wave of plastic deformation' at 20 deg. C that produces 'anomalously' high values of engineering ductility, especially when compared to deformation occurring at lower neutron exposures. Using the technique of digital optical extensometry the 'true stress {sigma}-true strain {epsilon}' curves were obtained. It was shown that a moving wave of plastic deformation occurs as a result of an increase in the intensity of strain hardening, d{sigma}/d{epsilon}({epsilon}). The increase in strain hardening is thought to arise from an irradiation-induced increase in the propensity of the {gamma} {yields} {alpha} martensitic transformation.

  7. Can gamma irradiation during radiotherapy influence the metal release process for biomedical CoCrMo and 316L alloys?

    Science.gov (United States)

    Wei, Zheng; Edin, Jonathan; Karlsson, Anna Emelie; Petrovic, Katarina; Soroka, Inna L; Odnevall Wallinder, Inger; Hedberg, Yolanda

    2018-02-09

    The extent of metal release from implant materials that are irradiated during radiotherapy may be influenced by irradiation-formed radicals. The influence of gamma irradiation, with a total dose of relevance for radiotherapy (e.g., for cancer treatments) on the extent of metal release from biomedical stainless steel AISI 316L and a cobalt-chromium alloy (CoCrMo) was investigated in physiological relevant solutions (phosphate buffered saline with and without 10 g/L bovine serum albumin) at pH 7.3. Directly after irradiation, the released amounts of metals were significantly higher for irradiated CoCrMo as compared to nonirradiated CoCrMo, resulting in an increased surface passivation (enhanced passive conditions) that hindered further release. A similar effect was observed for 316L showing lower nickel release after 1 h of initially irradiated samples as compared to nonirradiated samples. However, the effect of irradiation (total dose of 16.5 Gy) on metal release and surface oxide composition and thickness was generally small. Most metals were released initially (within seconds) upon immersion from CoCrMo but not from 316L. Albumin induced an increased amount of released metals from AISI 316L but not from CoCrMo. Albumin was not found to aggregate to any greater extent either upon gamma irradiation or in the presence of trace metal ions, as determined using different light scattering techniques. Further studies should elucidate the effect of repeated friction and fractionated low irradiation doses on the short- and long term metal release process of biomedical materials. © 2018 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 2018. © 2018 The Authors Journal of Biomedical Materials Research Part B: Applied Biomaterials Published by Wiley Periodicals, Inc.

  8. Radiation enhanced copper clustering processes in Fe-Cu alloys during electron and ion irradiations as measured by electrical resistivity

    International Nuclear Information System (INIS)

    Ishino, S.; Chimi, Y.; Bagiyono; Tobita, T.; Ishikawa, N.; Suzuki, M.; Iwase, A.

    2003-01-01

    To study the mechanism of radiation-enhanced clustering of copper atoms in Fe-Cu alloys, in situ electrical resistivity measurements are performed during irradiation with 100 MeV carbon ions and with 2 MeV electrons at 300 K. Two kinds of highly pure Fe-Cu alloys with Cu content of 0.02 and 0.6 wt% are used. The results are summarized as follows: - Although there is a steep initial resistivity increase below about 10 μdpa, the resistivity steadily decreases after this initial transient in Fe-0.6wt%Cu alloy, while in Fe-0.02wt%Cu alloy, the resistivity either decreases slowly or stays almost constant. The rate of change in resistivity depends on copper concentration. - The rate of change in resistivity per dpa is larger for electron irradiation than for ion irradiation. - Change in dose rate from 10 -8 to 10 -9 dpa/s slightly enhances the rate of resistivity change per dpa. The decrease in resistivity with dose is considered to be due to clustering or precipitation of copper atoms. The initial abrupt increase in resistivity is too large to be accounted for by initial introduction of point defects before copper clustering. Tentatively the phenomenon is explained as due to the formation of embryos of copper precipitates with a large strain field around them. Quantitative evaluation of the results using resistivity contribution of a unit concentration of Frenkel pairs and that of copper atoms gives an important conclusion that more than one copper atom are removed from solid solution by one Frenkel pair. The clustering efficiency is surprisingly high in the present case compared with the ordinary radiation-induced or radiation-enhanced precipitation processes

  9. Study of the interactions between irradiation and chemical order effects in ternary alloys Ni-Cr-Fe; Etude des interactions entre effets d`irradiation et effets d`ordre chimique dans les alliages ternaires Ni-Cr-Fe

    Energy Technology Data Exchange (ETDEWEB)

    Frely, E

    1997-12-31

    Because of its resistance to corrosion even under stress, the alloy 69 (nickel-based alloy with a chemically disordered F.c.c. structure) is a promising material for application in some of the inner parts of nuclear reactor. However, the eventual formation of an ordered NI{sub 2}Cr superstructure under irradiation or thermal ageing might diminish its performances. We have studied the binary model alloy Ni-Cr33at.% as well as the ternary alloys Ni-Cr3at.%-Fe5cat.% and Ni-Cr32at.%-Fe10at.%, the last one having a chemical composition similar to that of the industrial alloy. After irradiation experiments with 2.5 MeV electrons in the 300-500 deg C temperature range, all the model alloys show the Ni{sub 2}Cr superstructure. The samples irradiated at fluences between 2 and 8. 10 d.p.a. have been characterized by X-ray and neutron diffraction. The superlattice reflexions and the ordered domains have been observed by electron microscopy. The critical temperature of the order-disorder transformation, measured under 1 MeV electron irradiation, decreases linearly with iron content. The evolution of the chemical corder has been traced by means of in situ resistivity measurements. We have used the pair exchange based Dienes model of ordering kinetics for studying the long range order S (S between 0.5 and 0.8 after irradiation). The iron seems to remain in disorder in the sublattices. The similarity of the results under thermal ageing and under irradiation shows that the main effect of the electronic irradiation is to accelerate ordering. Under both treatments increasing the iron content or the dislocation density reduce the ordering kinetics. However, this effect is not sufficient to explain the lack of order in alloy 690 after a fluence of 1 d.p.a. (author). 95 refs.

  10. Potential for using rapid solidification for improved irradiation performance in the fusion environment

    International Nuclear Information System (INIS)

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

    1981-01-01

    Rapid solidification has a potential for improving perfomance in the fusion environment (first-wall materials, limiters, superconductors, ...) through structural refinements of crystalline materials and the preparation of amorphous materials with selected compositions. Compaction techniques which are used for rapidly solidified particulates allow as well for preparation of graded or layered materials. The following topics are being studied by using rapid solidification under a current M.I.T. alloy devlopment program: swelling resistance and high temperature strength of austenitic stainless steels; DBTT and high-temperature strength of ferritic steels; high strength copper alloys; simulation techniques (boron, lithium doping); irradiation damage in metallic glasses at high fluences

  11. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Singh, B.N; Edwards, D.J.; Bilde-Sørensen, Jørgen

    2004-01-01

    The phenomenon of plastic flow localization in the form of "cleared" channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels areinitiated during post-irradiation deformation...... has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons.Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron...... at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these results are discussedwith specific reference to the origin and consequences of plastic flow localization....

  12. Oxidation resistant high creep strength austenitic stainless steel

    Science.gov (United States)

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

    2010-06-29

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

  13. Effects of metallurgical variables on swelling of modified 316 and higher Ni austenitic stainless steels

    International Nuclear Information System (INIS)

    Shibahara, Itaru; Akasaka, Naoaki; Onose, Shoji

    1996-01-01

    The effects of solute elements and cold-work on swelling in modified 316 and higher Ni advanced austenitic stainless steels developed for FBR core material were investigated together with the posted model alloys. The Si, P, B, Ti, Nb modified and cold-worked steels exhibited an improved swelling resistance. In the temperature range between 400 and 500 C, the swelling was suppressed significantly by an addition of 0.8 wt% Si. The beneficial effect of Si appears to be reduced in the steels without Ti and Nb tending to form γ' precipitates. In the temperature range between 500 and 600 C, a needle-like phosphide precipitates played an important role in suppressing void growth. Additions of Ti and/or Nb were found to stabilize the phosphide phase and extended the swelling incubation period. In the improved austenitic steels, the synergistic effect of cold-working and P, B, Ti, Nb additions act beneficially to stabilize the dislocation structure and to form finely dispersed precipitates during irradiation

  14. Effect of He+ fluence on surface morphology and ion-irradiation induced defect evolution in 7075 aluminum alloys

    Science.gov (United States)

    Ni, Kai; Ma, Qian; Wan, Hao; Yang, Bin; Ge, Junjie; Zhang, Lingyu; Si, Naichao

    2018-02-01

    The evolution of microstructure for 7075 aluminum alloys with 50 Kev helium ions irradiation were studied by using optical microscopy (OM), scanning electron microscopy (SEM), x-ray diffraction (XRD) and transmission electron microscopy (TEM). The fluences of 1 × 1015, 1 × 1016 and 1 × 1017 ions cm-2 were selected, and irradiation experiments were conducted at room temperatures. The transmission process of He+ ions was simulated by using SRIM software, including distribution of ion ranges, energy losses and atomic displacements. Experimental results show that irradiated pits and micro-cracks were observed on irradiation sample surface, and the size of constituent particles (not including Mg2Si) decreased with the increasing dose. The x-ray diffraction results of the pair of peaks is better resolved in irradiated samples might indicate that the stressed structure consequence due to crystal defects (vacancies and interstitials) after He+ implantation. TEM observation indicated that the density of MgZn2 phase was significantly reduced after helium ion irradiation which is harmful to strength. Besides, the development of compressive stress produced a large amount of dislocation defects in the 1015 ions cm-2 sample. Moreover, higher fluence irradiation produced more dislocations in sample. At fluence of 1016 ions cm-2, dislocation wall formed by dislocation slip and aggregation in the interior of grains, leading to the refinement of these grains. As fluence increased to 1017 ions cm-2, dislocation loops were observed in pinned dislocation. Moreover, dislocation as effective defect sink, irradiation-induced vacancy defects aggregated to these sinks, and resulted in the formation of helium bubbles in dislocation.

  15. Modeling of the Recrystallization and Austenite Formation Overlapping in Cold-Rolled Dual-Phase Steels During Intercritical Treatments

    Science.gov (United States)

    Ollat, M.; Massardier, V.; Fabregue, D.; Buscarlet, E.; Keovilay, F.; Perez, M.

    2017-10-01

    Austenite formation kinetics of a DP1000 steel was investigated from a ferrite-pearlite microstructure (either fully recrystallized or cold-rolled) during typical industrial annealing cycles by means of dilatometry and optical microscopy after interrupted heat treatments. A marked acceleration of the kinetics was found when deformed ferrite grains were present in the microstructure just before austenite formation. After having described the austenite formation kinetics without recrystallization and the recrystallization kinetics of the steel without austenite formation by simple JMAK laws, a mixture law was used to analyze the kinetics of the cold-rolled steel for which austenite formation and recrystallization may occur simultaneously. In the case where the interaction between these two phenomena is strong, three main points were highlighted: (i) the heating rate greatly influences the austenite formation kinetics, as it affects the degree of recrystallization at the austenite start temperature; (ii) recrystallization inhibition above a critical austenite fraction accelerates the austenite formation kinetics; (iii) the austenite fractions obtained after a 1 hour holding deviate from the local equilibrium fractions given by Thermo-Calc, contrary to the case of the recrystallized steel. This latter result could be due to the fact that the dislocations of the deformed ferrite matrix could promote the diffusion of the alloying elements of the steel and accelerate austenite formation.

