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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  4. 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. The effect of alloying elements on the vacancy defect evolution in electron-irradiated austenitic Fe-Ni alloys studied by positron annihilation

    Energy Technology Data Exchange (ETDEWEB)

    Druzhkov, A.P. [Institute of Metal Physics, Ural Branch RAS, 18 Kovalevskaya St., 620041 Ekaterinburg (Russian Federation)], E-mail: druzhkov@imp.uran.ru; Perminov, D.A.; Davletshin, A.E. [Institute of Metal Physics, Ural Branch RAS, 18 Kovalevskaya St., 620041 Ekaterinburg (Russian Federation)

    2009-01-31

    The vacancy defect evolution under electron irradiation in austenitic Fe-34.2 wt% Ni alloys containing oversized (aluminum) and undersized (silicon) alloying elements was investigated by positron annihilation spectroscopy at temperatures between 300 and 573 K. It is found that the accumulation of vacancy defects is considerably suppressed in the silicon-doped alloy. This effect is observed at all the irradiation temperatures. The obtained results provide evidence that the silicon-doped alloy forms stable low-mobility clusters involving several Si and interstitial atoms, which are centers of the enhanced recombination of migrating vacancies. The clusters of Si-interstitial atoms also modify the annealing of vacancy defects in the Fe-Ni-Si alloy. The interaction between small vacancy agglomerates and solute Al atoms is observed in the Fe-Ni-Al alloy under irradiation at 300-423 K.

  8. 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.

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

    International Nuclear Information System (INIS)

    Dubuisson, P.; Levy, V.; Seran, J.L.

    1987-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  11. Effect of Ti additions on the swelling of electron irradiated austenitic steels and Ni alloys

    International Nuclear Information System (INIS)

    Gilbon, D.; Didout, G.; Le Naour, L.; Levy, V.

    1979-01-01

    It has been shown that titanium is a beneficial additive for the swelling of austenitic steels. The amplitude of the effects observed depends much on the nature and concentration of the other additives in the austenitic matrix [fr

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

    International Nuclear Information System (INIS)

    Huang, F.H.

    1992-02-01

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

  13. New developments in irradiation-induced microstructural evolution of austenitic alloys and their consequences on mechanical properties

    International Nuclear Information System (INIS)

    Garner, F.A.; Brager, H.R.; Hamilton, M.L.; Dodd, R.A.; Porter, D.L.

    1985-01-01

    A survey is presented of recent development in the study of radiation-induced changes in the microstructure of austenitic structural alloys that occur in fission reactors. The associated macroscopic consequences of these changes on both mechanical properties and dimensional stability are also reviewed. It is anticipated that some changes will occur in these phenomena as a result of the differences inherent in fission and fusion neutron spectra, but relevant data obtained to date do not indicate that the effects of helium and several other transmutation-related changes will be large. 78 refs., 12 figs

  14. The influence of nickel content on microstructures of Fe-Cr-Ni austenitic alloys irradiated with nickel ions

    International Nuclear Information System (INIS)

    Muroga, T.; Yoshida, N.; Garner, F.A.

    1990-11-01

    The objectives of this effort is to identify the mechanisms involved in the radiation-induced evolution of microstructure in materials intended for fusion applications. The results of this study are useful in interpreting the results of several other ongoing experiments involving either spectral or isotopic tailoring to study the effects of helium on microstructure evolution. Ion-irradiated Fe-15Cr-XNi (X = 20, 35, 45, 60, 75) ternary alloys and a 15Cr-85Ni binary alloy were examined after bombardment at 675 degree C and compared to earlier observations made on these same alloys after irradiation in EBR-II at 510 or 538 degree C. The response of the ion-irradiated microstructures to nickel content appears to be very consistent with that of neutron irradiation even though there are four orders of magnitude difference in displacement rate and over 200 degree C difference in temperature. It appears that the transition to higher rates of swelling during both types of irradiation is related to the operation of some mechanisms that is not directly associated with void nucleation. 6 refs., 8 figs

  15. Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in oxygenated high temperature water was studied. The IASCC failure has been considered as a degradation phenomenon potential not only in the present light water reactors but rather common in systems where the materials are exposed simultaneously to radiation and water environments. In this study, effects of the material and environmental factors on the IASCC of austenitic stainless steels were investigated in order to understand the underlying mechanism. The following three types of materials were examined: a series of model alloys irradiated at normal water-cooled research reactors (JRR-3M and JMTR), the material irradiated at a spectrally tailored mixed-spectrum research reactor (ORR), and the material sampled from a duct tube of a fuel assembly used in the experimental LMFBR (JOYO). Post-irradiation stress corrosion cracking tests in a high-temperature water, electrochemical corrosion tests, etc., were performed at hot laboratories. Based on the results obtained, analyses were made on the effects of alloying/impurity elements, irradiation/testing temperatures and material processing, (i.e., post-irradiation annealing and cold working) on the cracking behavior. On the basis of the analyses, possible remedies against IASCC in the core internals were discussed from viewpoints of complex combined effects among materials, environment and processing factors. (author). 156 refs.

  16. Austenitic alloys Fe-Ni-Cr dominating

    International Nuclear Information System (INIS)

    Gibson, R.C.; Korenko, M.K.

    1980-01-01

    Austenitic alloy essentially comprising 42 to 48% nickel, 11 to 13% chromium, 2.6 to 3.4% niobium, 0.2 to 1.2% silicon, 0.5 to 1.5% vanadium, 2.6 to 3.4% molybdenum, 0.1 to 0.3% aluminium, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06% zirconium, the balance being iron. The characteristic of this alloy is a conventional elasticity limit to within 2% of at least 450 MPa, with a maximum tensile strength of at least 500 MPa at a test temperature of 650 0 C after immersion annealing at 1038 0 C and 30% hardening. To this effect the invention concerns Ni-Cr-Fe high temperature alloys possessing excellent mechanical strength characteristics, that can be obtained with lower levels of nickel and chromium than those used in alloys of this kind in the present state of the technique, a higher amount of niobium than in the previous alloys and with the addition of 0.5 to 1.5% vanadium [fr

  17. Austin: austenitic steel irradiation E 145-02 Irradiation Report

    International Nuclear Information System (INIS)

    Genet, F.; Konrad, J.

    1987-01-01

    Safety measures for nuclear reactors require that the energy which might be liberated in a reactor core during an accident should be contained within the reactor pressure vessel, even after very long irradiation periods. Hence the need to know the mechanical properties at high deformation velocity of structure materials that have received irradiation damage due to their utilization. The stainless steels used in the structures of reactors undergo damage by both thermal and fast neutrons, causing important changes in the mechanical properties of these materials. Various austenitic steels available as structural materials were irradiated or are under irradiation in various reactors in order to study the evolution of the mechanical properties at high deformation velocity as a function of the irradiation damage rate. The experiment called AUSTIN (AUstenitic STeel IrradiatioN) 02 was performed by the JRC Petten Establishment on behalf of Ispra in support of the reactor safety programme

  18. Numerical analysis of stress effects on Frank loop evolution during irradiation in austenitic Fe-Cr-Ni alloy

    International Nuclear Information System (INIS)

    Tanigawa, H.; Katoh, Y.; Kohyama, A.

    1995-01-01

    Effects of applied stress on early stages of interstitial type Frank loop evolution were investigated by both numerical calculation and irradiation experiments. The final objective of this research is to propose a comprehensive model of complex stress effects on microstructural evolution under various conditions. In the experimental part of this work, the microstructural analysis revealed that the differences in resolved normal stress caused those in the nucleation rates of Frank loops on {111} crystallographic family planes, and that with increasing external applied stress the total nucleation rate of Frank loops was increased. A numerical calculation was carried out primarily to evaluate the validity of models of stress effects on nucleation processes of Frank loop evolution. The calculation stands on rate equations which describe evolution of point defects, small points defect clusters and Frank loops. The rate equations of Frank loop evolution were formulated for {111} planes, considering effects of resolved normal stress to clustering processes of small point defects and growth processes of Frank loops, separately. The experimental results and the predictions from the numerical calculation qualitatively coincided well with each other. (orig.)

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

    International Nuclear Information System (INIS)

    Taehtinen, S.; Pyykkoenen, M.; Singh, B.N.; Toft, P.

    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.)

  20. 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

  1. Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

    Science.gov (United States)

    Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeob; Kwon, Junhyun; Shin, Chansun

    2017-09-01

    We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni-Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

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

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. Microstructural investigations of fast reactor irradiated austenitic and ferritic-martensitic stainless steel fuel cladding

    International Nuclear Information System (INIS)

    Agueev, V.S.; Medvedeva, E.A.; Mitrofanova, N.M.; Romanueev, V.V.; Tselishev, A.V.

    1992-01-01

    Electron microscopy has been used to characterize the microstructural changes induced in advanced fast reactor fuel claddings fabricated from Cr16Ni15Mo3NbB and Cr16Ni15Mo2Mn2TiVB austenitic stainless steels in the cold worked condition and Cr13Mo2NbVB ferritic -martensitic steel following irradiation in the BOR-60, BN-350 and BN-600 fast reactors. The data are compared with the results obtained from a typical austenitic commercial cladding material, Cr16Ni15Mo3Nb, in the cold worked condition. The results reveal a beneficial effect of boron and other alloying elements in reducing void swelling in 16Cr-15Ni type austenitic steels. The high resistance of ferritic-martensitic steels to void swelling has been confirmed in the Cr13Mo2NbVB steel. (author)

  9. Study of irradiation damage structures in austenitic stainless steels

    International Nuclear Information System (INIS)

    Hamada, Shozo

    1997-08-01

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

  10. Study of irradiation damage structures in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    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. The

  12. Alloys under irradiation

    International Nuclear Information System (INIS)

    Martin, G.; Bellon, P.; Soisson, F.

    1997-01-01

    During the last two decades, some effort has been devoted to establishing a phenomenology for alloys under irradiation. Theoretically, the effects of the defect supersaturation, sustained defect fluxes and ballistic mixing on solid solubility under irradiation can now be formulated in a unified manner, at least for the most simple cases: coherent phase transformations and nearest-neighbor ballistic jumps. Even under such restrictive conditions, several intriguing features documented experimentally can be rationalized, sometimes in a quantitative manner and simple qualitative rules for alloy stability as a function of irradiation conditions can be formulated. A quasi-thermodynamic formalism can be proposed for alloys under irradiation. However, this point of view has limits illustrated by recent computer simulations. (orig.)

  13. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-15

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

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

    International Nuclear Information System (INIS)

    Volgin, Alexandre

    2012-01-01

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

  16. 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.))

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

    International Nuclear Information System (INIS)

    Hocquellet, Dominique

    1984-01-01

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

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

    Science.gov (United States)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

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

  19. 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

  20. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Study of the effects of austenitizing and tempering heat treatments on the alloy HT-9

    International Nuclear Information System (INIS)

    Redmon, J.W.

    1982-01-01

    This paper investigates the potential use of the ferritic alloy Sandvik HT-9 (12 Cr - 1 Mo) as an alternative to stainless steels used in high-neutron-fluence environments. The neutron radiation influences embrittlement where the impact-energy versus test-temperature curve is seen displaced to the right. As a result, commercially effective solutioning and tempering processes are needed to suppress this effect in the pre-irradiated condition. The effects of austenitizing treatments on the impact energy of HT-9 were identified. 18 figures, 6 tables

  4. 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.

  5. Austenite Grain Growth Behavior of AISI 4140 Alloy Steel

    Directory of Open Access Journals (Sweden)

    Lin Wang

    2013-01-01

    Full Text Available AISI 4140 alloy steel is widely applied in the manufacture of various parts such as gears, rams, and spindles due to its good performance of strength, toughness, and wear resistance. The former researches most focused on its deformation and recrystallization behaviors under high temperature. However, the evolution laws of austenite grain growth were rarely studied. This behavior also plays an important role in the mechanical properties of parts made of this steel. In this study, samples are heated to a certain temperature of 1073 K, 1173 K, 1273 K, and 1373 K at a heating rate of 5 K per second and hold for different times of 0 s, 120 s, 240 s, 360 s, and 480 s before being quenched with water. The experimental results suggest that the austenite grains enlarge with increasing temperature and holding time. A mathematical model and an application developed in Matlab environment are established on the basis of previous works and experimental results to predict austenite grains size in hot deformation processes. The predicted results are in good agreement with experimental results which indicates that the model and the application are reliable.

  6. 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.)

  7. Study of irradiation effects in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Etienne, A. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Material Department, University of California, Santa Barbara (United States); Pareige, P.; Radiguet, B. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Cunningham, N.J.; Odette, G.O. [Material Department, University of California, Santa Barbara (United States); Pokor, C. [EDF RD, departement MMC, site des Renardieres, Moret-sur-Loing (France)

    2011-07-01

    Chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after seventeen years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A very high number density (6.10{sup 23} m{sup -3}) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. In order to bring complementary experimental results and to determine the mechanism of formation of these Ni-Si nano-clusters, Fe{sup 5+} ion irradiations have been performed on a 316 austenitic stainless steel. As after neutron irradiation, the formation of solute enriched features is observed. Linear features and two kinds of clusters, rounded and torus shaped, are present. Considering that solute enriched features are probably formed by radiation induced segregation on point defect sinks, these different shapes are due to the nature of the sinks where segregation occurs. (authors)

  8. 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.)

  9. Numerical analysis of stress effects on Frank loop evolution during irradiation in austenitic Fe&z.sbnd;Cr&z.sbnd;Ni alloy

    Science.gov (United States)

    Tanigawa, Hiroyasu; Katoh, Yutai; Kohyama, Akira

    1995-08-01

    Effects of applied stress on early stages of interstitial type Frank loop evolution were investigated by both numerical calculation and irradiation experiments. The final objective of this research is to propose a comprehensive model of complex stress effects on microstructural evolution under various conditions. In the experimental part of this work, the microstructural analysis revealed that the differences in resolved normal stress caused those in the nucleation rates of Frank loops on {111} crystallographic family planes, and that with increasing external applied stress the total nucleation rate of Frank loops was increased. A numerical calculation was carried out primarily to evaluate the validity of models of stress effects on nucleation processes of Frank loop evolution. The calculation stands on rate equuations which describe evolution of point defects, small points defect clusters and Frank loops. The rate equations of Frank loop evolution were formulated for {111} planes, considering effects of resolved normal stress to clustering processes of small point defects and growth processes of Frank loops, separately. The experimental results and the predictions from the numerical calculation qualitatively coincided well with each other.

  10. 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.)

  11. Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

    International Nuclear Information System (INIS)

    Byun, T.S.; Farrell, K.; Lee, E.H.; Hunn, J.D.; Mansur, L.K.

    2001-01-01

    Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels

  12. 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.

  13. Effects of irradiation on the fracture behavior of austenitic stainless steels

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Stiegler, J.O.; Holmes, J.J.

    1977-01-01

    Fracture in irradiated materials occurs by mechanisms which occur in unirradiated materials in addition to mechanisms related to irradiation phenomena. The paper examines radiation effects in austenitic stainless steels for use as core structural materials in fast breeder reactors

  14. 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.

  15. Ion irradiation-induced precipitation of Cr23C6 at dislocation loops in austenitic steel

    International Nuclear Information System (INIS)

    Jin, Shuoxue; Guo, Liping; Luo, Fengfeng; Yao, Zhongwen; Ma, Shuli; Tang, Rui

    2013-01-01

    The irradiation-induced precipitates in argon ion-irradiated austenitic stainless steel at 550 °C were examined via transmission electron microscopy. The selected-area electron diffraction patterns of precipitates indicated unambiguously that the precipitates were Cr 23 C 6 carbides. It was observed directly for the first time that irradiation-induced Cr 23 C 6 precipitates formed at dislocation loops in austenitic stainless steel, and coarsened with increasing irradiation dose.

  16. ODS alloys for structures subjected to irradiation

    International Nuclear Information System (INIS)

    Carlan, Y. de

    2010-01-01

    ODS (oxide-dispersion-strengthened) materials are considered for cladding purposes for the fourth-generation sodium-cooled fast reactors. ODS materials afford many benefits. Indeed, these high-performance materials combine, at the same time, remarkable mechanical strength, in hot conditions, and outstanding irradiation behavior. New ODS steel grades, exhibiting better performance levels than the last-generation austenitic steels, afford not only negligible swelling under irradiation, owing to their 'ferritic' body-centered cubic structure - by contrast to austenitic grades, which feature a face-centered cubic structure - but equally outstanding creep properties, owing to the nano-reinforcements present in the matrix. ODS materials are obtained by powder metallurgy, the first fabrication step involves co-grinding a metal powder together with yttrium oxide (Y 2 O 3 ) powder. At this stage, an iron oxide may also be added, or an yttrium-rich intermetallic compound in order to provide the amounts of yttrium, and oxygen required for the formation of nano-oxides. The metal powder consists of a powder pre-alloyed to the chemical composition of the desired material. Once the powder has been obtained, consolidation of the ODS materials is achieved either by hot extrusion, or by hot isostatic pressing. (A.C.)

  17. High temperature ductility of austenitic alloys exposed to thermal neutrons

    International Nuclear Information System (INIS)

    Watanabe, K.; Kondo, T.; Ogawa, Y.

    1982-01-01

    Loss of high temperature ductility due to thermal neutron irradiation was examined by slow strain rate test in vacuum up to 1000 0 C. The results on two heats of Hastelloy alloy X with different boron contents were analyzed with respect to the influence of the temperatures of irradiation and tensile tests, neutron fluence and the associated helium production due to nuclear transmutation reaction. The loss of ductility was enhanced by increasing either temperature or neutron fluence. Simple extrapolations yielded the estimated threshold fluence and the end-of-life ductility values at 900 and 1000 0 C in case where the materials were used in near-core regions of VHTR. The observed relationship between Ni content and the ductility loss has suggested a potential utilization of Fe-based alloys for seathing of the neutron absorber materials

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

  20. 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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

    Microstructural evolution in austenitic stainless steels subjected to displacement damage at high temperature is strongly influenced by the interaction between helium atoms and second phase particles. Cavity nucleation occurs by the trapping of helium at partially coherent particle-matrix interfaces. The recent precipitate point defect collector theory describes the more rapid growth of precipitate-attached cavities compared to matrix cavities where the precipitate-matrix interface collects point defects to augment the normal point deflect flux to the cavity. Data are presented which support these ideas. It is shown that during nickel ion irradiation of a titanium-modified stainless steel at 675 0 C the rate of injection of helium has a strong effect on the total swelling and also on the nature and distribution of precipitate phases. (orig.)

  2. Microstructural stability of austenitic stainless steels on exposure to irradiation and elevated temperatures

    International Nuclear Information System (INIS)

    Parameswaran, P.; Radhika, M.; Saroja, S.; Vijayalakshmi, M.; Nanda Gopal, M.

    2011-01-01

    Cold worked 316 stainless steels employed as core material in fast reactors on exposure to neutron irradiation to 40 dpa at ∼ 450 deg C have resulted in microstructural changes in terms of formation of voids and extensive precipitation of carbides, eta phase and nickel silicides. As a consequence there is degradation in the mechanical properties of the material, particularly ductility. In order to achieve higher burnup it is essential to find better materials, which would exhibit less void swelling and retain the microstructure over long radiation doses. Accordingly alloy D9 with appropriate modifications of Ni and Cr content with Ti additions has been developed. Further modification of alloy D9 with respect to minor alloying additions namely Si and P is being studied, in order to enhance the radiation resistance for extending the service life of components. The effectiveness of these elements can be achieved if and only if they are retained in solution over long time of exposure at high temperatures and irradiation. Therefore, the thermal stability of the newly developed improved D9 alloys, with a constant Ti:C ratio and different levels of Si and P has been studied with respect to microstructural evolution and its influence on the mechanical properties. Thermal aging behavior of the alloy with varying titanium contents at elevated temperatures was also studied in detail to identify the optimum alloying levels. The alloys in the 20% cold worked condition exhibit austenitic grains interspersed with bands of fine cold worked grains. On aging in the temperature range of 873-1073K for various durations upto two years the alloy showed the presence of different phases such as M 23 C 6 , intermetallics and TiC whose quantity varies with temperature. The hardness values showed a trend of an initial increase in all the alloys but at longer times the hardness either showed saturation or a decrease followed by saturation. The microstructural parameters like grain size and

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

    International Nuclear Information System (INIS)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X.; Zheng, W.; Guzonas, D.A.

    2012-01-01

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

  4. Mechanisms of irradiation assisted stress corrosion cracking in austenitic stainless steels

    International Nuclear Information System (INIS)

    Was, G.S.; Busby, G.T.

    2004-01-01

    Full text of publication follows: Service and laboratory experience have shown that irradiation enhances the stress corrosion cracking of austenitic alloys in high temperature water. The degree of irradiation assisted stress corrosion cracking (IASCC) increases with dose as the microstructure undergoes significant changes, including dislocation loop formation, grain boundary segregation and hardening. These changes occur simultaneously and at comparable rates, complicating the attribution of IASCC to specific components of the microstructure. Each of the principal effects of irradiation have been considered as potential causes of IASCC, but the multivariable nature of the problem obscures a definitive determination of the mechanism. Rather, the mechanism of IASCC is more likely due to a combination of factors, some which have not yet been considered. Among these effects is the heterogeneity of deformation caused by the irradiated microstructure, and the interaction of localized deformation bands with grain boundaries. Current understanding and proposed mechanisms of IASCC will be reviewed, and recent progress on the role of heterogeneous deformation on IASCC will be presented. (authors)

  5. Radiation damage simulation studies in the Harwell VEC of selected austenitic and ferritic alloys for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Mazey, D J; Walters, G P; Buckley, S N; Hanks, W; Bolster, D E.J.; Murphy, S M

    1988-07-01

    Three austenitic (316 L, 316-Ti, 316-Nb); four high-nickel (IN 625, IN 706, PE 16, Fe-25Ni-8Cr) and four ferritic (CRM 12, FV 448, FV 607, FI) alloys have been irradiated with 46 MeV Ni or 20 MeV Cr ions in the Harwell VEC to simulated fusion-reactor doses up to 110 dpa (proportional to 10 MW-yr m/sup -2/) at temperatures from 425 to 625/sup 0/C. Gas production rates appropriate to fusion were obtained from a mixed beam of He+H/sub 2/ in the ratio 1:4 He:H with gas/dpa ratios of 13 appm He/dpa and 52 appm H/dpa. The 316 alloys showed irradiation-induced precipitation and swelling as high as 40% in ST 316-Ti after 110 dpa at 625/sup 0/C. Low swelling (e.g. <2% at 110 dpa) was observed in the high-nickel alloys. The ferritic/martensitic alloys showed negligible swelling (e.g. <0.2% in FV 607 after 100 dpa at 475/sup 0/C). The results demonstrate the high swelling behaviour of 316 alloys and the better swelling resistance of high-nickel and ferritic alloys under simulated fusion conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Blondé, R., E-mail: r.j.p.blonde@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Materials Innovation Institute, Mekelweg 2, 2628 CD Delft (Netherlands); Jimenez-Melero, E., E-mail: enrique.jimenez-melero@manchester.ac.uk [Dalton Cumbrian Facility, The University of Manchester, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); Zhao, L., E-mail: lie.zhao@tudelft.nl [Materials Innovation Institute, Mekelweg 2, 2628 CD Delft (Netherlands); Department of Materials Science and Engineering, Delft University of Technology, Mekelweg 2, 2628 CD Delft (Netherlands); Schell, N., E-mail: norbert.schell@hzg.de [Institute of Materials Research, Helmholtz-Zentrum Geesthacht, Max Planck Strasse 1, 21502 Geesthacht (Germany); Brück, E., E-mail: e.h.bruck@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Zwaag, S. van der, E-mail: s.vanderzwaag@tudelft.nl [Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS Delft (Netherlands); Dijk, N.H. van, E-mail: n.h.vandijk@tudelft.nl [Fundamental Aspects of Materials and Energy, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-01-31

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

  7. Strain hardening of cold-rolled lean-alloyed metastable ferritic-austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Papula, Suvi [Aalto University School of Engineering, Department of Mechanical Engineering, P.O. Box 14200, FI-00076 Aalto (Finland); Anttila, Severi [Centre for Advanced Steels Research, University of Oulu, P.O. Box 4200, 90014 Oulu (Finland); Talonen, Juho [Outokumpu Oyj, P.O. Box 245, FI-00181 Helsinki (Finland); Sarikka, Teemu; Virkkunen, Iikka; Hänninen, Hannu [Aalto University School of Engineering, Department of Mechanical Engineering, P.O. Box 14200, FI-00076 Aalto (Finland)

    2016-11-20

    Mechanical properties and strain hardening of two pilot-scale lean-alloyed ferritic-austenitic stainless steels having metastable austenite phase, present at 0.50 and 0.30 volume fractions, have been studied by means of tensile testing and nanoindentation. These ferritic-austenitic stainless steels have high strain-hardening capacity, due to the metastable austenite phase, which leads to an improved uniform elongation and higher tensile strength in comparison with most commercial lean duplex stainless steels. According to the results, even as low as 0.30 volume fraction of austenite seems efficient for achieving nearly 40% elongation. The austenite phase is initially the harder phase, and exhibits more strain hardening than the ferrite phase. The rate of strain hardening and the evolution of the martensite phase were found to depend on the loading direction: both are higher when strained in the rolling direction as compared to the transverse direction. Based on the mechanical testing, characterization of the microstructure by optical/electron microscopy, magnetic balance measurements and EBSD texture analysis, this anisotropy in mechanical properties of the cold-rolled metastable ferritic-austenitic stainless steels can be explained by the elongated dual-phase microstructure, fiber reinforcement effect of the harder austenite phase and the presence and interplay of rolling textures in the two phases.

  8. 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

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

    International Nuclear Information System (INIS)

    Hosoi, Y.; Okazaki, Y.; Wade, N.; Miyahara, K.

    1990-01-01

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

  10. 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)

  11. Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

    International Nuclear Information System (INIS)

    Busby, J.T.; Was, G.S.; Kenik, E.A.

    2002-01-01

    Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 deg. C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 deg. C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 deg. C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 deg. C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same

  12. 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

  13. Effect of structure and alloying elements on void formation in austenitic steels and nickel alloys

    International Nuclear Information System (INIS)

    Levy, V.; Azam, N.; Le Naour, L.; Didout, G.; Delaplace, J.