  16. Synthesis of Au and Au/Cu alloy nanoparticles on multiwalled carbon nanotubes by using microwave irradiation

    International Nuclear Information System (INIS)

    Rangari, Vijaya K.; Dey, Sanchita; Jeelani, Shaik

    2010-01-01

    Gold nanoparticles and gold-copper alloy nanoparticles were synthesized by reduction of chloroauric acid (HAuCl 4 .xH 2 O) and co-reduction of chloroauric acid (HAuCl 4 .xH 2 O) and Copper(II) acetate [(CH 3 COO) 2 Cu.H 2 O] by ethylene glycol through microwave irradiation technique. In this reaction ethylene glycol used as a solvent and also reducing agent. The cetyltrimethyl ammonium bromide (CTAB) used as surfactant. Au nanoparticles and Au-Cu nanoparticles on the surface of multiwalled carbon nanotube also produced by using same procedure. The XRD analysis confirmed the formation of Au and Au-Cu alloy nanoparticles on multiwalled carbon nanotubes(CNTs). The morphology and size of the particles were examined by the transmission electron microscopy. The EDS analysis on individual particles confirmed that the presence of two metals in a particle in case of alloy nanoparticle. The results presented here show that a variety of well defined metal and metal alloy nanoparticles can be produced by using the microwave polyol process with in a short period of time. (author)

  17. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe–Cr model alloys

    International Nuclear Information System (INIS)

    Bergner, F.; Pareige, C.; Hernández-Mayoral, M.; Malerba, L.; Heintze, C.

    2014-01-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe–Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α′-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys

  18. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe–Cr model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany); Pareige, C. [Groupe de Physique des Matériaux, Université et INSA de Rouen, UMR 6634 CNRS, Avenue de l’Université, BP 12, 76801 Saint Etienne du Rouvray (France); Hernández-Mayoral, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Malerba, L. [SCK-CEN, Nuclear Material Science Institute, Boeretang 200, B-2400 Mol (Belgium); Heintze, C. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2014-05-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe–Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α′-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

  19. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe-Cr model alloys

    Science.gov (United States)

    Bergner, F.; Pareige, C.; Hernández-Mayoral, M.; Malerba, L.; Heintze, C.

    2014-05-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe-Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α‧-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

  20. Impact of Ion Irradiation upon Structure and Magnetic Properties of NANOPERM-Type Amorphous and Nanocrystalline Alloys

    Directory of Open Access Journals (Sweden)

    Marcel Miglierini

    2015-01-01

    Full Text Available Structural modifications and their impact upon magnetic properties are studied in amorphous and nanocrystalline NANOPERM-type 57Fe75Mo8Cu1B16 alloy. They are introduced by irradiation with 130 keV N+ ions to the total fluencies of up to 2.5 × 1017 ions/cm2 under different cooling conditions. Increased temperature during the irradiation triggers formation of nanocrystallites of bcc-Fe in those subsurface regions that are affected by bombarding ions. No crystallization occurs when good thermal contact between the irradiated sample and a sample holder is assured. Instead, structural rearrangement which favours development of magnetically active regions was determined by the local probe methods of Mössbauer spectrometry. Dipole magnetic interactions dominate in subsurface regions on that side of the ribbons which was exposed to ion irradiation. Nevertheless, structural modifications demonstrate themselves also via macroscopic magnetic parameters such as temperature dependence of magnetization, Curie temperature, and hysteresis loops. Impact of only the temperature itself to the observed effects is assessed by the help of samples that were subjected just to heat treatment, that is, without ion irradiation.

  1. Defects in hyperpure Fe-based alloys created by 3 MeV e{sup -}-irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, X.H.; Moser, P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M.; Van Duysen, C. [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    Information about vacancy defects created in RPV (Reactor Pressure Vessels) steels after neutron irradiations are obtained via a simulation: the RPV steels are simulated by a series of high purity Fe-based alloys; the neutron irradiation is simulated by a 3 MeV electron irradiation; vacancy defects characteristics are obtained by positron lifetime techniques. Irradiations are made at 150 or 288 deg C, with a dose of 4*10{sup 19} e-/cm{sup 2}, and followed by isochronal annealing in the range 20-500 deg C. The observed vacancy defects are single trapped vacancies and small vacancy clusters, the size of which being lower than 10 empty atomic volumes (vacancy clusters containing more than 50 empty atomic volumes were never found). A large recovery step is observed between 200 and 400 deg C, after 150 deg C irradiation and attributed to vacancy-impurity detrapping, and also, vacancy cluster evaporation. The influence of C, Cu and Mo are presented. These results are in agreement with a model supposing, in pure Fe, single vacancy migration at -50 deg C and vacancy-impurity detrapping at 200 deg C. (authors). 4 figs., 15 refs.

  2. 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; steel alloys is also disclosed.

  3. Review of corrosion phenomena on zirconium alloys, niobium, titanium, inconel, stainless steel, and nickel plate under irradiation

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1975-01-01

    The role of nuclear fluxes in corrosion processes was investigated in ATR, ETR, PRTR, and in Hanford production reactors. Major effort was directed to zirconium alloy corrosion parameter studies. Corrosion and hydriding results are reported as a function of oxygen concentration in the coolant, flux level, alloy composition, surface pretreatment, and metallurgical condition. Localized corrosion and hydriding at sites of bonding to dissimilar metals are described. Corrosion behavior on specimens transferred from oxygenated to low-oxygen coolants in ETR and ATR experiments is compared. Mechanism studies suggest that a depression in the corrosion of the Zr--2.5Nb alloy under irradiation is due to radiation-induced aging. The radiation-induced onset of transition on several alloys is in general a gradual process which nucleates locally, causing areas of oxide prosity which eventually encompass the surface. Examination of Zry-2 process tubes reveals that accelerated corrosion has occurred in low-oxygen coolants. Hydrogen contents are relatively low, but show some localized profiles. Gross hydriding has occurred on process tubes containing aluminum spacers, apparently by a galvanic charging mechanism. Titanium paralleled Zry-2 in corrosion behavior under irradiation. Niobium corrosion was variable, but did not appear to be strongly influenced by radiation. Corrosion rates on Inconel and stainless steel were only slightly higher in-flux than out-of-reactor. Corrosion rates on nickel-plated aluminum appeared to vary substantially with preexposure treatments, but the rates generally were accelerated compared to rates on unirradiated coupons. (59 references, 11 tables, 12 figs.)

  4. Detection of helium in irradiated Fe9Cr alloys by coincidence Doppler broadening of slow positron annihilation

    Science.gov (United States)

    Cao, Xingzhong; Zhu, Te; Jin, Shuoxue; Kuang, Peng; Zhang, Peng; Lu, Eryang; Gong, Yihao; Guo, Liping; Wang, Baoyi

    2017-03-01

    An element analysis method, coincidence Doppler broadening spectroscopy of slow positron annihilation, was employed to detect helium in ion-irradiated Fe9Cr alloys. Spectra with higher peak to background ratio were recorded using a two-HPGe detector coincidence measuring system. It means that information in the high-momentum area of the spectra can be used to identify helium in metals. This identification is not entirely dependent on the helium concentration in the specimens, but is related to the structure and microscopic arrangement of atoms surrounding the positron annihilation site. The results of Doppler broadening spectroscopy and transmission electron microscopy show that vacancies and dislocations were formed in ion-irradiated specimens. Thermal helium desorption spectrometry was performed to obtain the types of He traps.

  5. Investigation of irradiation strengthening of bcc metals and their alloys. Progress report, January 1977--October 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Progress is reported in the areas of (a) the effect of neutron damage on the dislocation kinetics in bcc metals and their alloys, and (b) the effect of 3 He on the deformation characteristics of body centered cubic metals and their alloys. Results obtained from these projects are discussed

  6. Mechanism of nucleation and incipient growth of Re clusters in irradiated W-Re alloys from kinetic Monte Carlo simulations

    Science.gov (United States)

    Huang, Chen-Hsi; Gharaee, Leili; Zhao, Yue; Erhart, Paul; Marian, Jaime

    2017-09-01

    High-temperature, high-dose, neutron irradiation of W results in the formation of Re-rich clusters at concentrations one order of magnitude lower than the thermodynamic solubility limit. These clusters may eventually transform into brittle W-Re intermetallic phases, which can lead to high levels of hardening and thermal conductivity losses. Standard theories of radiation-enhanced diffusion and precipitation cannot explain the formation of these precipitates and so understanding the mechanism by which nonequilibrium clusters form under irradiation is crucial to predict material degradation and devise mitigation strategies. Here we carry out a thermodynamic study of W-Re alloys and conduct kinetic Monte Carlo simulations of Re cluster formation in irradiated W-2Re alloys. We use a generalized Hamiltonian for crystals containing point defects parametrized entirely with electronic structure calculations. Our model incorporates recently gained mechanistic information of mixed-interstitial solute transport, which is seen to control cluster nucleation and growth by forming quasispherical nuclei after an average incubation time of 13.5(±8.5 ) s at 1800 K. These nuclei are seen to grow by attracting more mixed interstitials bringing solute atoms, which in turn attracts vacancies leading to recombination and solute agglomeration. Owing to the arrival of both Re and W atoms from the mixed dumbbells, the clusters are not fully dense in Re, which amounts to no more than 50% of the atomic concentration of the cluster near the center. Our simulations are in qualitative agreement with recent atom probe examinations of ion-irradiated W-2Re systems at 773 K.

  7. Abnormal growth of austenite grain of low-carbon steel

    International Nuclear Information System (INIS)

    Yu Qingbo; Sun Ying

    2006-01-01

    Niobium is an important alloying element for the steel. To know further the effect of Nb in the steel, the contrast experiments on the austenite grain growth of the 0.015%Nb and free Nb steels were carried out using Gleeble 1500 thermomechanical simulator. The experimental results indicate that the austenite grain of 0.015%Nb steel is finer than that of Nb free steel at 1150-1230 deg. C. And when the heating temperature arrives the critical temperature 1240 deg. C, the austenite grain of Nb steel suddenly grows up, while the austenite grain of Nb free steel changes little. Finally, the austenite grain of Nb steel is obviously coarser than that of Nb free steel. By transmission electron microscopy (TEM) using a carbon extraction replica technique, the precipitates of Nb(C,N) were not observed in the 0.015%Nb steel. It is concluded that the grain-boundary internal adsorption of Nb atoms leads to the result

  8. Expanded austenite in nitrided layers deposited on austenitic and super austenitic stainless steel grades

    International Nuclear Information System (INIS)

    Casteletti, L.C.; Fernandes, F.A.P.; Heck, S.C.; Gallego, J.