    1977-01-01

    In the development of the fast breeder reactors the phenomenon of metal swelling due to the formation of radiation induced voids is a large problem. In the complex alloys small fluctuations in composition can have a considerable effect on swelling and a great deal of investigation on the effect of both major and minor alloying elements is needed to be able to predict swelling. To provide more insight a research program involving irradiation of both commercial or specially cast alloys by 500 keV Ni + ions or 1 MeV electrons has been developed. The major results are presented

  14. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Voyevodin, V.N.; Karpov, S.A.; Kopanets, I.E.; Ruzhytskyi, V.V. [National Science Center “Kharkov Institute of Physics and Technology” Kharkov, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Tolstolutskaya, G.D., E-mail: g.d.t@kipt.kharkov.ua [National Science Center “Kharkov Institute of Physics and Technology” Kharkov, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Garner, F.A. [Radiation Effects Consulting, Richland, WA (United States)

    2016-01-15

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

  16. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-15

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

  17. Evaluation of Ion Irradiation Behavior of ODS Alloys

    International Nuclear Information System (INIS)

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

    2006-08-01

    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

  18. 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

  19. 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

  20. 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.

  1. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Science.gov (United States)

    Lu, Z.; Faulkner, R. G.; Morgan, T. S.

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

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

    International Nuclear Information System (INIS)

    Kuramoto, Eiichi; Abe, Hironobu; Tanaka, Minoru; Nishi, Kazuya; Tomiyama, Noriyuki.

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X. [Univ. of Alberta, Dept. of Chemical and Materials Engineering, Edmonton, Alberta (Canada); Zheng, W. [Materials Technology Laboratory, NRCan, Ottawa, Ontario (Canada); Guzonas, D.A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

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

  4. Characterization of the martensite phase formed during hydrogen ion irradiation in austenitic stainless steel

    Science.gov (United States)

    Jin, Hyung-Ha; Lim, Sangyeob; Kwon, Junhyun

    2017-10-01

    Microstructural changes in austenitic stainless steel caused by hydrogen ion irradiation were investigated using transmission electron microscopy (TEM). It has been confirmed that the irradiation induced the formation of martensite along the grain boundary; the martensite phase exhibited a crystal orientation relationship with the adjacent austenite phase. The results of this study also indicate that the concentration of Cr in the martensite phase is lower compared to that in the austenite matrix. The TEM results showed the development of asymmetric radiation-induced segregation (RIS) near the grain boundary, which leads to local changes in the chemical composition such as reduction of Cr near the grain boundary. The asymmetric RIS serves as a prerequisite for the formation of the martensite under hydrogen irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  6. 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

  7. Weldability of newly developed austenitic alloy for cryogenic service

    International Nuclear Information System (INIS)

    Ogawa, T.; Koseki, T.

    1986-01-01

    The testing reported in this paper involved typical steels of the new grades such as STEEL-A (0.025C-14Ni-25Cr-0.35N), STEEL-B (0.04C-23Mn-13Cr-0.22N) and STEEL-C (0.20C-25Mn-5Cr), and commercial steels of Type 300 series. Weldments were made mainly using the GTAW, SMAW and SAW processes with experimental and commercial filler metals. Strength and toughness of weldments were examined at 77 K (-321 0 F) and 4 K. The strengthening of material through the addition of nitrogen was far greater in the weld metal that in the base metal at cryogenic temperature. In fact, 0.2% proof stress of weld metals bearging 0.20% to 0.40% nitrogen at 77 K exhibited a higher value by 60 to 150 MPa (8,740 to 21,760 psi) than that of the base metal. Impact absorbed energy of weld metals at 77 K decreased rapidly with nitrogen content, 60-90 J at 0.20%N to 20-50J at 0.35% N. Rather high impact absorbed energy was obtained when the weld metal solidified as primary austenitic phase, resulting in fully austenitic microstructure or austenite-eutectic ferrite mixture at ambient temperature. In addition, oxide inclusions, the number of which strongly depends on welding processes, were detrimental to toughness of weld metals at cryogenic temperature

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

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Rowcliffe, A.F.; Leitnaker, J.M.

    1977-01-01

    The present invention is based on the discovery that radiation-induced voids which occur during fast neutron irradiation can be controlled by small but effective additions of titanium and silicon. The void-suppressing effect of these metals in combination is demonstrated and particularly apparent in austenitic stainless steels. 3 figures, 3 tables

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

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Rowcliffe, A.F.; Leitnaker, J.M.

    1979-01-01

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

  10. Precipitation of Second Phases in High-Interstitial-Alloyed Austenitic Steel

    Science.gov (United States)

    Lee, Tae-Ho; Ha, Heon-Young; Kim, Sung-Joon

    2011-12-01

    The precipitation reaction of an austenitic stainless steel containing N + C was investigated using transmission electron microscopy. The main precipitate formed during isothermal aging at 1123 K (850 °C) was M23C6 carbide, and its morphology gradually changed in a sequence of intergranular (along grain boundary) → cellular (or discontinuous) → intragranular (within grain interior) form with aging time. Irrespective of different morphologies, the M23C6 was consistently related to austenite matrix in accordance with the cube-on-cube orientation relationship. Based on the analysis of electron diffraction, two variants of intragranular M23C6 were identified, and they were related to each other by twin relation. Prolonged aging produced other types of precipitates—the rod-shaped Cr2N and the coarse irregular intermetallic sigma phase. The similarities and differences in precipitation behavior between N only and N + C alloyed austenitic stainless steels are briefly discussed.

  11. 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.

  12. 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

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

    OpenAIRE

    E. Rożniata; R. Dziurka; J. Pacyna

    2011-01-01

    Purpose: Present work corresponds to the research on the kinetics of phase transformations of undercooled austenite of Mn-Ni iron based model alloy. The kinetics of phase transformations of undercooled austenite of investigated alloy was presented on CCT diagram (continuous cooling transformation). Also the methodology of a dilatometric samples preparation and the method of the critical points determination were described.Design/methodology/approach: The austenitising temperature was defined ...

  14. The Effects of Alloy Chemistry on Localized Corrosion of Austenitic Stainless Steels

    Science.gov (United States)

    Sapiro, David O.

    This study investigated localized corrosion behavior of austenitic stainless steels under stressed and unstressed conditions, as well as corrosion of metallic thin films. While austenitic stainless steels are widely used in corrosive environments, they are vulnerable to pitting and stress corrosion cracking (SCC), particularly in chloride-containing environments. The corrosion resistance of austenitic stainless steels is closely tied to the alloying elements chromium, nickel, and molybdenum. Polarization curves were measured for five commercially available austenitic stainless steels of varying chromium, nickel, and molybdenum content in 3.5 wt.% and 25 wt.% NaCl solutions. The alloys were also tested in tension at slow strain rates in air and in a chloride environment under different polarization conditions to explore the relationship between the extent of pitting corrosion and SCC over a range of alloy content and environment. The influence of alloy composition on corrosion resistance was found to be consistent with the pitting resistance equivalent number (PREN) under some conditions, but there were also conditions under which the model did not hold for certain commercial alloy compositions. Monotonic loading was used to generate SCC in in 300 series stainless steels, and it was possible to control the failure mode through adjusting environmental and polarization conditions. Metallic thin film systems of thickness 10-200 nm are being investigated for use as corrosion sensors and protective coatings, however the corrosion properties of ferrous thin films have not been widely studied. The effects of film thickness and substrate conductivity were examined using potentiodynamic polarization and scanning vibrating electrode technique (SVET) on iron thin films. Thicker films undergo more corrosion than thinner films in the same environment, though the corrosion mechanism is the same. Conductive substrates encourage general corrosion, similar to that of bulk iron

  15. 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.

  16. Severe embrittlement of neutron irradiated austenitic steels arising from high void swelling

    Energy Technology Data Exchange (ETDEWEB)

    Neustroev, V.S. [FSUE ' SSC RF Research Institute of Atomic Reactors' , Dimitrovgrad (Russian Federation)], E-mail: neustroev@niiar.ru; Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

    2009-04-30

    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 and introducing limitations on low temperature handling especially. It is shown that the degradation is actually a form of quasi-embrittlement arising from intense flow localization with high levels of localized ductility involving micropore coalescence and void-to-void cracking. Voids initially serve as hardening components whose effect is overwhelmed by the void-induced reduction in shear and Young's moduli at high swelling levels. Thus the alloy appears to soften even as the ductility plunges toward zero on a macroscopic level although a large amount of deformation occurs microscopically at the failure site. Thus the failure is better characterized as 'quasi-embrittlement' which is a suppression of uniform deformation. This case should be differentiated from that of real embrittlement which involves the complete suppression of the material's capability for plastic deformation.

  17. Overview of MIT, ADIP irradiation experiments

    International Nuclear Information System (INIS)

    Kohse, G.; Harling, O.K.; Grant, N.J.

    1985-06-01

    Various rapidly solidified austenitic, ferritic and copper alloys have been produced at MIT for inclusion in ADIP neutron irradiation experiments. A brief summary of the alloys and their preparation and the achieved or projected irradiation parameters is provided

  18. 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.

  19. Thermal Effects That Arise upon Different Heat Treatments in Austenitic Steels Alloyed with Titanium and Phosphorus

    Science.gov (United States)

    Arbuzov, V. L.; Berger, I. F.; Bobrovskii, V. I.; Voronin, V. I.; Danilov, S. E.; Kazantsev, V. A.; Kataev, N. V.; Sagaradze, V. V.

    2018-04-01

    Structural and microstructural changes that arise in the course of the heat treatment of Cr-Ni-Mo austenitic stainless steels with different concentrations of titanium and phosphorus have been studied. It has been found that the alloying with phosphorus decreases the lattice parameter of these steels. The phosphorus contribution to this effect is 0.015 ± 0.002 Å/at %. Aging at a temperature of 670 K for about 20 h leads to the precipitation of dispersed needle-like particles, which are most likely to be iron phosphides. In the temperature range of 700-800 K, in austenitic steels, the atomic separation of the solid solution occurs, the intensity of which decreases upon alloying with titanium or phosphorus at concentrations of 1.0 and 0.1 wt %, respectively. At higher temperatures (about 950 K), the formed precipitates of the Ni3Ti (γ') phase increase in size to 7-10 nm.

  20. Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

    International Nuclear Information System (INIS)

    Kishimoto, N.; Amekura, H.; Saito, T.

    1994-01-01

    Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

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

    International Nuclear Information System (INIS)

    Saez-Maderuelo, A.; Gomez-Briceno, D.; Diego, G.

    2015-01-01

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

  2. 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

  3. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Precipitation response of austenitic stainless steel to simulated fusion irradiation

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1978-01-01

    The precipitation response of annealed type 316 stainless steel irradiated in HFIR is studied and compared to previously observed thermal aging and fast reactor irradiation responses. Irradiation in HFIR simultaneously produces high levels of helium and displacement damage and partially simulates a fusion environment. Samples have been irradiated at temperatures from 550 to 680 0 C to fluences producing up to 3300 appm He and 47 dpa

  6. Microstructural features of dissimilar welds between 316LN austenitic stainless steel and alloy 800

    International Nuclear Information System (INIS)

    Sireesha, M.; Sundaresan, S.

    2000-01-01

    For joining type 316LN austenitic stainless steel to modified 9Cr-1Mo steel for power plant application, a trimetallic configuration using an insert piece (such as alloy 800) of intermediate thermal coefficient of expansion (CTE) has been sometimes suggested for bridging the wide gap in CTE between the two steels. Two joints are thus involved and this paper is concerned with the weld between 316LN and alloy 800. These welds were produced using three types of filler materials: austenitic stainless steels corresponding to 316,16Cr-8Ni-2Mo, and the nickel-base Inconel 182 1 . The weld fusion zones and the interfaces with the base materials were characterised in detail using light and transmission electron microscopy. The 316 and Inconel 182 weld metals solidified dendritically, while the 16-8-2(16%Cr-8%Ni-2%Mo) weld metal showed a predominantly cellular substructure. The Inconel weld metal contained a large number of inclusions when deposited from flux-coated electrodes, but was relatively inclusion-free under inert gas-shielded welding. Long-term elevated-temperature aging of the weld metals resulted in embrittling sigma phase precipitation in the austenitic stainless steel weld metals, but the nickel-base welds showed no visible precipitation, demonstrating their superior metallurgical stability for high-temperature service. (orig.)

  7. 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.)

  8. 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

  9. 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)

  10. 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

  11. Microstructure characterization in the weld joint of a high nickel austenitic alloy and Cr18-Ni8 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Na; Li, Yajiang; Wang, Juan [Shandong Univ., Jinan (CN). Key Lab. for Liquid - Solid Structural Evolution and Processing of Materials (Ministry of Education)

    2012-06-15

    High nickel austenitic alloy, 6 mm thick, and Cr18-Ni8 stainless steel with a thickness of 0.6 mm were joined by pulsed current tungsten inert gas arc welding without filler metal in this work. Metallographic examination, microhardness measurement and electron microprobe analysis were used to reveal microstructural characteristics in the joint. The results indicated that the weld metal consisted of {gamma}-austenite, {delta}-ferrite and carbides without the appearance of martensite. There were dendrite crystals at the edge of the weld metal near the high nickel austenitic alloy and isometric crystals in the center of the weld metal. The microhardness of the weld metal was the highest due to the existence of carbides and its finer structure. Graphite flakes were still embedded in the austenite matrix of the heat-affected zone without the formation of martensite. (orig.)