    2010-01-01

    In this work nitrided layers deposited on austenitic and super austenitic stainless steels were analyzed through optical microscopy and X-rays diffraction analysis (XRD). It was observed that the formation of N supersaturated phase, called expanded austenite, has promoted significant increment of hardness (> 1000HV). XRD results have indicated the anomalous displacement of the diffracted peaks, in comparison with the normal austenite. This behavior, combined with peaks broadening, it was analyzed in different nitriding temperatures which results showed good agreement with the literature. (author)

  9. Void swelling and defect processes in ti-modified steels using accelerator irradiation

    International Nuclear Information System (INIS)

    Nair, K.G.M.; Panigrahi, B.K.; Arunkumar, J.; David, C.; Balaji, S.; Rajaraman, R.; Ararebdram, G.; Abhaya, S.; Valsakumar, M.C.; Sunar, C.S.; Raj, B.

    2008-01-01

    The void swelling behaviour of (15Ni-14Cr)-O.25Ti and (15Ni-14Cr)-O.15Ti steels are studied using heavy ion irradiation for understanding the influence of titanium in the void swelling resistance of these D9 alloys. The cold worked samples have been pre-implanted with a uniform helium concentration of 30 appm spanning a width of about 640 nm. This was followed by a 5-MeV nickel ion irradiation to create a peak damage of ∼ 100 dpa at a damage rate of 7 x 10 -3 dpa/s at various irradiation temperatures between 700 and 970 K. The gross swelling in the implanted range is measured by step height measurements. It is found that the peak swelling temperatures and the magnitude of swelling for the alloys are different. The difference in void swelling behaviour with variation in titanium concentration in these two alloys is discussed on the basis of the role of titanium on the vacancy migration and TiC precipitate formation. Isochronal annealing study of the positron lifetime in the un-irradiated alloys reveals different TiC precipitates formation behaviour in the two alloys. Ab initio calculations of positron lifetime, using large super-cells, show that C vacancies at the TiC/austenite interface are the predominant positron trapping centres in these alloys. (authors)

  10. The Prediction of Long-Term Thermal Aging in Cast Austenitic Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Yang, Ying; Lach, Timothy G.

    2017-02-15

    Cast austenitic stainless steel (CASS) materials are extensively used for many massive primary coolant system components of light water reactors (LWRs) including coolant piping, valve bodies, pump casings, and piping elbows. Many of these components are operated in complex and persistently damaging environments of elevated temperature, high pressure, corrosive environment, and sometimes radiation for long periods of time. Since a large number of CASS components are installed in every nuclear power plant and replacing such massive components is prohibitively expensive, any significant degradation in mechanical properties that affects structural integrity, cracking resistance in particular, of CASS components will raise a serious concern on the performance of entire power plant. The CASS materials for nuclear components are highly corrosion-resistant Fe-Cr-Ni alloys with 300 series stainless steel compositions and mostly austenite (γ)–ferrite (δ) duplex structures, which result from the casting processes consisting of alloy melting and pouring or injecting liquid metal into a static or spinning mold. Although the commonly used static and centrifugal casting processes enable the fabrication of massive components with proper resistance to environmental attacks, the alloying and microstructural conditions are not highly controllable in actual fabrication, especially in the casting processes of massive components. In the corrosion-resistant Fe-Cr-Ni alloy system, the minor phase (i.e., the δ-ferrite phase) is inevitably formed during the casting process, and is in a non-equilibrium state subject to detrimental changes during exposure to elevated temperature and/or radiation. In general, relatively few critical degradation modes are expected within the current design lifetime of 40 years, given that the CASS components have been processed properly. It has been well known, however, that both the thermal aging and the neutron irradiation can cause degradation of static

  11. Corrosion And Thermal Processing In Cold Gas Dynamic Spray Deposited Austenitic Stainless Steel Coatings

    Science.gov (United States)

    2016-06-01

    testing (ASTM G5) of low pressure cold spray austenitic stainless steel coatings. Several different powders and heat treatments will be applied to...diffusion eliminating the local low chromium region. The low carbon type stainless steel alloys as used here are generally considered to be...maximum 200words) This thesis presents research on the corrosion properties and effects of heat treatment on austenitic stainless steel coatings

  12. Materials design of high nitrogen manganese austenitic stainless TWIP steels for strip casting

    OpenAIRE

    Mosecker, Linda

    2016-01-01

    High nitrogen manganese austenitic stainless TWIP steels achieve attractive mechanical properties and excellent strain hardening behavior. However, high nitrogen steel melting methods are generally associated with high pressures to enhance the nitrogen solubility in the melt. Thin strip casting offers an attractive option that not only shortens the process route but also allows the alloying with nitrogen at atmospheric pressure. In the present work, the materials design of austenitic Fe-Cr-Mn...

  13. Influence of an external magnetic field on damage by self-ion irradiation in Fe90Cr10 alloy

    Directory of Open Access Journals (Sweden)

    Fernando José Sánchez

    2016-12-01

    Full Text Available The effect of an external magnetic field (B=0.5 T on Fe90Cr10 specimens during Fe ion irradiation, has been investigated by means of Conversion Electron Mössbauer Spectroscopy (CEMS. The analysis has revealed significant differences in the average hyperfine magnetic field (=0.3 T between non-irradiated and irradiated samples as well as between irradiations made with B (w/ B and without B (w/o B. It is considered that these variations can be due to changes in the local environment around the probe nuclei (57Fe; where vacancies and Cr distribution play a role. The results indicate that the Cr distribution in the neighbourhood of the iron atoms could be changed by the application of an external field. This would imply that an external magnetic field may be an important parameter to take into account in predictive models for Cr behaviour in Fe–Cr alloys, and especially in fusion conditions where intense magnetic fields are required for plasma confinement.

  14. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  15. Study of clustering point defects under irradiation in dilute iron alloys; Etude de la formation sous irradiation des amas de defauts ponctuels dans les alliages ferritiques faiblement allies

    Energy Technology Data Exchange (ETDEWEB)

    Duong-Hardouin Duparc, T.H.A

    1998-12-31

    In low copper steels for nuclear reactor pressure vessel, point defect clustering plays an important role in hardening. These clusters are very small and invisible by transmission electron microscopy. In order to study the hardening component which results from the clustering of freely migrating point defects, we irradiated in a high voltage electron microscope Fe, the FeCu{sub 0.13%}, FeP{sub 0.015%} and FeN{sub 33ppm} alloys and the complex FeMn{sub 1.5%}Ni{sub 0.8%}Cu{sub 0.13%}P{sub 0.01%} alloy the composition of which is close to the matrix of pressure vessel steel. We studied the nucleation of dislocation loops and their growth velocity. The observations and the analyses have shown that in the complex model alloy, the interstitial dislocation loops are smaller and their density is more important than for the others alloys. The diffusion coefficients of interstitials and vacancies are obtained with the help of a simplified model. The densities of dislocation loops and their growth velocities obtained experimentally are reproduced by means of a cluster dynamics model we have developed. This is achieved self-consistently by using as a first trial the approximated coefficients obtained with the simplified model. The results of calculations have shown that the binding energy of di-interstitials must be very important in the binary iron alloys and only 0.95 eV in iron. Copper, nitrogen and phosphorus stabilize di-interstitials in iron. Finally the distribution of interstitial loops at 290 deg C and at 2.10{sup -9} dpa/s is calculated with the diffusion coefficient of point defects adjusted in FeCu. A distribution of small loops appears which gives an increase of hardening estimated to 10 Hv instead of 33 Hv experimentally observed. This low value can be improved by assuming in agreement with molecular dynamics simulations that a little fraction of di-interstitials is created at 2.5 MeV. (author) 111 refs.

  16. Austenitic stainless steel for high temperature applications

    International Nuclear Information System (INIS)

    Johnson, G. D.; Powell, R. W.

    1985-01-01

    This invention describes a composition for an austenitic stainless steel which has been found to exhibit improved high temperature stress rupture properties. The composition of this alloy is about (in wt. %): 12.5 to 14.5 Cr; 14.5 to 16.5 Ni; 1.5 to 2.5 Mo; 1.5 to 2.5 Mn; 0.1 to 0.4 Ti; 0.02 to 0.08 C; 0.5 to 1.0 Si; 0.01 maximum, N; 0.02 to 0.008 P; 0.002 to 0.008 B; 0.004-0.0010 S; 0.02-0.05 Nb; 0.01-0.05 V; 0.005-0.02 Ta; 0.02-0.05 Al; 0.01-0.04 Cu; 0.02-0.05 Co; 0.03 maximum, As; 0.01 maximum, O; 0.01 maximum, Zr; and with the balance of the alloy being essentially iron. The carbon content of the alloy is adjusted such that wt. % Ti/(wt. % C+wt. % N) is between 4 and 6, and most preferably about 5. In addition the sum of the wt. % P+wt. % B+wt. % S is at least 0.03 wt. %. This alloy is believed to be particularly well suited for use as fast breeder reactor fuel element cladding

  17. Non-uniformity of hot plastic strain of stainless steels with austenitic-ferritic structure

    International Nuclear Information System (INIS)

    Laricheva, L.P.; Peretyat'ko, V.N.; Rostovtsev, A.N.; Levius, A.M.

    1987-01-01

    Non-uniformity of hot strain of stainless steels of various alloying was investigated. Steels with austenite and δ-ferrite structure of two classes were chosen for investigation: 08Kh18N10T steel of austenitic class and 08Kh21N5T steel of austenitic-ferritic class. Tests were conducted for samples subjected to preliminary thermal treatment: heating up to 1250 deg C, holding during 0.5 h, cooling in water. The heat treatment enabled to produce large grains of austenite and δ-ferrite (about 30 μm) in 08Kh21N5T steel, and sufficient amount of δ-ferrite (up to 50%) in 08Kh18N10T steel. It is shown that hot strain of austenitic-ferritic steels is non-uniform. δ-ferrite strain is more pronounced as compared to austenite. The ratio of mean δ-ferrite strain to the mean austenite strain grows with increase of the degree of general steel strain and temperature. The ratio of mean phase strains in 08Kh18N10T steel is higher as compared to 08Kh21N5T steel, general strain and temperature being equal. Temperature effect on the ratio of δ-ferrite and austenite strains is more pronounced for 08Kh18N10T steel. It is explaind by the value of ratios of phase strain resistance and temperature effect on them

  18. Thermodynamic stability of austenitic Ni-Mn-Cu cast iron

    Directory of Open Access Journals (Sweden)

    A. Janus

    2014-07-01

    Full Text Available The performed research was aimed at determining thermodynamic stability of structures of Ni-Mn-Cu cast iron castings. Examined were 35 alloys. The castings were tempered at 900 °C for 2 hours. Two cooling speeds were used: furnace-cooling and water-cooling. In the alloys with the nickel equivalent value less than 20,0 %, partial transition of austenite to martensite took place. The austenite decomposition ratio and the related growth of hardness was higher for smaller nickel equivalent value and was clearly larger in annealed castings than in hardened ones. Obtaining thermodynamically stable structure of castings requires larger than 20,0 % value of the nickel equivalent.