  12. Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

    Science.gov (United States)

    Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

    Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

  13. Effect of cyclic electron irradiation on mechanical properties of austenite steel

    International Nuclear Information System (INIS)

    Tsepelev, A.B.; Sadykhov, S.I.O.; Chernov, A.I.; Sevost'yanov, M.A.

    2006-01-01

    To check the supposition on the possibility of radiation-stimulated process enhancement under cyclic irradiation conditions an experimental investigation is carried out to elucidate the effect of the mode of irradiation (continuous or cyclic) on mechanical properties of chromium-manganese austenitic stainless steel type 10Kh12G20V. The effect of some radiation hardening is observed under cyclic irradiation, however, the data obtained cannot be considered as good evidence for the validity of proposed model of dynamic preference if the scatter in experimental data is taken into account [ru

  14. Localized corrosion of high alloyed austenitic stainless steels

    International Nuclear Information System (INIS)

    Morach, R.; Schmuki, P.; Boehni, H.

    1992-01-01

    The susceptibility of several high alloyed stainless steels against localized corrosion was investigated by traditional potentiostatic and -kinetic methods and the current transient technique. Different test cells, proposed in literature, were evaluated for use in testing of plate materials. The AVESTA-cell showed to be not useful for potentiokinetic current density potential curves, but useable for pitting experiments. After pickling and prepassivation epoxy embedded materials proved to be resistant to crevice corrosion at the metal-resin interface. The electrode in form of a wire was the most reliable crevice free cell design. The grinding of the samples in the pretreatment procedure was found to have a large effect on the pitting corrosion behaviour. Using different paper types with varying grit, a drop in pitting potential for rougher surfaces and an increase in metastable pitting activity was found. Increasing surface roughness led also to changes in the electronic structure of the passive film reflected by a lower bandgap energy. High alloyed stainless steels showed no breakdown potential within the examined potential range. Compared to 18/8 type stainless steels significantly less transients were found. The number of transients decreases with increasing molybdenum and chromium content

  15. Study on creep-fatigue life of irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji

    2001-01-01

    The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2 dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. The creep damage of both unirradiated and irradiated specimens was underestimated by the time fraction rule or the ductility exhaustion rule. The creep damage calculated by the time fraction rule or the ductility exhaustion rule increased by the irradiation. The predictions derived from the linear damage rule are unsafe as compared with the experimental fatigue lives. (author)

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

    International Nuclear Information System (INIS)

    Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi; Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao.

    1996-01-01

    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,α) 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.)

  17. 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.)

  18. 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

  19. Characterization of the dissimilar welding - austenitic stainless steel with filler metal of the nickel alloy

    International Nuclear Information System (INIS)

    Soares, Bruno Amorim; Schvartzman, Monica Maria de Abreu Mendonca; Campos, Wagner Reis da Costa

    2007-01-01

    In elevated temperature environments, austenitic stainless steel and nickel alloy has a superior corrosion resistance due to its high Cr content. Consequently, this alloys is widely used in nuclear reactors components and others plants of energy generation that burn fossil fuel or gas, chemical and petrochemical industries. The object of the present work was to research the welding of AISI 304 austenitic stainless steel using the nickel alloy filler metals, Inconel 625. Gas tungsten arc welding, mechanical and metallographic tests, and compositional analysis of the joint were used. A fundamental investigation was undertaken to characterize fusion boundary microstructure and to better understand the nature and character of boundaries that are associated with cracking in dissimilar welds. The results indicate that the microstructure of the fusion zone has a dendritic structure, inclusions, and precipitated phases containing Ti and Nb are present in the inter-dendritic region. In some parts near to the fusion line it can be seen a band in the weld, probably a eutectic phase with lower melting point than the AISI 304, were the cracking may be beginning by stress corrosion. (author)

  20. First stage of the structural evolution of austenite in Cu-Al-Ni shape memory alloys

    International Nuclear Information System (INIS)

    Pelosin, V.; Gerland, M.; Riviere, A.

    2001-01-01

    Two shape memory Cu-Al-Ni alloys, a polycrystal and a single crystal, exhibiting a martensitic transformation close to 130 C (in the as-quenched state) have been studied. Specimens have been quenched after heat treatment at 850 C. The structural evolutions of the high temperature phase (austenite) have been studied for thermal treatments performed below 200 C. Investigations have been carried out using electrical resistivity measurements, TEM (transmission electron microscopy) observations and X-ray diffraction analysis. The main structural modifications are observed in the polycrystalline alloy and concern first, the reordering process of the austenite structure (B2→L2 1 ), and second, the precipitation of the (Cu 9 Al 4 ) γ 2 phase. In the single crystal alloy, the evolutions are very slight and localized on the structural defects. Particular attention is paid to the role of the quenched-in vacancy elimination on the observed mechanisms. In addition, the incidence of the structural evolution on the transformation temperatures is also discussed. (orig.)

  1. 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)

  2. Irradiation of copper alloys in FFTF

    International Nuclear Information System (INIS)

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

    1984-01-01

    Nine copper-base alloys in thirteen material conditions have been inserted into the MOTA-18 experiment for irradiation in FFTF at approx.450 0 C. The alloy Ni-1.9Be is also included in this experiment, which includes both TEM disks and miniature tensile specimens

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

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jae-Hyeok, E-mail: jhshim@kist.re.kr [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); High Temperature Energy Materials Research Center, Korea Institute of Science and Technology, Seoul 136-791 (Korea, Republic of); Povoden-Karadeniz, Erwin [Christian Doppler Laboratory for Early Stages of Precipitation, Vienna University of Technology, A-1040 Vienna (Austria); Kozeschnik, Ernst [Institute of Materials Science and Technology, Vienna University of Technology, A-1040 Vienna (Austria); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2015-07-15

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

  4. 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

  5. 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.)

  6. Fatigue and creep–fatigue deformation of an ultra-fine precipitate strengthened advanced austenitic alloy

    International Nuclear Information System (INIS)

    Carroll, M.C.; Carroll, L.J.

    2012-01-01

    An advanced austenitic alloy, HT-UPS (high-temperature ultrafine-precipitation-strengthened), has been identified as an ideal candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS alloys demonstrate improved creep resistance relative to 316 stainless steel (SS) through additions of Ti and Nb, which precipitate to form a widespread dispersion of stable nanoscale metallic carbide (MC) particles in the austenitic matrix. To investigate the behavior in more representative conditions than are offered by uniaxial creep tests, the low-cycle continuous fatigue and combined creep–fatigue response of an HT-UPS alloy have been investigated at 650 °C and 1.0% total strain, with an R-ratio of −1 and hold times at peak tensile strain of up to 150 min. The cyclic deformation response of HT-UPS is directly compared to that of standard 316 SS. The measured values for total cycles to failure between the two alloys are similar, despite differences in peak stress profiles and in qualitative observations of the deformed microstructures. Crack propagation is primarily transgranular in both fatigue and creep–fatigue of each alloy at the investigated conditions. Internal grain boundary damage in the form of fine cracks resulting from the tensile hold is present following the application of hold times of 60 min and longer, and considerably more internal cracks are quantifiable in 316 SS than in HT-UPS. The dislocation substructures observed in the deformed material differ substantially; an equiaxed cellular structure is observed in the microstructure of 316 SS, whereas HT-UPS exhibits widespread and relatively homogenous tangles of dislocations pinned by the nanoscale MC precipitates. The significant effect of the fine distribution of precipitates on observed fatigue and creep–fatigue response is described in three distinct behavioral regions as the microstructure evolves with continued cycling.

  7. Fatigue and creep-fatigue deformation of an ultra-fine precipitate strengthened advanced austenitic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Carroll, M.C., E-mail: Mark.Carroll@INL.gov [Idaho National Laboratory, 1955 Fremont, PO Box 1625, Idaho Falls, ID 83415-2218 (United States); Carroll, L.J. [Idaho National Laboratory, 1955 Fremont, PO Box 1625, Idaho Falls, ID 83415-2218 (United States)

    2012-10-30

    An advanced austenitic alloy, HT-UPS (high-temperature ultrafine-precipitation-strengthened), has been identified as an ideal candidate material for the structural components of fast reactors and energy-conversion systems. HT-UPS alloys demonstrate improved creep resistance relative to 316 stainless steel (SS) through additions of Ti and Nb, which precipitate to form a widespread dispersion of stable nanoscale metallic carbide (MC) particles in the austenitic matrix. To investigate the behavior in more representative conditions than are offered by uniaxial creep tests, the low-cycle continuous fatigue and combined creep-fatigue response of an HT-UPS alloy have been investigated at 650 Degree-Sign C and 1.0% total strain, with an R-ratio of -1 and hold times at peak tensile strain of up to 150 min. The cyclic deformation response of HT-UPS is directly compared to that of standard 316 SS. The measured values for total cycles to failure between the two alloys are similar, despite differences in peak stress profiles and in qualitative observations of the deformed microstructures. Crack propagation is primarily transgranular in both fatigue and creep-fatigue of each alloy at the investigated conditions. Internal grain boundary damage in the form of fine cracks resulting from the tensile hold is present following the application of hold times of 60 min and longer, and considerably more internal cracks are quantifiable in 316 SS than in HT-UPS. The dislocation substructures observed in the deformed material differ substantially; an equiaxed cellular structure is observed in the microstructure of 316 SS, whereas HT-UPS exhibits widespread and relatively homogenous tangles of dislocations pinned by the nanoscale MC precipitates. The significant effect of the fine distribution of precipitates on observed fatigue and creep-fatigue response is described in three distinct behavioral regions as the microstructure evolves with continued cycling.

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Qiaofeng [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Su, Qing [Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Wang, Fei [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Zhang, Chenfei; Lu, Yongfeng [Department of Electrical Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nastasi, Michael [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Cui, Bai, E-mail: bcui3@unl.edu [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States)

    2017-06-15

    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. - Highlights: •Laser shock peening generates a dislocation network, stacking faults and deformation twins in stainless steels. •Dislocations and incoherent twin boundaries serve as effective sinks for the annihilation of irradiation defects. •Incoherent twin boundaries remain as stable and effective defect sinks at 300 °C.

  10. Irradiation creep of dispersion strengthened copper alloy

    International Nuclear Information System (INIS)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

    1997-01-01

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

  11. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

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

    International Nuclear Information System (INIS)

    Simon, N.

    1992-06-01

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

  13. 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

  14. 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)

  15. 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

  16. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  17. 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.

  18. Irradiation creep in simple binary alloys

    International Nuclear Information System (INIS)

    Nagakawa, J.; Sethi, V.K.; Turner, A.P.L.

    1981-07-01

    Creep enhancement during 21-MeV deuteron irradiation was examined at 350 0 C for two simple binary alloys with representative microstructures, i.e., solid-solution (Ni - 4 at. % Si) and precipitation-hardened (Ni - 12.8 at. % Al) alloys. Coherent precipitates were found to be very effective in suppressing irradiation-enhanced creep. Si solute atoms depressed irradiation creep moderately and caused irradiation hardening via radiation-induced segregation. The stress-dependence of irradiation creep in Ni - 4 at. % Si should a transition, which seems to reflect a change of mechanism from dislocation climb due to stress-induced preferential absorption (SIPA) to climb-controlled dislocation glide enhanced by irradiation

  19. Alloying effect on the structure and properties of austenitic heat-resistant steels

    International Nuclear Information System (INIS)

    Levitin, V.V.; Grabovskij, V.Ya.; Korostelev, V.F.; Ryvkin, Yu.A.

    1978-01-01

    Investigated have been mechanical properties at test temperatures of 20-95O deg C, wear resistance, softening at thermomechanical cycling and microstructure of cast austenitic chromium-nickel steels (13%Cr + 35%Ni), produced by electroslag remelting with variations in Ti, Mo, Nb and W contents. Regression equations for relationship of the investigated characteristics to alloying element content have been obtained. Titanium, molybdenum and niobium increasing hardness and strength limit at room and high temperatures promote a decrease in ductility. Tungsten increases strength properties, wear resistance and thermal stability of the steels without negative effect on the impact strength. The impact strength decrease with an increase in alloying is due to brittle precipitations along the boundaries of as-cast grains, containing Ti, Mo, Nb and Si

  20. 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

  1. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    Science.gov (United States)

    Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

    2012-12-01

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

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

    International Nuclear Information System (INIS)

    Mohamed, H.G.; Matta, M.K.

    2001-01-01

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

  3. Fracture toughness of irradiated wrought and cast austenitic stainless steels in BWR environment

    International Nuclear Information System (INIS)

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

    2007-01-01

    Experimental data are presented on the fracture toughness of wrought and cast austenitic stainless steels (SSs) that were irradiated to a fluence of ∼ 1.5 x 10 21 n/cm 2 (E > 1 MeV) * (∼ 2.3 dpa) at 296-305 o C. To evaluate the possible effects of test environment and crack morphology on the fracture toughness of these steels, all tests were conducted in normal-water-chemistry boiling water reactor (BWR) environments at ∼ 289 o C. Companion tests were also conducted in air on the same material for comparison. The fracture toughness J-R curves for SS weld heat-affected-zone materials in BWR water were found to be comparable to those in air. However, the results of tests on sensitized Type 304 SS and thermally aged cast CF-8M steel suggested a possible effect of water environment. The available fracture toughness data on irradiated austenitic SSs were reviewed to assess the potential for radiation embrittlement of reactor-core internal components. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components are also discussed. (author)

  4. High Ni austenite stainless steel resistant to neutron irradiation degradation

    International Nuclear Information System (INIS)

    Yonezawa, Toshio; Iwamura, Toshihiko; Kanasaki, Hiroshi; Fujimoto, Koji; Nakata, Shizuo; Ajiki, Kazuhide; Nakamura, Mitsuhiro.