  19. Effects of 600 MeV proton irradiation on nucleation and growth of precipitates and helium bubbles in a high-purity Al-Mg-Si alloy

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Leffers, Torben; Victoria, M.

    1986-01-01

    Solution treated specimens of a high-purity Al-0.75%Mg-0.42%Si alloy were irradiated with 600 MeV protons at 150 and 240°C to a dose level of 0.47 and 0.55 dpa, respectively. Mg2Si-type precipitates formed during irradiation at 150 and 240°C; at 240°C, however, a large number of precipitates seem...

  20. Morphology variation, composition alteration and microstructure changes in ion-irradiated 1060 aluminum alloy

    Science.gov (United States)

    Wan, Hao; Si, Naichao; Wang, Quan; Zhao, Zhenjiang

    2018-02-01

    Morphology variation, composition alteration and microstructure changes in 1060 aluminum irradiated with 50 keV helium ions were characterized by field emission scanning electron microscopy (FESEM) equipped with x-ray elemental scanning, 3D measuring laser microscope and transmission electron microscope (TEM). The results show that, helium ions irradiation induced surface damage and Si-rich aggregates in the surfaces of irradiated samples. Increasing the dose of irradiation, more damages and Si-rich aggregates would be produced. Besides, defects such as dislocations, dislocation loops and dislocation walls were the primary defects in the ion implanted layer. The forming of surface damages were related with preferentially sputtering of Al component. While irradiation-enhanced diffusion and irradiation-induced segregation resulted in the aggregation of impurity atoms. And the aggregation ability of impurity atoms were discussed based on the atomic radius, displacement energy, lattice binding energy and surface binding energy.

  1. Effect of double ion implantation and irradiation by Ar and He ions on nano-indentation hardness of metallic alloys

    Science.gov (United States)

    Dayal, P.; Bhattacharyya, D.; Mook, W. M.; Fu, E. G.; Wang, Y.-Q.; Carr, D. G.; Anderoglu, O.; Mara, N. A.; Misra, A.; Harrison, R. P.; Edwards, L.

    2013-07-01

    In this study, the authors have investigated the combined effect of a double layer of implantation on four different metallic alloys, ODS steel MA957, Zircaloy-4, Ti-6Al-4V titanium alloy and stainless steel 316, by ions of two different species - He and Ar - on the hardening of the surface as measured by nano-indentation. The data was collected for a large number of indentations using the Continuous Stiffness Method or "CSM" mode, applying the indents on the implanted surface. Careful analysis of the data in the present investigations show that the relative hardening due to individual implantation layers can be used to obtain an estimate of the relative hardening effect of a combination of two separate implanted layers of two different species. This combined hardness was found to lie between the square root of the sum of the squares of individual hardening effects, (ΔHA2 + ΔHB2)0.5 as the lower limit and the sum of the individual hardening effects, (ΔHA + ΔHB) as the upper limit, within errors, for all depths measured. The hardening due to irradiation by different species of ions was calculated by subtracting the average hardness vs. depth curve of the un-irradiated or "virgin" material from that of the irradiated material. The combined hardening of the irradiated samples due to Ar and He irradiation was found to be described well by an approximate upper bound given by the simple linear sum of the individual hardening (L) and a lower bound given by the square root of the sum of the squares (R) of the individual hardening effects due to Ar and He irradiation along the full depth of the indentation. The peak of the combined hardness of Ar and He irradiated material appears at the depth predicted by both the R and the L curves, in all samples. The combined hardness increase due to Ar and He irradiation lies near the upper limit (L curve) for the ODS steel MA957, somewhere in between L and R curves for Zircaloy-4, and near the R curve for the stainless steel 316

  2. Effect of MeV ion irradiation on the coefficient of thermal expansion of Fe-Ni Invar alloys: A combinatorial study

    Energy Technology Data Exchange (ETDEWEB)

    Zheng Xuan, E-mail: xuan.zheng@seagate.com [Department of Materials Science and Engineering, Frederick-Seitz Materials Research Laboratory, University of Illinois, Urbana, IL 61801 (United States); Cahill, David G. [Department of Materials Science and Engineering, Frederick-Seitz Materials Research Laboratory, University of Illinois, Urbana, IL 61801 (United States); Zhao Jicheng [Department of Materials Science and Engineering, Ohio State University, Columbus, OH 43210 (United States)

    2010-02-15

    The effect of low-dose MeV ion irradiation on the thermal conductivity and coefficient of thermal expansion (CTE) of Fe-Ni alloys has been studied over the entire composition range using a combinatorial method. With our experimental precision of {+-}5%, the thermal conductivity of Fe-Ni alloys is unchanged by 2.3MeVAr{sup +} irradiation at a fluence of 1.2x10{sup 14}ionscm{sup -2}. The CTE of Fe-Ni Invar alloys with Ni concentrations between 30 and 37 at.% increases significantly upon 2.3MeVAr{sup +} irradiation. At the Invar composition of Fe{sub 65}Ni{sub 35}, the CTE increases from 0.5x10{sup -6}K{sup -1} before ion irradiation to 4.3x10{sup -6}K{sup -1} after irradiation for an ion fluence of 1.2x10{sup 14}ionscm{sup -2} or, equivalently, 0.1 displacements per atom. We attribute this increase in CTE to changes in atomic short-range order created by ion irradiation.

  3. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C. (ITER R and D Task no. T213)

    International Nuclear Information System (INIS)

    Singh, B.N.; Stubbins, J.F.; Toft, P.

    2000-03-01

    The fatigue behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of ∼2.5 x 10 17 n/m 2 s (E> 1 MeV) to influence levels of 1.0 - 1.5 x 10 24 n/m 2 (E> 1 MeV) at 250 and 350 deg. C. These irradiations were carried out in temperature controlled rigs where the irradiation temperature was monitored and controlled continuously throughout the whole irradiation experiment. Both unirradiated and irradiated specimens were fatigue tested in vacuum at the irradiation temperatures of 250 and 350 deg. C in a strain controlled mode with a loading frequency of 0.5Hz. Post-fatigue microstructures were examined using transmission electron microscopy and the fracture surfaces were investigated using scanning electron microscope. The present investigations demonstrated that the fatigue life decreases with increasing temperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures and the fracture surface morphology. Finally, the main conclusions and their implications are summarised. (au)

  4. Phase-field Model for Interstitial Loop Growth Kinetics and Thermodynamic and Kinetic Models of Irradiated Fe-Cr Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yulan; Hu, Shenyang Y.; Sun, Xin; Khaleel, Mohammad A.

    2011-06-15

    strength of interstitial loop for interstitials. In part II, we present a generic phase field model and discuss the thermodynamic and kinetic properties in phase-field models including the reaction kinetics of radiation defects and local free energy of irradiated materials. In particular, a two-sublattice thermodynamic model is suggested to describe the local free energy of alloys with irradiated defects. Fe-Cr alloy is taken as an example to explain the required thermodynamic and kinetic properties for quantitative phase-field modeling. Finally the great challenges in phase-field modeling will be discussed.

  5. Influence of irradiation parameters on damage accumulation in metals and alloys

    DEFF Research Database (Denmark)

    Singh, B.N.; Zinkle, S.J.

    1994-01-01

    It is well known that a fraction of defects produced during irradiation accumulate in crystalline solids in the form of clusters of self-interstitial atoms (SIAs) and vacancies, loops, tetrahedra, dislocation segments and cavities. The irradiation parameters such as recoil energy, damage rate...

  6. Some data of second sequence non standard austenitic ingot, A2

    International Nuclear Information System (INIS)

    Nurdin Effendi; Aziz K Jahja; Bandriana; Wisnu Ari Adi

    2012-01-01

    Synthesis of second sequence austenite stainless steel named A2 using extracted minerals from Indonesian mines has been carried out. The starting materials for austenite alloy consist of granular ferro scrap, nickel, ferro-chrome, ferro-manganese, and ferro-silicon. The second sequence composition differs from the former first sequence. This A2 sequence contained more nickel, meanwhile titanium element had not been added explicitly to it, and just been found from raw materials contents or impurities, as well as carbon content in the alloy. However before the actual alloying work started, the first important step was to carry out the determination of the fractional amount of each starting material necessary to form an austenite stainless steel alloy as specified. Once the component fraction of each base alloy-element was determined, the raw materials are weighed on the mini-balance. After the fractional quantities of each constituent have been computed, an appropriate amount of these base materials are weighed separately on the micro scale. The raw materials were then placed in the induction foundry furnace, which was operated by an electromagnetic inductive-thermal system. The foundry furnace system performs the stirring of the molten materials automatically. The homogenized molten metals were poured down into sand casting prepared in advance. Some of the austenite stainless steel were normalized at 600°C for 6 hours. The average density is 7.8 g cm -1 and the average hardness value of 'normalized' austenite stainless-steels is in the range of 460 on the Vickers scale. The microstructure observation concludes that an extensive portion of the sample's structure is dendritic and the surface turns out to be homogenous. X-ray diffraction analysis shows that the material belongs to the fcc crystallographic system, which fits in with the austenite class of the alloy. The experimental fractional elemental composition data acquired by OES method turn out to

  7. Shape memory alloys – characterization techniques

    Indian Academy of Sciences (India)

    Shape memory alloys are the generic class of alloys that show both thermal and mechanical memory. The basic physics involved in the shape memory effect is the reversible thermoelastic martensitic transformation. In general, there exists two phases in shape memory alloys, viz., a hightemperature phase or austenitic ...

  8. Development of Computational Tools for Predicting Thermal- and Radiation-Induced Solute Segregation at Grain Boundaries in Fe-based Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-30

    Radiation-induced segregation (RIS) has been frequently reported in structural materials such as austenitic, ferritic, and ferritic-martensitic stainless steels (SS) that have been widely used in light water reactors (LWRs). RIS has been linked to secondary degradation effects in SS including irradiation-induced stress corrosion cracking (IASCC). Earlier studies on thermal segregation in Fe-based alloys found that metalloids elements such as P, S, Si, Ge, Sn, etc., embrittle the materials when enrichment was observed at grain boundaries (GBs). RIS of Fe-Cr-Ni-based austenitic steels has been modeled in the U.S. 2015 fiscal year (FY2015), which identified the pre-enrichment due to thermal segregation can have an important role on the subsequent RIS. The goal of this work is to develop thermal segregation models for alloying elements in steels for future integration with RIS modeling.

  9. The effect of metal primer application and Nd:YAG laser irradiation on the shear-bond strength between polymethyl methacrylate and cobalt-chromium alloy.

    Science.gov (United States)

    Yilmaz, Asude; Akyil, Musa Şamil; Hologlu, Bilal

    2011-01-01

    The purpose of this study was to compare the shear-bond strengths (SBSs) of an acrylic resin and a cobalt-chromium (Co-Cr) alloy after applying a metal primer, Nd:YAG laser irradiation, or both to the sandblasted surface of the Co-Cr alloy. The serviceability of a removable partial denture (RPD) is dependent on the bond strength at the resin-alloy interface. No previously published studies exist on the use of Nd:YAG lasers for preparing the surface of a Co-Cr alloy in an RPD to obtain a high-strength bond between PMMA and the alloy. One-hundred twenty Co-Cr alloy specimens were sandblasted and randomly assigned to four equal groups: Group I, sandblasting; Group II, sandblasting + metal primer; Group III, sandblasting + Nd:YAG laser; and Group IV, sandblasting + Nd:YAG laser + metal primer. To establish the most appropriate fluence for modifying the surface of the sandblasted cast specimens, we conducted a preliminary study. Nd:YAG laser irradiation at a fluence of 46.9 J/cm(2) was selected. After the various surface treatments, each alloy specimen was embedded in PMMA to determine the SBS between PMMA and the alloy. Group II and III specimens exhibited higher SBSs than did those of the Group I specimens (p < 0.05), and Group IV specimens showed higher SBSs than did those of the Group II and III specimens (p < 0.05). A significant difference existed in failure types among groups (p < 0.05). Failure type was predominantly adhesive for groups I and III, but predominantly mixed for groups II and IV. Nd:YAG laser irradiation at a fluence of 46.9 J/cm(2) roughens the sandblasted surface of a Co-Cr alloy and increases the strength of the bond between PMMA and the alloy. This bond strength can be increased further by applying a metal primer to the laser-irradiated surface.

  10. Lightweight Multifunctional Linear Cellular Alloy Ballistic Structures

    Science.gov (United States)

    2006-04-26

    densities of 10, 15 and 20 % with the dimensions shown in Table 1. The alloy compositions were high strength maraging steel (M200) and Super Invar ... alloys made from LCA processing3 are shown in Table 3. Super Invar in the as-reduced state is a ductile (25-30%) austenitic alloy . When cooled to...Final Report for Lightweight Multifunctional Linear Cellular Alloy Ballistic Structures from Structured Alloys , Inc. Joe K

  11. Design of Wear-Resistant Austenitic Steels for Selective Laser Melting

    Science.gov (United States)

    Lemke, J. N.; Casati, R.; Lecis, N.; Andrianopoli, C.; Varone, A.; Montanari, R.; Vedani, M.

    2018-03-01

    Type 316L stainless steel feedstock powder was modified by alloying with powders containing carbide/boride-forming elements to create improved wear-resistant austenitic alloys that can be readily processed by Selective Laser Melting. Fe-based alloys with high C, B, V, and Nb contents were thus produced, resulting in a microstructure that consisted of austenitic grains and a significant amount of hard carbides and borides. Heat treatments were performed to modify the carbide distribution and morphology. Optimal hard-phase spheroidization was achieved by annealing the proposed alloys at 1150 °C for 1 hour followed by water quenching. The total increase in hardness of samples containing 20 pct of C/B-rich alloy powder was of 82.7 pct while the wear resistance could be increased by a factor of 6.

  12. Irradiation enhancement of water corrosion of Zr alloys: materials and water radiochemistry aspects of oxide growth

    International Nuclear Information System (INIS)

    Iltis, X.; Salot, R.; Lefebvre, F.; Lemaignan, C.

    1994-01-01

    It is well known that under irradiation the corrosion rates of Zircaloy's are significantly increased. The aim of this paper is to present the two main mechanisms by which the irradiation is supposed to contribute to such a corrosion rate enhancement. The first one concerns the effects of irradiation on the base metal and their consequences on the zirconia nucleation and growth. The second one implies the contribution of water radiolysis and is proposed as an explanation for all the reported cases of local or gross corrosion enhancement. (authors). 3 figs., 13 refs

  13. Microstructural evolution and hardening of GH3535 alloy under energetic Xe ion irradiation at room temperature and 650 °C

    Science.gov (United States)

    Huang, Hefei; Gao, Jie; Radiguet, Bertrand; Liu, Renduo; Li, Jianjian; Lei, Guanhong; Huang, Qing; Liu, Min; Xie, Ruobing

    2018-02-01

    The GH3535 alloy was irradiated with 7 MeV Xe26+ ions to a dose of 10 dpa at room temperature (RT) and 650 °C, and subsequently examined using Transmission Electron Microscopy (TEM) and nanoindentation. High numbers of nano-sized black dots, identified as dislocation loops were observed in both irradiated samples. The dislocation loops detected at the high temperature irradiated sample (size/number density: 9.5 nm/1.9 × 1021 m-3) were found to be larger in size but less in amount as compared to that of the case of RT irradiation (6.9 nm/18.7 × 1021 m-3). In addition, the large-sized Mo-Cr rich precipitates (16.4 nm/3.7 × 1021 m-3) were observed in the sample irradiated at 650 °C. Moreover, the Xe bubbles, with smaller size (2.9 nm) but higher number density (77.8 × 1021 m-3) among the irradiated induced defects, were also detected in the case of high temperature irradiated sample via the diffusion and aggregation of Xe atoms. Nanoindentaion measurements showed a hardening phenomenon for the irradiated sample, and the hardness increment is higher in the case of high temperature irradiated sample. Dispersed barrier-hardening (DBH) model was applied to predict the hardening produced from the irradiation induced defects. The yield strength increment calculated based on TEM observations and the nanohardness increment measured using nanoindentation are in excellent agreement.

  14. Modifications in morphological, structural, electrical and mechanical properties of Fe-1.0 wt.% Cu alloy on irradiation with 532 nm-6 ns Nd:YAG laser shots

    Science.gov (United States)

    Butt, M. Z.; Ur-Rehman, Khalil; Ali, Dilawar; Aftab, Muzamil; Usman Tanveer, M.

    2017-09-01

    The role of laser irradiation in tailoring the mechanical and electrical properties of Fe-1.0 wt.% Cu alloy has been explored, and compared with that of energetic particle irradiation. Mechanically polished/annealed Fe-1.0 wt.% Cu alloy specimens were irradiated in a vacuum ~10-3 Torr with 1-2500 laser shots. The laser fluence and laser intensity at the laser spot on the target surface were 86.54 J cm-2 and 14.4  ×  109 W cm-2, respectively. Different surface morphological features, e.g. dips, ripples, ridges, nanohillocks, microcones, micropillars, cavities, grooves, nano- and micro-size droplets, as well as tadpole-like structures, etc, were observed by SEM. Surface roughness was maximum for 1 laser shot, decreased by 40% for 250 laser shots, and then reduced by 70% in the range 500-2500. Crystallite size D and lattice strain ɛ were determined by Williamson-Hall analysis of x-ray diffractographs. An increasing linear relationship between electrical resistivity and Vickers hardness was found. In a plot of hardness versus D -1/2, crossover from Hall-Petch to Inverse Hall-Petch relation was observed at a critical value of D  ≈  11 nm. This behaviour was also followed by electrical resistivity. It is established that the nature of microstructural changes produced in the alloy on laser irradiation and on energetic particle irradiation is different.

  15. Influence of neutron irradiation on the magnetic properties of the Fe Ni pure alloy and with impurities of Si and Mo

    International Nuclear Information System (INIS)

    Lucki, George

    1971-01-01

    Hysteresis loop, Initial permeability and Curie Temperature measurements were conducted on several pure and polluted (with Si and Mo) Fe Ni 50-50% at. alloys. Isochronal annealings were performed between 25 deg 65 deg C, on each composition in three different ways: quenched (anisotropic) samples; quenched and irradiated samples; quenched irradiated samples annealed with saturating magnetic field. The experiment showed a sharp decrease in all parameters of the polluted alloys. Fast neutron irradiation results indicated that the magnetic properties are affected by the defects created during irradiation. The effect of thermal treatment, magnetic annealing and irradiation is greatest in anisotropic alloys. It is considered that magnetic annealing introduces a uniaxial anisotropy that tends to increase the remanence and hence the squareness of the hysteresis loop; but an increase in both remanence and coercivity was measured even in absence of the magnetic field. Magnetic after effect has been detected and a simple model for the diffusion of defects is presented. Many models have been proposed to explain the resultant properties, the most feasible being that based upon short-range ordering, proposed by Neel and Taniguchi, together with the interesting hypothesis of Heidenreich and Nesbitt. (author)

  16. Characterization of the sodium corrosion behavior of commercial austenitic steels

    International Nuclear Information System (INIS)

    Shiels, S.A.; Bagnall, C.; Keeton, A.R.; Witkowski, R.E.; Anantatmula, R.P.

    1980-01-01

    During the course of an on-going evaluation of austenitic alloys for potential liquid metal fast breeder reactor (LMFBR) fuel pin cladding application, a series of commercial alloys was selected for study. The data obtained led to the recognition of an underlying pattern of behavior and enabled the prediction of surface chemistry changes. The changes in surface topographical development from alloy to alloy are shown and the important role played by the element molybdenum in this development is indicated. The presentation also illustrates how a total damage equation was evolved to encompass all aspects of weight loss and metal/sodium interactions: wall thinning ferrite layer formation and intergranular attack. The total damage equation represents a significant departure from the classical description of sodium corrosion in which weight loss is simply translated into wall thinning

  17. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Edwards, D.J.; Singh, B.N.; Bilde-Sørensen, Jørgen

    2005-01-01

    The formation of ‘cleared’ channels in neutron irradiated metals and alloys have been frequently reported for more than 40 years. So far, however, no unambiguous and conclusive evidence showing as to how and where these channels are initiated has emerged. In the following we present experimental...... results illustrating initiation and propagation of channels during post-irradiation deformation of neutron irradiated copper and a copper alloy. The observations strongly suggest that the channels are initiated at boundaries, large inclusions and even at previously formed cleared channels. Some...... of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these results are discussed with specific reference to the origin and consequences of plastic flow localization....

  18. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-03-28

    Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

  19. Market Opportunities for Austenitic Stainless Steels in SO2 Scrubbers

    Science.gov (United States)

    Michels, Harold T.

    1980-10-01

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

  20. Effect of ultraviolet irradiation on the osseointegration of a titanium alloy with bone

    Directory of Open Access Journals (Sweden)

    Ashish Yadav

    2017-01-01

    Full Text Available Introduction: Attempt has been made to analyze the potential of titanium (Ti alloy for osteointegration by the effect of surface photo functionalization in different aspects as follows: in Ringer's solution, in vitro cell growth, and in vivo study on rabbit. The present study was aimed to investigate the influence of ultraviolet (UV light on surface topography, corrosion behavior, and bioactivity of indigenously manufactured samples of Ti alloy mini-implant. Materials and Methods: The study includes surface modification of Ti samples by UV treatment, corrosion testing of the specimens using Potentiostat (GAMRY System, qualitative examination of modified surface topography using scanning electron microscope, and cellular viability test on Ti alloy surface (3-(4,5-Dimethylthiazol-2-yl-2,5-diphenyltetrazolium bromide ASSAY. To find the effect of UV light on implant bone integration, biochemical test was performed on the femur of rabbits. Results and Discussion: Corrosion resistance of untreated Ti alloy in Ringer's solution was found to be less, whereas corrosion rate was more. Corrosion resistance of UV-treated samples was found to increase significantly, thereby lowering the corrosion rate. Cell growth in UV-treated specimen was observed to be higher than that in untreated samples. It is important to mention that cell growth was significantly enhanced on samples which were UV treated for longer duration of time. Conclusions: There was a marked improvement in cell growth on UV-treated Ti alloy samples. Hence, it is expected that it would enhance the process of osseointegration of Ti with bone. Another important finding obtained was that the removal torque values of UV-treated implants were higher than that of untreated implants. The overall result reveals that UV treatment of implants does help us in speeding up the osseointegration process.