    1997-01-01

    The composition of the stainless steel of the present invention comprises from 0.005 to 0.08% of C, up to 3% of Mn, up to 0.2% of Si+P+S, from 25 to 40% of Ni, from 25 to 40% of Cr, up to 3% of Mo, up to 0.3% of Nb+Ta, up to 0.3% of Ti, up to 0.001% of B and the balance of Fe. A solid solubilization treatment at a temperature of from 1,000 to 1,150degC is applied to the stainless steel having the composition. The stainless steel is excellent in stress corrosion cracking-resistance at a working circumstance of a LWR type reactor (high temperature and high pressure water at from 270 to 350degC/from 70 to 160 atm even after undergoing neutron irradiation of about 1 x 10 22 n/cm 2 (E>1 MeV) which is a maximum neutron irradiation amount undergone till the final stage of the working life of the LWR-type reactor. In addition, the average thermal expansion coefficient at from room temperature to 400degC ranges from 15x10 -6 - 19x10 -6 /K. (I.N.)

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

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop A125), and a precipitation-hardened copper alloy (Cu-Cr-Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop A125 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface. (orig.)

  7. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2015-11-15

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

  9. Changes in grain boundary composition induced by neutron irradiation of austenitic stainless steels

    International Nuclear Information System (INIS)

    Asano, K.; Nakata, K.; Fukuya, K.; Kodama, M.

    1992-01-01

    The radiation induced segregation of solutes to the grain boundary in austenitic stainless steels were studied. Type 304 and type 316 steel samples neutron irradiated at 561K up to 9.2x10 25 n/m 2 were obtained and minute compositional profiles across grain boundaries were examined using an analytical scanning transmission electron microscope equipped with a field emission electron gun. Chromium was slightly enriched at grain boundaries at the lowest irradiation dose but decreased with increasing fluence. Higher fluence irradiation resulted in depletion in chromium and molybdenum, and enrichment in nickel, silicon and phosphorus. These changes in grain boundary chemistry were limited within about 5nm of the boundary. Significant depletion of chromium and enrichment of impurities on the grain boundary occurred at fluences roughly coincidental with that of SCC susceptibility change obtained from another project

  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. Plastic deformation and fracture behaviors of nitrogen-alloyed austenitic stainless steels

    International Nuclear Information System (INIS)

    Wang Songtao; Yang Ke; Shan Yiyin; Li Laifeng

    2008-01-01

    The plastic deformation and fracture behaviors of two nitrogen-alloyed austenitic stainless steels, 316LN and a high nitrogen steel (Fe-Cr-Mn-0.66% N), were investigated by tensile test and Charpy impact test in a temperature range from 77 to 293 K. The Fe-Cr-Mn-N steel showed ductile-to-brittle transition (DBT) behavior, but not for the 316LN steel. X-ray diffraction (XRD) confirmed that the strain-induced martensite occurred in the 316LN steel, but no such transformation in the Fe-Cr-Mn-N steel. Tensile tests showed that the temperature dependences of the yield strength for the two steels were almost the same. The ultimate tensile strength of the Fe-Cr-Mn-N steel displayed less significant temperature dependence than that of the 316LN steel. The strain-hardening exponent increased for the 316LN steel, but decreased for the Fe-Cr-Mn-N steel, with decreasing temperature. Based on the experimental results and the analyses, a modified scheme was proposed to explain the fracture behaviors of austenitic stainless steels

  12. 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)

  13. The evolution of mechanical property change in irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Lucas, G.E.

    1993-01-01

    The evolution of mechanical properties in austenitic stainless steels during irradiation is reviewed. Changes in strength, ductility and fracture toughness are strongly related to the evolution of the damage microstructure and microstructurally-based models for strengthening reasonably correlate the data. Irradiation-induced defects promote work softening and flow localization which in turn leads to significant reductions in ductility and fracture toughness beyond about 10 dpa. The effects of irradiation on fatigue appear to be modest except at high temperature where helium embrittlement becomes important. The swelling-independent component of irradiation creep strain increases linearly with dose and is relatively insensitive to material variables and irradiation temperature, except at low temperatures where accelerated creep may occur as a result of low vacancy mobility. Creep rupture life is a strong function of helium content, but is less sensitive to metallurgical conditions. Irradiation-induced stress corrosion cracking appears to be related to the evolution of radiation-induced segregation/depletion at grain boundaries, and hence may not be significant at low irradiation temperatures. (orig.)

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

    International Nuclear Information System (INIS)

    Mazey, D.J.; Walters, G.P.; Buckley, S.N.; Bullough, R.; Hanks, W.; Bolster, D.E.J.; Sowden, B.C.; Lurcook, D.; Murphy, S.M.

    1985-03-01

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

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

    Directory of Open Access Journals (Sweden)

    Serna, A.

    2003-12-01

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

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

  18. Annealing effect on redistribution of atoms in austenite of Fe-Ni-Mo and Fe-Ni-Si alloys

    International Nuclear Information System (INIS)

    Rodionov, Yu.L.; Isfandiyarov, G.G.; Zambrzhitskij, V.N.

    1980-01-01

    Using the Moessbauer spectrum method, studied has been the change in the fine atomic structure of the Fe-(28-36)%Ni austenite alloys with Mo and Si additives during annealing in the 200-800 deg C range. Also, the energy of the activation of processes, occurring at the annealing temperatures of below 500 deg C has been researched. On the basis of the obtained results a conclusion is drawn that the annealing of the investigated alloys at 300-500 deg C is conducive to the redistribution of the atoms of the alloying element and to the formation of regions with a higher content of Ni and Mo(Si) atoms

  19. 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.

  20. Neutron resistant irradiation alloy and usage thereof

    International Nuclear Information System (INIS)

    Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

    1997-01-01

    A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

  1. 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

  2. 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

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

    Science.gov (United States)

    Hug, E.; Prasath Babu, R.; Monnet, I.; Etienne, A.; Moisy, F.; Pralong, V.; Enikeev, N.; Abramova, M.; Sauvage, X.; Radiguet, B.

    2017-01-01

    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 56Fe5+ 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.

  4. 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.

  5. 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.)

  6. 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)

  7. Effects of alloying elements on defect structures in the incubation period for void swelling in austenitic stainless steels

    International Nuclear Information System (INIS)

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

    2013-01-01

    Positron lifetime measurements were used to study the effects of alloying elements on the defect structure during the incubation period for void swelling in several fcc model alloys. Pure Ni, four model alloys (Fe–Cr–Ni, Fe–Cr–Ni–Mo–Mn, Fe–Cr–Ni–Mo–Mn–Si and Fe–Cr–Ni–Mo–Mn–Si–Ti), and four commercial alloys (SUS316LSS, SUS316SS, SUS304SS and Ti added modified SUS316SS) were irradiated with electrons and neutrons. Even at 363 and 573 K to a dose of 0.2 dpa, an effect of alloying elements was observed. At 363 K irradiation, voids were formed only in Ni and Fe–Cr–Ni. At 573 K irradiation, voids were formed in Ni and all model alloys, though the concentration depended on the alloying elements. In commercial alloys, precipitates were formed instead of vacancy clusters, which prevented void growth

  8. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K., E-mail: ksato@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Xu, Q.; Yoshiie, T. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dai, Y. [Spallation Neutron Source Division, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kikuchi, K. [Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Tokai-mura, Naka-gun, Ibaraki 319-1106 (Japan)

    2012-12-15

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<{approx}0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

  9. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    International Nuclear Information System (INIS)

    Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

    2012-01-01

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<∼0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

  11. 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; Yichuan, Zhang

    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)

  12. 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.

  13. 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

  14. Plastic strain characterization in austenitic stainless steels and nickel alloys by electron backscatter diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Saez-Maderuelo, A., E-mail: alberto.saez@ciemat.es [CIEMAT, Av. Complutense, 22-28040 Madrid (Spain); Castro, L.; Diego, G. de [CIEMAT, Av. Complutense, 22-28040 Madrid (Spain)

    2011-09-01

    Stress corrosion cracking (SCC) is enhanced by cold work and causes many problems in components of the nuclear power plants. Besides, during manufacturing, installation, welding and service of the material, residual strains can be produced increasing the susceptibility to SCC. For this reason, it is important to characterize the degree of plastic strain due to dislocation accumulation in each crystal. Electron backscatter diffraction (EBSD), in conjunction with scanning electron microscope (SEM), has been a great advance in this field because it enables to estimate the plastic strain in a quick and easy way. Nevertheless, over the last few years, a lot of different mathematical expressions to estimate the plastic strain have appeared in the literature. This situation hinders the election of one of them by a novel scientist in this field. Therefore, in this paper some of the more common expressions used in the calculation of the angular misorientation have been presented and discussed in order to clarify their more important aspects. Then, using one of these expressions (average local misorientation), curves relating misorientation density with known levels of strain will be obtained for an austenitic stainless steel 304L and nickel base alloy 690, which have shown a linear behaviour that is in good agreement with results found in the literature. Finally, using curves obtained in previous steps, levels of plastic strain in a plate of nickel base alloy 600 welded with weld metal 182 were estimated between 8 and 10% for a high temperature mill annealing sample.

  15. Plastic strain characterization in austenitic stainless steels and nickel alloys by electron backscatter diffraction

    International Nuclear Information System (INIS)

    Saez-Maderuelo, A.; Castro, L.; Diego, G. de

    2011-01-01

    Stress corrosion cracking (SCC) is enhanced by cold work and causes many problems in components of the nuclear power plants. Besides, during manufacturing, installation, welding and service of the material, residual strains can be produced increasing the susceptibility to SCC. For this reason, it is important to characterize the degree of plastic strain due to dislocation accumulation in each crystal. Electron backscatter diffraction (EBSD), in conjunction with scanning electron microscope (SEM), has been a great advance in this field because it enables to estimate the plastic strain in a quick and easy way. Nevertheless, over the last few years, a lot of different mathematical expressions to estimate the plastic strain have appeared in the literature. This situation hinders the election of one of them by a novel scientist in this field. Therefore, in this paper some of the more common expressions used in the calculation of the angular misorientation have been presented and discussed in order to clarify their more important aspects. Then, using one of these expressions (average local misorientation), curves relating misorientation density with known levels of strain will be obtained for an austenitic stainless steel 304L and nickel base alloy 690, which have shown a linear behaviour that is in good agreement with results found in the literature. Finally, using curves obtained in previous steps, levels of plastic strain in a plate of nickel base alloy 600 welded with weld metal 182 were estimated between 8 and 10% for a high temperature mill annealing sample.

  16. 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.

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

    International Nuclear Information System (INIS)

    Maziasz, P.J.; Roche, T.K.

    1981-01-01

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

  18. Comparison of ferritic and austenitic plasma nitriding and nitrocarburizing behavior of AISI 4140 low alloy steel

    International Nuclear Information System (INIS)

    Fattah, M.; Mahboubi, F.

    2010-01-01

    This paper compares the ferritic and austenitic plasma nitriding and nitrocarburizing behavior of AISI 4140 low alloy steel carried out to improve the surface corrosion resistance. The gas composition for plasma nitriding was 85% N 2 -15% H 2 and that for plasma nitrocarburizing was 85% N 2 -12% H 2 -3% CO 2 . Both treatments were performed for 5 h, for different process temperatures of 570 and 620 o C for ferritic and austenitic plasma treatment, respectively. Optical microscopy, X-ray diffraction and potentiodynamic polarization technique in 3.5% NaCl solution, were used to study the treated surfaces. The results of X-ray analysis revealed that with increasing the treatment temperature from 570 to 620 o C for both treatments, the amount of ε phase decreased and γ' phase increased. Nitrocarburizing treatment resulted in formation of a more amount of ε phase with respect to nitriding treatment. However, the highest amount of ε phase was observed in the ferritic nitrocarburized sample at 570 o C. The sample nitrided at 620 o C exhibited the thickest layer. The potentiodynamic polarization results revealed that after plasma nitriding and nitrocarburizing at 570 o C, corrosion potential increased with respect to the untreated sample due to the noble nitride and carbonitride phases formed on the surface. After increasing the treatment temperature from 570 to 620 o C, corrosion potential decreased due to the less ε phase development in the compound layer and more porous compound layer formed at 620 o C with respect to the treated samples at 570 o C.