  1. Effects of HVEM irradiation on ordered phases in Ni-Ti

    International Nuclear Information System (INIS)

    Pelton, A.R.

    1983-01-01

    Various ordered phases in the Ni-Ti system were subjected to electron irradiation in the Berkeley HVEM. Austenitic NiTi (B2 structure) disorders and turns amorphous with room-temperature irradiations at accelerating potentials between 1 and 1.5 MeV. Total doses for the onset of amorphiticity range between 0.7 x 10 22 and 3 x 10 22 e.cm -2 (0.4 to 1.0dpa). At 90K the dose requirement decreases to 4 x 10 20 e.cm -2 (approx. 10 -2 dpa). Martensitic NiTi (distorted monoclinic structure) readily detwins and transforms to austenite when irradiated for short times (approx. 10 seconds). Vapor-deposited amorphous films were crystallized to produce NiTi, Phase X (ordered nickel-rich phase with unknown structure) and Ni 3 Ti (DO 24 structure). Upon electron irradiation, NiTi and Phase X disorder and become amorphous, while Ni 3 Ti disorders but does not turn amorphous with doses up to 4 x 10 22 e.cm -2 at 90K. These results are discussed in terms of the requirement of a critical concentration of defects and their relative mobilities. Brimhall's solubility criteria for amorphization of ordered alloys by ion bombardment is apparantly applicable to electron-induced crystalline to amorphous transitions in this alloy

  2. Effects of Austenitizing Conditions on the Microstructure of AISI M42 High-Speed Steel

    Directory of Open Access Journals (Sweden)

    Yiwa Luo

    2017-01-01

    Full Text Available The influences of austenitizing conditions on the microstructure of AISI M42 high-speed steel were investigated through thermodynamic calculation, microstructural analysis, and in-situ observation by a confocal scanning laser microscope (CSLM. Results show that the network morphology of carbides could not dissolve completely and distribute equably in the case of the austenitizing temperature is 1373 K. When the austenitizing temperature reaches 1473 K, the excessive increase in temperature leads to increase in carbide dissolution, higher dissolved alloying element contents, and unwanted grain growth. Thus, 1453 K is confirmed as the best austenitizing condition on temperature for the steel. In addition, variations on the microstructure and hardness of the steel are not obvious when holding time ranges from 15 to 30 min with the austenitizing temperature of 1453 K. However, when the holding time reaches 45 min, the average size of carbides tends to increase because of Ostwald ripening. Furthermore, the value of Ms and Mf decrease with the increase of cooling rate. Hence, high cooling rate can depress the martensitic transformation and increase the content of retained austenite. As a result, the hardness of the steel is the best (65.6 HRc when the austenitizing temperature reaches 1453 K and is held for 30 min.

  3. Comparing the possibilities of austenite content determination in austempered ductile iron

    Directory of Open Access Journals (Sweden)

    D. Myszka

    2011-07-01

    Full Text Available The article presents various methods for assessment of the austenite volume fraction in Austempered Ductile Iron (ADI. Tests were carried out on two types of ADI, i.e. unalloyed and alloyed with the addition of 0.72%Cu and 0.27%Mo, heat treated under different conditions of isothermal transformation to obtain different austenite volume fractions. The test material was then subjected to metallographic examinations, X-ray diffraction (XRD analysis, an analysis using the author's genuine programme of artificial neural networks, image analysis and magnetic measurements. The results were compared with each other indicating the possibility of a quantitative measurement of austenite and other phases present in cast iron. It was found that different methods of measurement are not fully consistent with each other but show similar results of the austenite content.

  4. Application of Moessbauer effect in the study of austenite retained in low carbon steel

    International Nuclear Information System (INIS)

    Azevedo, A.L.T. de; Silva, E.G. da

    1979-01-01

    Moessbauer effect measurements of two samples of low carbon alloy having micro-structure of granular bainite type and martensite type have been done. The concentration of the retained austenite in both samples was determined by Moessbauer effect and x-rays there, being agreement for the higher austenite content sample. Concentration of carbon in the MA (Martensite - Austenite) constituents of bainite is also ditermined, the results being in agreement with metallographic considerations. Carbon enrichments are shown as responsible by the stabilization of the austenite in the granular bainite. Spectra of both samples present three magnetic configurations for α-iron with medium magnetic fields iqual to 335, 307 and 280 KOe. (A.R.H.) [pt

  5. Role of Austenitization and Pre-Deformation on the Kinetics of the Isothermal Bainitic Transformation

    Science.gov (United States)

    Lambers, H.-G.; Tschumak, S.; Maier, H. J.; Canadinc, D.

    2009-06-01

    The role of time-temperature path on the isothermal austenite-to-bainite phase transformation of low alloy 51 CrV 4 steel was investigated and the corresponding microstructures were analyzed. The important finding is that an incomplete initial austenitization treatment leaves undissolved carbides in the matrix, such that lower carbon and chromium content in the matrix result, eventually accelerating the phase transformation. Furthermore, the residual carbides constitute additional nucleation sites for the bainite plates, speeding up the process even further. Also, both plastic pre-deformation of the supercooled austenite and application of external elastic stresses during the phase transformation lead to transformation plasticity by enhancing the stress fields, providing a driving force for the growth of bainite plates along a preferred orientation. Overall, the current results constitute the first step toward establishing a database for constructing a realistic microstructure-based model for simulating metal forming operations involving austenite-to-bainite phase transformation.

  6. Effects of proton irradiation on nanocluster precipitation in ferritic steel containing fcc alloying additions

    International Nuclear Information System (INIS)

    Zhang, Z.W.; Liu, C.T.; Wang, X-.L.; Miller, M.K.; Ma, D.; Chen, G.; Williams, J.R.; Chin, B.A.

    2012-01-01

    Newly developed precipitate-strengthened ferritic steels with and without pre-existing nanoscale precipitates were irradiated with 4 MeV protons to a dose of ∼5 mdpa at 50 °C and subsequently examined by nanoindentation and atom probe tomography. Irradiation-enhanced precipitation and coarsening of pre-existing nanoscale precipitates were observed. Cu partitions to the precipitate core along with a segregation of Ni, Al and Mn to the precipitate/matrix interface after both thermal aging and proton irradiation. Proton irradiation induces the precipitation reaction and coarsening of pre-existing nanoscale precipitates, and these results are similar to a thermal aging process. The precipitation and coarsening of nanoscale precipitates are responsible for the changes in hardness. The observation of the radiation-induced softening is essentially due to the coarsening of the pre-existing Cu-rich nanoscale precipitates. The implication of the precipitation on the embrittlement of reactor-pressure-vessel steels after irradiation is discussed.

  7. Mechanical properties and microstructure of neutron irradiated cold worked Al-6063 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Munitz, A.; Shtechman, A.; Cotler, C.; Dahan, S. [Nuclear Res. Center-Negev, Beer-Sheva (Israel); Talianker, M. [Ben-Gurion Univ., Beer-Sheva (Israel). Dept. of Materials Science

    1998-01-01

    The impact of neutron irradiation on the mechanical properties and fracture morphology of cold worked Al-6063 were studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens (50 mm long and 6 mm wide gauge sections) were punched out from an Al-6063 23% cold worked tubes, which had been exposed to prolonged neutron irradiation of up to 4.5 x 10{sup 25} thermal neutrons/m{sup 2} (E < 0.625 eV). The temperature ranged between 41 and 52 C. The tensile specimens were then tensioned till fracture in an Instron tensiometer with strain rate of 2 x 10{sup -3} s{sup -1}. The uniform elongation and the ultimate tensile strength increase as functions of fluence. Metallographic examination and fractography reveal a decrease in the local area reduction of the final fracture necking. This reduction is accompanied with a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. No voids could be observed up to the maximum fluence. The dislocation density of cold worked Al decreases with the thermal neutron fluence. Prolonged annealing of unirradiated cold worked Al-6063 at 52 C revealed similar results. It thus appears that under our irradiation conditions the temperature during irradiation is the major factor influencing the mechanical properties and the microstructure during irradiation. (orig.). 23 refs.

  8. Reprint of: Effects of cold deformation, electron irradiation and extrusion on deuterium desorption behavior in Zr-1%Nb alloy

    Science.gov (United States)

    Morozov, O.; Mats, O.; Mats, V.; Zhurba, V.; Khaimovich, P.

    2018-01-01

    The present article introduces the data of analysis of ranges of ion-implanted deuterium desorption from Zr-1% Nb alloy. The samples studied underwent plastic deformation, low temperature extrusion and electron irradiation. Plastic rolling of the samples at temperature ∼300 K resulted in plastic deformation with the degree of ε = 3.9 and the formation of nanostructural state with the average grain size of d = 61 nm. The high degree of defectiveness is shown in thermodesorption spectrum as an additional area of the deuterium desorption in the temperature ranges 650-850 K. The further processing of the sample (that had undergone plastic deformation by plastic rolling) with electron irradiation resulted in the reduction of the average grain size (58 nm) and an increase in borders concentration. As a result the amount of deuterium desorpted increased in the temperature ranges 650-900 K. In case of Zr-1% Nb samples deformed by extrusion the extension of desorption area is observed towards the temperature reduction down to 420 K. The formation of the phase state of deuterium solid solution in zirconium was not observed. The structural state behavior is a control factor in the process of deuterium thermodesorption spectrum structure formation with a fixed implanted deuterium dose (hydrogen diagnostics). It appears as additional temperature ranges of deuterium desorption depending on the type, character and defect content.

  9. Mechanistic understanding of irradiation-induced corrosion of zirconium alloys in nuclear power plants: Stimuli, status, and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Ishigure, K.; Nechaev, A.F.; Reznichenko, E.A.; Cox, B.; Lemaignan, C.; Petrik, N.G.

    1990-05-01

    Failures in the basic materials used in nuclear power plants continue to be costly and insidious, despite increasing industry vigilance to catch failures before they degrade safety. For instance, the overall costs to the US industry from materials problems could amount to as much as $10 billion annually. Moreover, estimates indicate that the cost of a pipe failure in a nuclear plant is one hundred times greater than the cost of a similar failure in a coal-fired plant. There are important practical stimuli and much scope for further understanding of the effects of irradiation on Zr-alloys (and other materials used in nuclear installations) by careful experimentation. Moreover, these studies need to address the effect of irradiation on all components of heterogeneous systems: the metal, the oxide and the environment, and especially those processes recurring at the interphases between these components. The present paper is aimed at providing specialists with some systematic information on the subject and with important considerations on the key items for further experimentation.

  10. Non-isothermal irradiation creep of nickel alloys Inconel 706 and PE-16

    International Nuclear Information System (INIS)

    Gilbert, E.R.; Chin, B.A.

    1984-06-01

    The results of in-reactor step temperature change experiments conducted on two nickel alloys, PE-16 and Inconel 706, were evaluated to determine the creep behavior under nonisothermal conditions. The effect of the temperature changes was found to be significantly different for the two alloys. Following a step temperature change, the creep rate of PE-16 adjusted to the rate found in isothermal tests at the new temperature. In contrast for Inconel 706, a reduction in temperature from 540 to 425 0 C produced a 300% increase in creep above that measured at 540 0 C in isothermal tests. The response of in-reactor creep in Inconel 706 to temperature changes was attributed to the dissolution of the gamma double-prime phase and subsequent loss of precipitation-strengthening at temperatures below 500 C

  11. Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the

  12. Enhancement of Curie Temperature (T c) and Magnetization of Fe-Ni Invar alloy Through Cu Substitution and with He+2 Ion Irradiation

    Science.gov (United States)

    Khan, Sajjad Ahmad; Ziya, Amer Bashir; Ibrahim, Ather; Atiq, Shabbar; Usman, Muhammad; Ahmad, Naseeb; Shakeel, Muhammad

    2016-04-01

    The magnetic properties of ternary Fe-Ni-Cu invar alloys are affected by ion irradiation, which goes on increasing with increasing ion fluence (Φ), and by increasing Cu content. In the present study, the ions used are He+2 with 2 MeV energy and with 1 × 1013 cm-2, 1 × 1014 cm-2, 5 × 1014 cm-2, 1 × 1015 cm-2 and 5 × 1015 cm-2 fluence (dose) for irradiation purpose. The face centered cubic structure of the alloy was investigated after ion irradiation using x-ray diffraction (XRD) and found unchanged. However, the peaks become broader with increasing ion dose. Additionally, the lattice fluctuations were observed in XRD study. Curie temperature (T c) is also increased after irradiation. Many factors are considered here for the reason for increasing T c, such as the stopping of incident ions, atomic mixing effect at micro scale level owing to ion irradiation, which might change local concentration and ordering already reported in diffuse scattering, and as a result the Fe-Fe interatomic distance and the Fe-Fe coupling are changed. A comparative study shows that the effect of irradiation on T c and magnetization with increasing ion fluence is more distinctive than the addition of Cu.

  13. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    International Nuclear Information System (INIS)

    Singh, B.N.; Edwards, D.J.; Bilde-Soerensen, J.B.

    2004-10-01

    The phenomenon of plastic flow localization in the form of 'cleared' channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels are initiated during post-irradiation deformation has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons. Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron irradiated at 323 and 373K to displacement doses in the range of 0.01 to 0.3 dpa (displacement per atom) and tensile tested at the irradiation temperature. The stress-strain curves clearly indicated the occurrence of a yield drop. The post-deformation microstructural examinations revealed that the channels are formed already in the elastic regime and their density increases with increasing plastic strain. The channels appear to have been initiated at grain boundaries, twin boundaries, at relatively large inclusions and even at the previously formed cleared channels. Even though the channels are produced throughout the whole tensile test, no clear evidence has been found for the operation of Frank-Read sources in the volume between the channels. Channels have been observed to penetrate through annealing twins, in some cases stopping at the opposite twin boundary and in other cases penetrating even through the opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimens activate dislocation sources at the sites of stress concentration at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these

  14. Change of Cr atoms distribution in Fe85Cr15 alloy caused by 250 keV He+ ion irradiation to different doses

    International Nuclear Information System (INIS)

    Dubiel, S.M.; Żukrowski, J.

    2015-01-01

    Highlights: • Effect of He-ion irradiation dose on Fe 85 Cr 15 alloy. • Irradiation-induced clustering of Cr atoms. • Irradiation-caused reorientation of the surface magnetization vector. • Irradiation-caused increase of Fe-site spin-density. - Abstract: Redistribution of Cr atoms in a Fe 85 Cr 15 alloy caused by its irradiation with 250 keV He + ions to different doses, D = 8 ⋅ 10 16 , 16 ⋅ 10 16 and 48 ⋅ 10 16 ions/cm 2 was investigated by means of conversion electrons Mössbauer spectroscopy. The redistribution was expressed in terms of the Warren–Cowley short-range order parameters α 1 , α 2 and α 12 pertaining to the first (1NN), second (2NN) and both i.e. 1NN + 2NN shells, respectively. Clear evidence was found, both for non-irradiated and irradiated samples that the actual distribution of Cr atoms is characteristic of the shell, and for a given shell it depends on the irradiation dose. In particular, α 1 is positive, hence indicates an under population of Cr atoms in 1NN with respect to the random case, α 2 is negative, giving evidence thereby that 2NN is overpopulated by Cr atoms, and α 12 is weakly positive. Under the applied irradiation the number of Cr atoms in both neighbor shells decreased signifying thereby a clustering of Cr atoms. The underlying decrease of Cr concentration within the 1NN–2NN volume around the probe Fe atoms was estimated at 1.5 at.% ranging between 2.1 for the lowest and 0.8 at.% for the highest dose

  15. Effects of fluence and fluence rate of proton irradiation upon magnetism in Fe{sub 65}Ni{sub 35} Invar alloy

    Energy Technology Data Exchange (ETDEWEB)

    Matsushita, Masafumi, E-mail: matsushita.masafumi.me@ehime-u.ac.jp [Department of Mechanical Engineering, Ehime University, 3-Bunkyocho, Matsuyama 790-8977 (Japan); Wada, Hideki [Department of Mechanical Engineering, Ehime University, 3-Bunkyocho, Matsuyama 790-8977 (Japan); Matsushima, Yasushi [Department of Physics, Okayama University, 2-naka-tsushima, Kitaku, Okayama 700-8530 (Japan)

    2015-11-15

    Curie temperature, T{sub C}, of the Fe-Ni Invar alloys increase due to irradiation with electron and some kinds of ions. In this study, proton irradiation effects upon magnetism in an Fe{sub 65}Ni{sub 35} alloy have been investigated. It is found that the increment of T{sub C,} ∆T{sub C}, increases with increasing fluence. The magnetic hysteresis curve of the alloy was found to be unaffected by irradiation. Comparing ∆T{sub C} and the calculated energy transfer from the ions to the sample, it seemed that ∆T{sub C} was found to be related to the number of vacancies formed in nuclear collision events. In addition, ∆T{sub C} was influenced by the fluence rate, i.e., the deposited energy per unit time. - Highlights: • Proton irradiation effect on T{sub C} of Fe{sub 65}Ni{sub 35} was investigated. • Increment of T{sub C}, ∆T{sub C}, was confirmed in ion passed through and stopped samples. • The relationships among ∆T{sub C} and the deposited energy and vacancies were discussed. • It was reasonable to consider that ∆T{sub C} was related to the number of vacancies. • ∆T{sub C} was influenced by fluence rate, i.e. the energy deposition rate.

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

    Energy Technology Data Exchange (ETDEWEB)

    Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

    2016-11-15

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

  17. Effect of fast neutron irradiation upon the omega transformation process in zirconium--niobium alloys

    International Nuclear Information System (INIS)

    Bremer, B.W.

    1974-01-01

    The effect of fast neutrons (E greater than 1 MeV) upon the beta-omega transformation was investigated. Irradiation-produced vacancies promoted the omega transformation by migrating to the regions of high compressive stress associated with the one-dimensional omega embryos, thereby allowing rearrangement of the strain fields. This rearrangement allows the omega embryos to attain large sizes. Growth of these embryos is diffusion controlled. However, irradiation produced no increase in growth rate. It is concluded the vacancies are effectively trapped by these strain fields even at the aging temperature, 400 0 C. The omega hardening mechanism is shown to be related solely to lattice misfit, independent of irradiation, and to saturate when the magnitude of the strain causes a breakdown of the coherent interface, thereby creating one or two interfacial dislocations. Aging at 400 0 C results in alpha growth into the interomega beta phase, producing an additional hardness increase additive to that resulting from the omega phase. At higher aging temperatures 500 0 C, the omega is rapidly replaced by alpha. The alpha microstructure consists of ultra-fine grains, 1000 A, each composed of one 12 interrelated crystallographic variants. Fast neutron irradiation has no effect upon the omega metastable equilibrium phase diagram

  18. Post irradiated microstructural characterization of Zr–1Nb alloy by X ...

    Indian Academy of Sciences (India)

    Abstract. Zr–1Nb samples were irradiated with 116 MeV O5+ ions at different doses ranging from 5 × 1017 to 8 ×. 1018 O5+/m2. X-ray diffraction line profile analysis was performed to characterize the microstructural parameters of these samples. Average domain size, microstrain and dislocation density were estimated as a ...

  19. Neutron irradiation effects on magnetic properties of iron-nickel Invar alloys

    International Nuclear Information System (INIS)

    Morita, H.; Tanji, Y.; Hiroyoshi, H.; Nakagawa, Y.

    1983-01-01

    The Curie temperature of fcc Fe-Ni containing 30-50% Ni is reaised by neutron irradiation, although no appreciable change is detected in the X-ray diffraction pattern. These results are related to a tendency to two-phase separation of the fcc phase. (orig.)

  20. Vacancies supersaturation induced by fast neutronn irradiation in FeNi alloys

    International Nuclear Information System (INIS)

    Lucki, G.; Watanabe, S.; Chambron, W.; Verdoni, J.

    1976-01-01

    Isothermal annealings have been performed between 400 and 555 0 C with and without fast neutron (1 MeV) irradiation. Pure FeNi (50-50 at %) was irradiated in the Melousine reactor in Grenoble and FeNiMO (50-50 at % + 50 ppm.) in the IEAR 1 reactor at the Instituto de Energia Atomica in Sao Paulo. The toroidal shaped specimens were fabricated from Johnson Mathey zone refined ingots and were initially annealed at 800 0 C during 1 h in hydrogen atmosphere and then slowly cooled (4 h) inside the furnace. Magnetic After Effect Measurements (MAE) permitted the evaluation of activation energies during fast neutron irradiation (1.54eV) and without irradiation (3.14eV) for pure FeNi and respectively (1.36eV) and 2.32eV) for FeNiMO. Since the time constants of relaxation process are inversely proportional to the vacancies comcentration a quantitative evaluation of vacancies supersaturation was made it decreases from value 700 at 410 0 C to the value 40 at 190 0 C for pure FeNi and from 765 to 121 for FeNiMO in the same temperature range

  1. Microstructural evolutions of zirconium alloys under irradiation. Link with the irradiation growth phenomenon; Evolutions microstructurales des alliages de zirconium sous irradiation. Liens avec le phenomene de croissance

    Energy Technology Data Exchange (ETDEWEB)

    Simonot, C.