  19. 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)

  20. Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

    International Nuclear Information System (INIS)

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

    1998-01-01

    A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

  1. 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)

  2. The Effects of CO{sub 2} Pressure on Corrosion and Carburization Behaviors of Chromia-forming Austenitic Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jung; Kim, Sung Hwan; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    By applying S-CO{sub 2} cycle to SFR, the inherent safety could be improved by alleviating the concern of explosive reaction between high temperature steam and liquid sodium as well as increased thermal efficiency at 500-550 .deg. C compared to helium Brayton cycle. Meanwhile, from the material point of view, a compatibility such as corrosion and carburization of candidate materials in S-CO{sub 2} environment should be evaluated to assure the long-term integrity of IHX. It has been previously reported that Ni-base alloys and high-Cr Fe-base austenitic alloys showed a good corrosion resistance by the formation of thin chromia layer while carburization behaviors of those materials were not properly investigated. Corrosion and carburization behaviors of three chromia-forming austenitic alloys (Ni-base alloys and Alloy 800HT) were evaluated in S-CO{sub 2} (200 bar) and CO{sub 2} (1 bar) environment at 550.650 .deg. C for 1000 h. For all test materials, a good corrosion resistance was exhibited by the formation of thin chromia (Cr{sub 2}O{sub 3}) with small amount of minor oxides such as Mn1.5Cr1.5O{sub 4}, Al{sub 2}O{sub 3}, and TiO{sub 2}.

  3. Investigation of irradiation induced inter-granular stress corrosion cracking susceptibility on austenitic stainless steels for PWR by simulated radiation induced segregation materials

    Energy Technology Data Exchange (ETDEWEB)

    Yonezawa, Toshio; Fujimoto, Koji; Kanasaki, Hiroshi; Iwamura, Toshihiko [Mitsubishi Heavy Industries Ltd., Takasago R and D Center, Takasago, Hyogo (Japan); Nakada, Shizuo; Ajiki, Kazuhide [Mitsubishi Heavy Industries Ltd., Kobe Shipyard and Machinery Works, Kobe, Hyogo (Japan); Urata, Sigeru [General Office of Nuclear and Fossil Power Production, Kansai Electric Power Co., Inc., Osaka (Japan)

    2000-07-01

    An Irradiation Assisted Stress Corrosion Cracking (IASCC) has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated Type 304 and Type 316 CW stainless steels. Low chromium, high nickel and silicon (12%Cr-28%Ni-3%Si) steel showed high susceptibility to PWSCC (Primary Water Stress Corrosion Cracking) by SSRT (Slow Strain Rate Tensile) test in simulated PWR primary water. PWSCC susceptibility of the test steels increases with a decrease of chromium content and a increase of nickel and silicon contents. The aged test steel included coherent M{sub 23}C{sub 6} carbides with matrices at the grain boundaries showed low PWSCC susceptibility. This tendency is in very good agreement with that of the PWSCC susceptibility of nickel based alloys X-750 and 690. From these results, if there is the possibility of IASCC for austenitic stainless steels in PWRs, in the future, the IASCC shall be caused by the PWSCC as a result of irradiation induced grain boundary segregation. (author)

  4. In-situ observation of weld joint of austenitic stainless steel due to helium irradiation

    International Nuclear Information System (INIS)

    Hamada, S.; Hojou, K.; Hishinuma, A.

    1992-01-01

    Microstructural evolution during helium ions irradiation in a weld metal containing 10% delta-ferrite of a weld joint of Ti-modified austenitic stainless steel were in-situ observed through a transmission electron microcopy. Very fine helium bubbles were observed in high number density in both a delta ferrite phase and a matrix to a dose of 3 x 10 19 ions·m -2 . Entirely different behavior appeared in both phases with increasing dose. Bubbles in a delta-ferrite phase were readily converted into voids during slight increment of dose, and these rapidly grew with additional increasing of dose. On the other hand, finer bubbles in a matrix were very stable during irradiation and did not grow any more up to 2 x 10 20 ions·m -2 . Swelling became much larger in a delta-ferrite phase than in a fcc matrix phase, resultantly ; This means an inverse phenomenon for conventional results that swelling is smaller in a ferrite phase than in a fcc phase. Sigma phase radiation-enhanced precipitated at the grain boundary between a delta-ferrite phase and a matrix at a dose 9 x 10 19 ions·m -2 . This phase grew in two dimensions with increasing dose. The chemical composition of the sigma phase observed during irradiation showed Cr and Mo enrichment, and Fe and Ni depletion compared with those of a sigma phase thermally produced. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro, E-mail: chimi.yasuhiro@jaea.go.jp [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kitsunai, Yuji [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Kasahara, Shigeki [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Chatani, Kazuhiro; Koshiishi, Masato [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Nishiyama, Yutaka [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2016-07-15

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

  6. Effect of Hot Rolling on the Microstructure and Mechanical Properties of Nitrogen Alloyed Austenitic Stainless Steel

    Science.gov (United States)

    Chenna Krishna, S.; Karthick, N. K.; Jha, Abhay K.; Pant, Bhanu; Cherian, Roy M.

    2018-05-01

    In the present investigation, the effect of multi-pass hot rolling in the temperature range of 700-1000 °C on the microstructure and mechanical properties of nitrogen alloyed austenitic stainless steel was studied with the aid of optical microscopy, tensile testing and x-ray diffraction measurements. The microstructural changes that occurred in the hot rolled specimens were elongation of grains in rolling direction, nucleation of new grains at the grain boundaries of elongated grains and growth of nucleated grains to form fully recrystallized grains. Elongated grains formed at lower rolling temperature (700-800 °C) due to inadequate strain/temperature for the initiation of dynamic recrystallization. At higher rolling temperature (900-1000 °C), fine grains formed due to dynamic recrystallization. Tensile properties showed strong dependency on the rolling temperature. Tensile strength increased with the decrease in the rolling temperature at the cost of ductility. Maximum strength was observed in samples hot rolled at 700 °C with yield strength of 917 MPa and ductility of 25%. This variation in the tensile properties with the rolling temperature is attributed to changes in the dislocation density and grain structure. The estimated yield strength from the dislocation density, solid solution and grain boundary strengthening closely matched with experimentally determined yield strength confirming the role of dislocation density and grain size in the strengthening.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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)

  13. 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)

  14. Microstructure of Au-ion irradiated 316L and FeNiCr austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Jublot-Leclerc, S., E-mail: stephanie.jublot-leclerc@csnsm.in2p3.fr [CSNSM, Univ Paris-Sud, CNRS, Université Paris Saclay, 91405 Orsay (France); Li, X. [CSNSM, Univ Paris-Sud, CNRS, Université Paris Saclay, 91405 Orsay (France); Legras, L.; Lescoat, M.-L. [EDF R& D, Groupe Métallurgie, Les Renardières, 77818 Moret sur Loing (France); Fortuna, F.; Gentils, A. [CSNSM, Univ Paris-Sud, CNRS, Université Paris Saclay, 91405 Orsay (France)

    2016-11-15

    Thin foils of 316L were irradiated in situ in a Transmission Electron Microscope with 4 MeV Au ions at 450 °C and 550 °C. Similar irradiations were performed at 450 °C in FeNiCr. The void and dislocation microstructure of 316L is found to depend strongly on temperature. At 450 °C, a dense network of dislocation lines is observed in situ to grow from black dot defects by absorption of other black dots and interstitial clusters whilst no Frank loops are detected. At 550 °C, no such network is observed but large Frank loops and perfect loops whose sudden appearance is concomitant with a strong increase in void density as a result of a strong coupling between voids and dislocations. Moreover, differences in both alloys microstructure show the major role played by the minor constituents of 316L, increasing the stacking fault formation energy, and possibly leading to significant differences in swelling behaviour. - Highlights: • 316L and FeNiCr were ion irradiated in situ in a TEM at elevated temperature. • The minor constituents of 316L play a major role in the resulting microstructure. • A dense network of dislocations develops in both alloys from black dot defects. • The nucleation and growth of voids and dislocations are strongly correlated. • The Frank loop mean size saturates at similar dpa values as in neutron irradiation.

  15. EDX microanalysis of neutron-irradiated alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1981-09-01

    Energy-dispersive X-ray (EDX) spectrometry of 50 nm thick specimens in the scanning transmission electron microscope provides quantitative elemental analyses of selected regions as small as 20 nm in diameter. To analyze highly radioactive neutron-irradiated alloys it is necessary to reduce the high counting deadtimes caused by energetic γ-Compton scattering in the Si(Li) detector, and to account for spurious background contributions from γ-rays and characteristic x-ray emissions. Several simple methods for overcoming effects of specimen radioactivity are described, including use of a tungsten collimator to attenuate γ and x-rays coming from the thick edges of self-supporting disk specimens. These methods allow analyses of Fe-Cr-Ni based alloys with γ-activities up to 1000 μC/sub i/. Techniques used to maintain high spatial resolution and accuracy in quantitatve analysis are also described, and their use is illustrated

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

    International Nuclear Information System (INIS)

    Kim, Won Sam; Kim, Dae Whan; Hwang, Seong Sik

    2007-01-01

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

  17. Void-assisted grain boundary migration in ion-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Vaidya, W.V.

    1983-01-01

    A number of austenitic stainless steels (15 wt% Cr-15 wt% Ni) were irradiated in the solution-annealed condition with 46 MeV Nisup(6+)-ions to a dose-level of 64 dpa at 848 K. Though the microstructure was initially well-equilibrated, under irradiation a general interface migration was observed, the most pronounced being at grain boundaries followed by that at incoherent and even at coherent twins. Changes at the migrating interfaces, features of the migration and variations in the near grain boundary voidage are described. After considering various possibilities which might have caused the migration, it is shown that the observed migration was void-assisted. This has led to the conclusion that voids by nature do not constitute an obstacle for the migrating interface but on the contrary, they offer driving force. Therefore, migration becomes feasible even in the solution-annealed specimens in which inherently there should be a least tendency for such a migration. (orig.)

  18. Void-assisted grain boundary migration in ion-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Vaidya, W.V.

    1983-01-01

    A number of austenitic stainless steels (15 wt% Cr-15 wt% Ni) were irradiated in the solution-annealed condition with 46 MeV Ni 6+ -ions to a dose-level of 64 dpa at 848 K. Though the microstructure was initially well-equilibrated, under irradiation a general interface migration was observed, the most pronounced being at grain boundaries followed by that at incoherent and even at coherent twins. Changes at the migrating interfaces, features of the migration and variations in the near grain boundary voidage are described. After considering various possibilities which might have caused the migration, it is shown that the observed migration was void-assisted. This has led to the conclusion that voids by nature do not constitute an obstacle for the migrating interface but on the contrary, they offer driving force. Therefore, migration becomes feasible even in the solution-annealed specimens in which inherently there should be a least tendency for such a migration. (orig.)

  19. 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

  20. 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

  1. Evaluation of strain-rate sensitivity of ion-irradiated austenitic steel using strain-rate jump nanoindentation tests

    Energy Technology Data Exchange (ETDEWEB)

    Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Hamaguchi, Dai; Ando, Masami; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

    2016-11-01

    Highlights: • We examined strain-rate jump nanoindentation on ion-irradiated stainless steel. • We observed irradiation hardening of the ion-irradiated stainless steel. • We found that strain-rate sensitivity parameter was slightly decreased after the ion-irradiation. - Abstract: The present study investigated strain-rate sensitivity (SRS) of a single crystal Fe–15Cr–20Ni austenitic steel before and after 10.5 MeV Fe{sup 3+} ion-irradiation up to 10 dpa at 300 °C using a strain-rate jump (SRJ) nanoindentation test. It was found that the SRJ nanoindentation test is suitable for evaluating the SRS at strain-rates from 0.001 to 0.2 s{sup −1}. Indentation size effect was observed for depth dependence of nanoindentation hardness but not the SRS. The ion-irradiation increased the hardness at the shallow depth region but decreased the SRS slightly.

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

    International Nuclear Information System (INIS)

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

    1997-09-01

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

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

    International Nuclear Information System (INIS)

    Lee, E.H.

    2001-01-01

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

  4. Modification of the Strength Anisotropy in an Austenitic ODS Steel

    International Nuclear Information System (INIS)

    Kim, T. K.; Jang, J.; Kim, S. H.; Lee, C. B.; Bae, C. S.; Kim, D. H.

    2007-01-01

    Among many candidate alloys for Gen IV reactors, the oxide dispersion strengthened (ODS) alloy is widely considered as a good candidate material for the in-reactor component, like cladding tube. The ODS alloy is well known due to its good high temperature strength, and excellent irradiation resistance. For the previous two decades in the nuclear community, the ODS alloy developments have been mostly focused on the ferritic martensitic (F-M) steel-based ones. On the other hand, the austenitic stainless steels (e.g. 316L or 316LN) have been used as a structural material due to its good high temperature strength and a good compatibility with a media. However, the austenitic stainless steel showed unfavorable characteristics in the dimensional stability under neutron irradiation and cracking behavior with the media. It is thus expected that the austenitic ODS steels restrain the dimension stability under neutron irradiation. However, the ODS alloys usually reveal the anisotropic characteristic in mechanical strength in the hoop and longitudinal directions, which is attributed to the grain morphology strongly developed parallel to the rolling direction with a high aspect ratio. This study focuses on a modification of the strength anisotropy of an austenitic ODS alloy by a recrystallization heat treatment

  5. Synergistic enhancing effect of N+C alloying on cyclic deformation behaviors in austenitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Kang, J. [State Key Laboratory of Metastable Materials Science and Technology, Yanshan University, Qinhuangdao 066004 (China); Zhang, F.C., E-mail: zfc@ysu.edu.cn [State Key Laboratory of Metastable Materials Science and Technology, Yanshan University, Qinhuangdao 066004 (China); Long, X.Y. [State Key Laboratory of Metastable Materials Science and Technology, Yanshan University, Qinhuangdao 066004 (China); Yang, Z.N. [National Engineering Research Center for Equipment and Technology of Cold Strip Rolling, Yanshan University, Qinhuangdao 066004 (China)

    2014-07-29

    Cyclic plastic and elastic strain controlled deformation behaviors of Mn18Cr7 austenitic steel with N0.6C0.3 synergistic enhancing alloying have been investigated using tension-compression low cycle fatigue and three-point bending high cycle fatigue testing. Results of cyclic deformation characteristic and fatigue damage mechanism have been compared to that in Mn12C1.2 steel. Mn18Cr7N0.6C0.3 steel always shows cyclic softening caused by enhanced planar sliding due to the interaction between N+C and the substitutional atoms as well as the dislocation, which is totally different from cyclic hardening in Mn12C1.2 steel caused by the interaction between C members of C–Mn couples with the dislocation. Enhanced effective stress is obtained due to the solid solution strengthening effect caused by the short range order at low strain amplitude while this effect does not work at high strain amplitude. Internal stress contributes most to the cyclic softening with the increase of strain amplitudes. Significant planar slip characteristic can be observed resulting from low stacking fault energy and high short range order effects in Mn18Cr7N0.6C0.3 steel and finally the parallel or intersecting thin sheets with dislocation tangles separated by dislocation free sheets are obtained with the prolonged cycles under cyclic elastic or plastic strain controlled fatigue testing. There exist amounts of small cracks on the surface of the Mn18Cr7N0.6C0.3 steel because fatigue crack initiation is promoted by the cyclic plastic strain localization. However, the zigzag configuration of the cracks reveals that the fatigue crack propagation is highly inhibited by the planar slip characteristic, which eventually improves the fatigue life.

  6. The use of potential drop techniques for the evaluation of environment assisted cracking of austenitic alloys

    International Nuclear Information System (INIS)

    Lidar, P.; Hwang, I.S.; Ballinger, R.G.

    1992-01-01

    The DC potential drop technique has been adapted for strain measurements. An AC potential drop technique has been developed for crack detection. It has been demonstrated that a DC with switched polarity can have a strain accuracy of ±0.12% after yielding and a multi-frequency AC system with up to 200 kHz frequency can have a sensitivity of 50 μm for crack detection. The techniques have been applied to environment assisted cracking tests of Ni-Cr-Fe alloys in 350 degrees C water and austenitic stainless steel in 288 degrees C water using both static and dynamic loading. Long term signal stability is maintained given: (1) rigid probe attachments; (2) local preamplification; and (3) adequate lead grounding and shielding. The AC potential drop technique is found to be more suitable for constant load than for dynamic loading and is also compatible with aggressive environments. During dynamic loading, sensitivity can be significantly reduced due to serrated yielding. An analytical model has been developed to predict the DC and AC potential drops in a round bar geometry. A sensitivity analysis has been made to determine test to test variation associated with variations in supplied current, specimen dimensions, and probe spacing. Multifrequency analysis shows that the measured data agrees with the prediction at frequencies up to about 100 kHz. Above 100 kHz induced signal in the probes results in increases in potential drop and phase angle. The induced signal is reproducible and therefore may be related to the crack mouth opening displacement

  7. 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)

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

    International Nuclear Information System (INIS)

    Kenik, E.A.

    1997-01-01

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

  9. 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.

  10. 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)

  11. 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)

  12. 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)

  13. 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

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

    International Nuclear Information System (INIS)

    Yoshimoto, A.

    1988-08-01

    Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 10 22 n/cm 2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 650 0 C. (author) [pt

  15. 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.

  16. 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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, A.V., E-mail: Alexey.V.Subbotin@gmail.com [Scientific and Production Complex Atomtechnoprom, Moscow 119180 (Russian Federation); Panyukov, S.V., E-mail: panyukov@lpi.ru [PN Lebedev Physics Institute, Russian Academy of Sciences, Moscow 117924 (Russian Federation)

    2016-08-15

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

  19. Effect of laser and/or electron beam irradiation on void swelling in SUS316L austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Subing [School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Yang, Zhanbing, E-mail: yangzhanbing@ustb.edu.cn [School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); State Key Laboratory of Advanced Metallurgy, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Hui [School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Watanabe, Seiichi; Shibayama, Tamaki [Center for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University, Sapporo, Hokkaido 060-8628 (Japan)

    2017-05-15

    Large amounts of void swelling still limit the application of austenitic stainless steels in nuclear reactors due to radiation-induced lattice point defects. In this study, laser and/or beam irradiation was conducted in a temperature range of 573–773 K to explore the suppression of void swelling. The results show that during sequential laser-electron beam irradiation, the void nucleation is enhanced because of the vacancy clusters and void nuclei formed under pre-laser irradiation, causing greater void swelling than single electron beam irradiation. However, simultaneous laser-electron dual-beam irradiation exhibits an obvious suppression effect on void swelling due to the enhanced recombination between interstitials and vacancies in the temperature range of 573–773 K; especially at 723 K, the swelling under simultaneous dual-beam irradiation is 0.031% which is only 22% of the swelling under electron beam irradiation (0.137%). These results provide new insight into the suppression of void swelling during irradiation. - Highlights: •The temperature dependence of void swelling under simultaneous laser-electron dual-beam irradiation has been investigated. •Pre-laser irradiation enhances void nucleation at temperatures from 573 K to 773 K. •Simultaneous laser-electron dual-beam irradiation suppresses void swelling in the temperature range of 573–773 K.

  20. On abnormal decomposition of supercooled austenite in carbon and alloy steels

    International Nuclear Information System (INIS)

    Parusov, V.V.; Dolzhenkov, I.I.; Podobedov, L.V.; Vakulenko, I.A.

    1980-01-01

    Residual stresses which appear as a result of thermal cycling in the temperature range of 300-700 deg C are investigated in an austenitic class steel (03Kh18N11) to ground the assumption on the effect of plastic deformation, appearing due to thermal stresses, on the mechanism of supercooled austenite decomposition. The determination of residual stresses is carried out with the help of X-ray diffraction analysis. It is established that the deformation brings about an increase in density of dislocation the interaction of which leads to the formation of a typical austenite substructure which conditions the proceeding of the eutectoid transformation according to an abnormal mechanism. It is noted, that the grain pearlite formation due to plastic and microplastic deformation of supercooled austenite induced by thermal stresses should be taken into account when developing steel heat treatment shedules [ru

  1. 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)

  2. Fracture toughness of irradiated stainless steel alloys

    International Nuclear Information System (INIS)

    Mills, W.J.

    1986-01-01

    The postirradiation fracture toughness responses of Types 316 and 304 stainless steel (SS) wrought products, cast CF8 SS and Type 308 SS weld deposit were characterized at 427 0 C using J/sub R/-curve techniques. Fast-neutron irradiation of these alloys caused an order of magnitude reduction in J/sub c/ and two orders of magnitude reduction in tearing modulus at neutron exposures above 10 dpa, where radiation-induced losses in toughness appeared to saturate. Saturation J/sub c/ values for the wrought materials ranged from 28 to 31 kJ/m 2 ; the weld exhibited a saturation level of 11 kJ/m 2 . Maximum allowable flaw sizes for highly irradiated stainless steel components stressed to 90% of the unirradiated yield strength are on the order of 3 cm for the wrought material and 1 cm for the weld. Electron fractographic examination revealed that irradiation displacement damage brought about a transition from ductile microvoid coalescence to channel fracture, associated with local separation along planar deformation bands. The lower saturation toughness value for the weld relative to that for the wrought products was attributed to local failure of ferrite particles ahead of the advancing crack which prematurely initiated channel fracture

  3. 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

  4. The effect of alloying and treatment on martensite transformation during deformation in Fe-Cr-Mn steels with unstable austenite

    International Nuclear Information System (INIS)

    Malinov, L.S.; Konop, V.I.; Sokolov, K.N.

    1977-01-01

    The effect is studied of alloying with chromium (6-10%), silicon (1-2%), molybdenum (1-3%), and copper (2%), the heat treatment conditions, and the deformation conditions, or the martensitic transformation and mechanical properties of Fe-Cr-Mn steels of the transitional class based on 0G8AM2S. It is shown that appropriate alloying and treatment, taking into account the degree of stability of the austenite, can ensure a complex of high mechanical properties of the steels investigated. For instance, the treatment of steel 0Kh10AG8MD2S by the technique: hardening+ 40% deformation at 400 deg C + 10% deformation at room temperature has yielded the following mechanical properties: sigmasub(B)=150 kgf/mm 2 , sigmasub(T)=110 kgf/mm 2 , sigma=18%, psi=32%

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

    Institute of Scientific and Technical Information of China (English)

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

    2003-01-01

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

  6. 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

  7. Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

    Science.gov (United States)

    Gussev, M. N.; Cakmak, E.; Field, K. G.

    2018-06-01

    Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

  8. 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

  9. Irradiation-assisted stress corrosion cracking in HTH Alloy X-750 and Alloy 625

    International Nuclear Information System (INIS)

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-01-01

    In-reactor testing of bolt-loaded compact tension specimens was performed in 360 C water to determine the irradiation-assisted stress corrosion cracking (IASCC) behavior of HTH Alloy X-750 and direct-aged Alloy 625. New data confirm previous results showing that high irradiation levels reduce SCC resistance in Alloy X-750. Heat-to-heat variability correlates with boron content, with low boron heats showing improved IASCC properties. Alloy 625 is resistant to IASCC, as no cracking was observed in any Alloy 625 specimens. Microstructural, microchemical and deformation studies were performed to characterize the mechanisms responsible for IASCC in Alloy X-750 and the lack of an effect in Alloy 625. The mechanisms under investigation are: boron transmutation effects, radiation-induced changes in microstructure and deformation characteristics, and radiation-induced segregation. Irradiation of Alloy X-750 caused significant strengthening and ductility loss that was associated with the formation of cavities and dislocation loops. High irradiation levels did not cause significant segregation of alloying or trace elements in Alloy X-750. Irradiation of Alloy 625 resulted in the formation of small dislocation loops and a fine body-centered-orthorhombic phase. The strengthening due to the loops and precipitates was apparently offset by a partial dissolution of γ double-prime precipitates, as Alloy 625 showed no irradiation-induced strengthening or ductility loss. In the nonirradiated condition, an IASCC susceptible HTH heat containing 28 ppm B showed grain boundary segregation of boron, whereas a nonsusceptible HTH heat containing 2 ppm B and Alloy 625 with 20 ppm B did not show significant boron segregation. Transmutation of boron to helium at grain boundaries, coupled with matrix strengthening, is believed to be responsible for IASCC in Alloy X-750, and the absence of these two effects results in the superior IASCC resistance displayed by Alloy 625

  10. Proton irradiation studies on Al and Al5083 alloy

    Science.gov (United States)

    Bhattacharyya, P.; Gayathri, N.; Bhattacharya, M.; Gupta, A. Dutta; Sarkar, Apu; Dhar, S.; Mitra, M. K.; Mukherjee, P.

    2017-10-01

    The change in the microstructural parameters and microhardness values in 6.5 MeV proton irradiated pure Al and Al5083 alloy samples have been evaluated using different model based techniques of X-ray diffraction Line Profile Analysis (XRD) and microindendation techniques. The detailed line profile analysis of the XRD data showed that the domain size increases and saturates with irradiation dose both in the case of Al and Al5083 alloy. The corresponding microstrain values did not show any change with irradiation dose in the case of the pure Al but showed an increase at higher irradiation doses in the case of Al5083 alloy. The microindendation results showed that unirradiated Al5083 alloy has higher hardness value compared to that of unirradiated pure Al. The hardness increased marginally with irradiation dose in the case of Al5083, whereas for pure Al, there was no significant change with dose.

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Renault-Laborne, A., E-mail: alexandra.renault@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Garnier, J.; Malaplate, J. [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gavoille, P. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sefta, F. [EDF R& D, MMC, Site des Renardières, F-77818, Morêt-sur-Loing Cedex (France); Tanguy, B. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-07-15

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

  13. 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

  14. 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)

  15. Influence of helium embrittlement on post-irradiation creep rupture behaviour of austenitic and martensitic stainless steels

    International Nuclear Information System (INIS)

    Wassilew, C.

    1982-01-01

    The author has investigated the influence of helium embrittlement on the creep rupture properties of the austenitic stainless steels 1.4970 and 1.4962 and the martensitic stainless steel 1.4914 after irradiation in the BR-2 reactor in Mol, Belgium. The results show that austenitic steels react much more strongly to the embrittlement effect of the helium than do martensitic steels. The causes of the lower embrittlement tendency of the martensitic than of both austenitic stainless steels were analysed carefully. A new embrittlement model was developed on the basis of data derived from the creep rupture experiments, and reinforced by a simple metallographic investigation of the fracture zone and its immediate environment. This model pays specific attention to the role of the twin planes as the most efficient area of increased vacancy production, and takes into account the ability of the twin boundaries to transport these vacancies with reduced energy and low loss into the high-angle grain boundaries. (author)

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

    International Nuclear Information System (INIS)

    1976-01-01

    A specification is provided for an austenitic stainless steel consisting of Fe, Cr and Ni, with small amounts of Mo, Mn, Si and Ti. The specification includes a fuel element and a method for cladding a reactor fuel. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    Science.gov (United States)

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

    2015-09-01

    Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 1019 ions/m2 (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite-austenite phase boundary and presence of M23C6 carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M23C6 carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M23C6 carbides at 350 °C and 400 °C.