    1995-07-18

    This study deals with the irradiation-induced growth and microstructural evolutions of Zircaloy-4 type materials (ZrSn{sub 1.2-1.7} Fe{sub 0.18-0.24} Cr{sub 0.07-0.13} O{sub 0.09-0.15}), used as cladding and guide-tubes in PWR`s fuel assemblies. The main objective was to obtain a better understanding of the growth acceleration which may occur at high doses for the recrystallized metallurgical state. The elongation values of stress-free tubes irradiated at 400 deg in experimental reactors give clear indication of accelerated growth after a critical dose. Microstructural investigations reveal some large basal plane dislocation loops with vacancy character, which is an unexpected defect configuration for an hexagonal material with a c/a ratio less than the ideal value. In addition, a significant redistribution of iron and chromium solute elements comes from the dissolution of the initial Zr(Fe,Cr){sub 2} phases. In a guide-tube irradiated to high dose at 320 deg in a power reactor, a large density of these c-component loops was also observed in coincidence with a large iron re-solution due to the progressive partial amorphization of Laves phases. By contrast, as long as a negligible amount of iron is available in the matrix (start of progressive) amorphization at 350 deg or complete amorphization without any chemical change at 280 deg, only prism plane loops with interstitial and vacancy character are observed and the steady-state growth rate is low. A mechanism taking into account the Diffusional Anisotropy Difference of the radiation induced point defects seems to be the most suitable to explain the correlations between microstructural evolutions and growth rates. However it does not allow to predict the dose necessary for the formation of the basal plane loops responsible for the growth acceleration. (Abstract Truncated)

  2. Gas accumulation at grain boundaries during 800 MeV proton irradiation of aluminium and aluminium-alloys

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Horsewell, Andy; Sommer, W. F.

    1986-01-01

    ) showed a complete absence of voids or bubbles in the grain interiors of the aluminium and the aluminium-alloys. Bubbles were clearly visible by TEM at grain boundaries in pure Al and the AlMg3 alloy; but bubbles were not visible in the Al6061 alloy. The bubble density in the AlMg3 alloy was considerably...

  3. Properties of gallium arsenide alloyed with Ge and Se by irradiation in nuclear reactor thermal column

    International Nuclear Information System (INIS)

    Kolin, N.G.; Osvenskij, V.B.; Tokarevskij, V.V.; Kharchenko, V.A.; Ievlev, S.M.

    1985-01-01

    Dependences of electrophysical properties as well as lattice unit spacing and density of nuclear-alloyed gallium arsenide on the fluence of reactor neutrons and heat treatment are investigated. Neutron radiation of gallium arsenide with different energy spectra is shown to differently affect material properties. Fast neutrons make the main contribution to defect formation. Concentration of compensating acceptor defects formed under GaAs radiation in a thermal column practically equals concentration of introduced donor impurities. Radiation defects of acceptor type are not annealed in the material completely even at 900-1000 deg C

  4. Development of nondestructive inspection techniques for measurement of void swelling in irradiated microscopy disks

    International Nuclear Information System (INIS)

    Sagisaka, M.; Isobe, Y.; Garner, F.A.; Fujita, S.; Okita, T.

    2011-01-01

    Results are presented of resistivity change measurements made on a model Fe-Cr-Ni-Ti-C austenitic alloy irradiated in the Fast Flux Test Facility to doses ranging from 1.87 to 67.8 dpa. Two different electrical resistivity measurement systems were developed to overcome problems associated with examination of small microscopy specimens in order to investigate small changes in resistivity arising from voids and other radiation-induced microstructural features. A correlation is shown between resistivity changes arising primarily from void swelling. However, contributions arising probably from radiation-induced redistribution and perhaps precipitation of carbon and titanium can dominate the resistivity at low dpa levels.

  5. Study of the precipitation and of the hardening microscopic mechanisms under irradiation in dilute ferritic alloys; Etude de la precipitation et des mecanismes microscopiques de durcissement sous irradiation dans des alliages ferritiques dilues

    Energy Technology Data Exchange (ETDEWEB)

    Mathon, M.H

    1995-07-01

    The copper precipitation plays a significant role in the embrittlement process of reactor vessel steels under neutron irradiation at 300 deg C. In order to understand the copper precipitation mechanisms, we have studied model ferritic binary FeCu and ternary alloys FeCuX (X=Mn,Ni, Cr, P). These materials have been either Irradiated with 2.5 MeV electrons In the 175-360 deg C temperature range or thermal aged at 500 deg C. The evolution of materials has been followed by resistivity measurements under irradiation, by small angle neutron scattering and by Vickers microhardness measurements. We have shown the similarity of copper precipitation under thermally ageing at 500 deg C and electron Irradiation at 300 deg C, in FeCu{sub 1,34%}. This result confirms that the main effect of electronic irradiation is to accelerate precipitation. Nevertheless, we have observed that irradiation induces an additional contribution to hardening attributed to point defect clusters. Concerning the ternary alloys, we observed that at 300 deg C the addition of a third element has no significant effect on the copper precipitation kinetic under irradiation but that at lower temperature manganese slows down precipitation kinetic. In order to reproduce the experimental results obtained on FeCu{sub 1,34%} by using a cluster kinetics model, we have to suppose that the precipitation is heterogeneous and controlled by interface reactions for the small size clusters. In addition, neutron or electron irradiated industrial steels have been studied by small angle neutron scattering. The results revealed the presence of nano-metric solute clusters which contain few copper atoms and which are not linked to the formation of displacement cascades. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Meric de Bellefon, G., E-mail: mericdebelle@wisc.edu [University of Wisconsin-Madison (United States); Duysen, J.C. van [EDF R& D (France); University of Tennessee-Knoxville (United States); Unité Matériaux et Transformation (UMET) CNRS, Université de Lille (France)

    2016-07-15

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

  7. The Formation of Multipoles during the High-Temperature Creep of Austenitic Stainless Steels

    DEFF Research Database (Denmark)

    Howell, J.; Nielsson, O.; Horsewell, Andy

    1981-01-01

    It is shown that multipole dislocation configurations can arise during power-law creep of certain austenitic stainless steels. These multipoles have been analysed in some detail for two particular steels (Alloy 800 and a modified AISI 316L) and it is suggested that they arise either during...

  8. Estimation of the solubility limit of Cr in Fe at 300 oC from small-angle neutron scattering in neutron-irradiated Fe-Cr alloys

    International Nuclear Information System (INIS)

    Bergner, F.; Ulbricht, A.; Heintze, C.

    2009-01-01

    The solubility limit of Cr in Fe (α-Fe-Cr) at low temperatures is a matter of debate. We report a direct estimation of the solubility limit at 300 o C from small-angle neutron scattering (SANS) data obtained for neutron-irradiated Fe-Cr alloys. The SANS results indicate that the equilibrium concentration of α' was reached via irradiation-enhanced diffusion. The solubility limit was estimated using an iterative approach based on the SANS invariant and the lever rule of phase equilibrium.

  9. Diagnostic experimental results on the hydrogen embrittlement of austenitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Gavriljuk, V.G.; Shivanyuk, V.N.; Foct, J

    2003-03-14

    Three main available hypotheses of hydrogen embrittlement are analysed in relation to austenitic steels based on the studies of the hydrogen effect on the interatomic bonds, phase transformations and microplastic behaviour. It is shown that hydrogen increases the concentration of free electrons, i.e. enhances the metallic character of atomic interactions, although such a decrease in the interatomic bonding cannot be a reason for brittleness and rather assists an increased plasticity. The hypothesis of the critical role of the hydrogen-induced {epsilon} martensite was tested in the experiment with the hydrogen-charged Si-containing austenitic steel. Both the fraction of the {epsilon} martensite and resistance to hydrogen embrittlement were increased due to Si alloying, which is at variance with the pseudo-hydride hypothesis. The hydrogen-caused early start of the microplastic deformation and an increased mobility of dislocations, which are usually not observed in the common mechanical tests, are revealed by the measurements of the strain-dependent internal friction, which is consistent with the hypothesis of the hydrogen-enhanced localised plasticity. An influence of alloying elements on the enthalpy E{sub H} of hydrogen migration in austenitic steels is studied using the temperature-dependent internal friction and a correlation is found between the values of E{sub H} and hydrogen-caused decrease in plasticity. A mechanism for the transition from the hydrogen-caused microplasticity to the apparent macrobrittle fracture is proposed based on the similarity of the fracture of hydrogenated austenitic steels to that of high nitrogen steels.

  10. Dynamic recrystallization in friction surfaced austenitic stainless steel coatings

    Energy Technology Data Exchange (ETDEWEB)

    Puli, Ramesh, E-mail: rameshpuli2000@gmail.com; Janaki Ram, G.D.

    2012-12-15

    Friction surfacing involves complex thermo-mechanical phenomena. In this study, the nature of dynamic recrystallization in friction surfaced austenitic stainless steel AISI 316L coatings was investigated using electron backscattered diffraction and transmission electron microscopy. The results show that the alloy 316L undergoes discontinuous dynamic recrystallization under conditions of moderate Zener-Hollomon parameter during friction surfacing. - Highlights: Black-Right-Pointing-Pointer Dynamic recrystallization in alloy 316L friction surfaced coatings is examined. Black-Right-Pointing-Pointer Friction surfacing leads to discontinuous dynamic recrystallization in alloy 316L. Black-Right-Pointing-Pointer Strain rates in friction surfacing exceed 400 s{sup -1}. Black-Right-Pointing-Pointer Estimated grain size matches well with experimental observations in 316L coatings.

  11. Comparison of damaging behavior of oxide scales grown on austenitic stainless steels using tensile test and cyclic thermogravimetry

    OpenAIRE

    Fedorova, Elena N.; Braccini, Muriel; Parry, Valérie; Pascal, Celine; Mantel, Marc; Roussel-Dherbey, Francine; Oquab, Djar; Wouters, Yves; Monceau, Daniel

    2016-01-01

    Two austenitic stainless steels, AISI 304L and AISI 303, were submitted to cyclic oxidation and to staticmechanical loading after isothermal oxidation at 1000◦C. Alloy 303 contains ten times more S than 304Land some Mn addition. During the steel process, it formed manganese sulfides that lead to the formationof a less resistant oxide scale. Both alloys showed similar behavior during thermal cycling but breakawayoxidation and intensive spallation occurred much sooner for alloy 303 than for all...

  12. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    International Nuclear Information System (INIS)

    Jones, R.H.; Gelles, D.S.

    1993-08-01

    Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry

  13. In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy

    International Nuclear Information System (INIS)

    Zhang, Xuan; Li, Meimei; Park, Jun-Sang; Kenesei, Peter; Almer, Jonathan

    2016-01-01

    The effect of neutron irradiation on tensile deformation of a Fe-9wt.%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tensile tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the