  19. Influence of silicon on void nucleation in irradiated alloys

    International Nuclear Information System (INIS)

    Esmailzadeh, B.; Kumar, A.; Garner, F.A.

    1984-01-01

    The addition of silicon to pure nickel, Ni-Cr alloys and Fe-Ni-Cr alloys raises the diffusivity of each of the alloy components. The resultant increase in the effective vacancy diffusion coefficient causes large reductions in the nucleation rate of voids during irradiation. This extends the transient regime of swelling, which is controlled not only by the amount of silicon in solution but also by the precipitation kinetics of precipitates rich in nickel and silicon

  20. 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

  1. Analytical electron microscopy of neutron-irradiated reactor alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1982-01-01

    Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

  2. Development of high nickel austenitic steels for the application to fast reactor cores, (I). Alloy design with the aid of the d-electrons concept

    International Nuclear Information System (INIS)

    Murata, Yoshinori; Morinaga, Masahiko; Yukawa, Natsuo; Ukai, Shigeharu; Nomura, Shigeo; Okuda, Takanari; Harada, Makoto

    1999-01-01

    The design of high nickel austenitic steels for the core materials of the fast reactors was performed following the d-electrons concept devised on the basis of molecular orbital calculations of transition-metal based alloys. In this design two calculated parameters are mainly utilized. The one is the d-orbital energy level (Md) of alloying transition elements, and the other is the bond order (Bo) that is a measure of the covalent bond strength between atoms. Using the Md-bar - Bo-bar phase stability diagram accurate prediction become possible for the phase stability of the austenite phase and 5% swelling at 140 dpa for nickel ions. Here, Md-bar and Bo-bar are the compositional average of Md and Bo parameters, respectively. On the basis of the phase stability diagram and preliminary experiments, guidelines for the alloy design of carbo-nitrides precipitated high nickel austenitic steels were constructed. Following the guidelines several new austenitic steels were designed for the fast reactors core material. (author)

  3. Microstructural and crystallographic characteristics of modulated martensite, non-modulated martensite, and pre-martensitic tweed austenite in Ni-Mn-Ga alloys

    International Nuclear Information System (INIS)

    Zhou, Le; Schneider, Matthew M.; Giri, Anit; Cho, Kyu; Sohn, Yongho

    2017-01-01

    A combinatorial approach using diffusion couples and TEM analyses was carried out to investigate the composition-dependent martensitic transformation in NiMnGa alloys. The compositions cover a large portion of the off-stoichiometric Ni 2 MnGa compositions and some Mn-rich compositions. Crystallographic variations of the martensitic phase, including non-modulated (NM) martensite, modulated (5M or 7M) martensite, and austenitic phase were identified in the diffusion couples and investigated with respect to their microstructure and crystallography. The 5M and 7M martensitic structures were only found near the interphase boundary between austenite and martensite, while the NM martensitic structures were found mostly away from the interphase boundary. The tetragonality ratio (c/a) for NM martensite generally increases with e/a ratio, but was also dependent on the composition. The habit plane and martensitic microstructure that consists of twinned variants with differing orientations were documented using electron diffraction. The pre-martensitic state was observed in the austenitic phase that was located near the interphase boundary between austenite and martensite, with distinctive tweed microstructure and a strain field originating from the local lattice distortions. The combinatorial approach proves to be efficient and systematic in studying the composition-dependent martensitic transformation in NiMnGa alloys and can be potentially applied to other shape memory alloys.

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

    International Nuclear Information System (INIS)

    Enrique, R.A.; Bellon, P.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Enrique, R.A.; Bellon, P.

    1999-01-01

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

  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. Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Eiholzer, C.R.; Toloczko, M.B.; Kumar, A.S.

    1991-11-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550 degrees c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength

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

    International Nuclear Information System (INIS)

    Pokor, C.

    2003-01-01

    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)

  10. Irradiation of aluminium alloy materials with electron beam

    International Nuclear Information System (INIS)

    Konno, Osamu; Masumoto, Kazuyoshi

    1982-01-01

    It is a theme with a room for discussion to employ the stainless steel composed of longer half-life materials for the vacuum system of accelerators, from the viewpoint of radiation exposure. Therefore, it is desirable to use aluminium of shorter half-life in place of stainless steel. As a result of investigation on the above theme in the 1.2 GeV electron linac project in Tohoku University, it has been concluded that aluminium alloy vacuum chambers can reduce exposure dose by about one or two figures as compared with stainless steel ones. Of course, aluminium alloy contains trace amounts of Mg, Si, Ti, Cr, Mn, Fe, Zn, Cu and others. Therefore, four kinds of aluminium alloy considered to be usable have been examined for induced radioactivity by electron beam irradiation. Stainless steel SUS 304 has been also irradiated for comparison. Radiation energy has been 30 MeV and 200 MeV. When stainless steel and aluminium alloy were compared, aluminium alloy was very effective for reducing surface dose in low energy irradiation. In 200 MeV irradiation, the dose ratio of aluminium alloy to stainless steel became 1/30 to 1/100 after one week, though the dose difference between these two materials became smaller in 100 days or more after irradiation. If practical inspection and repair are implemented during the period from a few days to one week after shutdown, the aluminium alloy is preferable for exposure dose reduction even in high energy irradiation. (Wakatsuki, Y.)

  11. 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.

  12. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this obervation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. Measurements of radiation enhanced diffusion are less time consuming and expensive than irradiation creep tests and information on this property can be obtained rather quickly, prior to the selection of stainless steel alloys for creep tests. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. Finally, a few uniaxial tensile creep tests will be performed in fully instrumented rigs. (Auth.)

  13. Influence of alloy elements on physical and mechanical properties of single crystalline austenitic stainless steels

    International Nuclear Information System (INIS)

    Okamoto, Kazutaka; Kaneda, Junya; Yoshinari, Akira; Aono, Yasuhisa

    2000-01-01

    The single crystalline austenitic stainless steels based on 316 L were developed to improve their resistance to intergranular corrosion and stress corrosion cracking. However the mechanical properties of the single crystals were lower than those of polycrystalline. The precipitation hardening methods were applied to the single crystal for the increase of their mechanical strength by addition of niobium and heat treatments. In this paper, the influences of niobium concentration on the several physical and mechanical properties of these single crystalline austenitic stainless steels were studied. The thermal conductivity, coefficients of thermal expansion and elastic constants of the single crystals were almost the same as those of polycrystalline independently of the niobium concentration. The mechanical properties of the single crystals strongly depended on the niobium concentration and the orientation. In the specimen which contains 1.0 mass% niobium, 0.2% proof stress were remarkably improved; 370 MPa, 337 MPa and 403 MPa were obtained in , and orientations at the room temperature. The creep rupture strength and the high cycle fatigue strength were also improved by addition of niobium. In the -orientated specimen which contains 1.0 mass% niobium, the creep rupture strength at 873 K for 103 hours, 245 MPa and the high cycle fatigue strength at 773 K for 107 cycles, 220 MPa were obtained. Furthermore, the single crystalline pipe, bolts and nuts were successfully manufactured for the application of these single crystals. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  15. 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

  16. 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.

  17. Irradiation Stability of Uranium Alloys at High Exposures

    International Nuclear Information System (INIS)

    McDonell, W.R.

    2001-01-01

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

  18. Fatigue micro-crack initiation behavior and effect of irradiation damage on it in austenitic stainless steel

    International Nuclear Information System (INIS)

    Nakai, Ryosuke; Sato, Yuki; Nogami, Shuhei; Hasegawa, Akira

    2012-01-01

    The effect of irradiation on slip band formation and growth and micro-crack initiation behavior under low cycle fatigue in SUS316L austenitic stainless steel was investigated using accelerator-based proton irradiation and a low cycle fatigue test at room temperature in air. The micro-crack initiation was observed at slip band, grain boundary, twin boundary, and triple junction regardless of the total strain range and the proton irradiation. In unirradiated specimens, the micro-crack initiation life dropped by 75-90% due to the increase of the plastic strain range. Under the condition the plastic strain range was 0.4%, the micro-crack initiation was observed mainly at the grain boundary. On the other hand, under the condition the plastic strain range was 1.0%, the number fractions of the micro-crack initiation in slip band and twin boundary were increased. In proton-irradiated specimens, the micro-crack initiation life decreased by 50-80% and the micro-crack initiation was observed mainly at slip band and twin boundary. (author)

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

    Science.gov (United States)

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

    2015-04-01

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

  20. Property change of advanced tungsten alloys due to neutron irradiation

    International Nuclear Information System (INIS)

    Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki

    2013-01-01

    This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  2. Austenitic stainless steel alloys with high nickel contents in high temperature liquid metal systems

    International Nuclear Information System (INIS)

    Konvicka, H.R.; Schwarz, N.F.

    1981-01-01

    Fe-Cr-Ni base alloys (nickel content: from 15 to 70 wt%, Chromium content: 15 wt%, iron: balance) together with stainless steel (W.Nr. 1.4981) have been exposed to flowing liquid sodium at 730 0 C in four intervals up to a cumulative exposure time of 1500 hours. Weight change data and the results of post-exposition microcharacterization of specimens are reported. The corrosion rates increase with increasing nickel content and tend to become constant after longer exposure times for each alloy. The corrosion rate of stainless steel is considerably reduced due to the presence of the base alloys. Different kinetics of nickel poor (up to 35% nickel) and nickel rich (> 50% nickel) alloys and nickel transport from nickel rich to nickel poor material is observed. (orig.)

  3. Moessbauer study of amorphous alloys irradiated with energetic heavy ions

    International Nuclear Information System (INIS)

    Kuzmann, E.; Spirov, I.N.

    1984-01-01

    The Moessbauer spectroscopy was applied to study radiation damages in amorphous alloys irradiated with 40 Ar (E=225 MeV) or 132 Xe (E=120 MeV) ions at room temperature. In the magnetically splitted Moessbauer spectra the dose-dependent decreases of the intensity of the 2nd and 5th lines as well as of the average hyperfine magnetic field were observed. The changes weAe also analysed using the hyperfine field distribution obtained from the spectra. The results are interpreted in terms of defect creation and structural changes of shortrange order of irradiated amorphoys alloys

  4. 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

  5. Ductility behavior of irradiated path B alloys

    International Nuclear Information System (INIS)

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

    1983-01-01

    The objective of this study was to assess the practicality of using five Path B alloys in their current form as structural materials in the Fusion First-Wall/Blanket by evaluating both their postirradiation ductility and the corresponding microstructures

  6. 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

  7. 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.)

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

    International Nuclear Information System (INIS)

    Barbu, A.; Ardell, A.J.

    1975-07-01

    The microstructures of Ni + ion-irradiated Ni--Si solid-solution alloys, containing 2, 4, 6 and 8 at. percent Si were investigated as a function of dose, dose-rate, and temperature. Results of transmission electron microscopy and other data show the precipitation of γ' (Ni 3 Si) in all samples irradiated at 500 0 C. Characteristics of the precipitates are described and a mechanism for their formation is suggested. (U.S.)

  9. Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

    International Nuclear Information System (INIS)

    Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

    2011-01-01

    The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

  10. Study of irradiation creep of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

    Elthalabawy, Waled M.; Khan, Tahir I.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Bajaj, R.; Mills, W.J.; Lebo, M.R.; Hyatt, B.Z.; Burke, M.G.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Mills, W.J.; Lebo, M.R.; Bajaj, R.; Kearns, J.J.; Hoffman, R.C.; Korinko, J.J.

    1994-01-01

    In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360 degree C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K 1 and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750

  14. Martensite shear phase reversion-induced nanograined/ultrafine-grained Fe-16Cr-10Ni alloy: The effect of interstitial alloying elements and degree of austenite stability on phase reversion

    Energy Technology Data Exchange (ETDEWEB)

    Misra, R.D.K., E-mail: dmisra@louisiana.edu [Center for Structural and Functional Materials, University of Louisiana at Lafayette, Madison Hall Room 217, P.O. Box 44130, Lafayette, LA 70504-1430 (United States); Zhang, Z.; Venkatasurya, P.K.C. [Center for Structural and Functional Materials, University of Louisiana at Lafayette, Madison Hall Room 217, P.O. Box 44130, Lafayette, LA 70504-1430 (United States); Somani, M.C.; Karjalainen, L.P. [Department of Mechanical Engineering, University of Oulu, P.O. Box 4200, Oulu 90014 (Finland)

    2010-11-15

    Research highlights: {yields} Development of a novel process involving phase-reversion annealing process. {yields} Austensite stability strongly influences development of nanograined structure. {yields} Interstitial elements influence microstructural evolution during annealing. - Abstract: We describe here an electron microscopy study of microstructural evolution associated with martensitic shear phase reversion-induced nanograined/ultrafine-grained (NG/UFG) structure in an experimental Fe-16Cr-10Ni alloy with very low interstitial content. The primary objective is to understand and obtain fundamental insights on the influence of degree of austenite stability (Fe-16Cr-10Ni, 301LN, and 301 have different austenite stability index) and interstitial elements (carbon and nitrogen) in terms of phase reversion process, microstructural evolution during reversion annealing, and temperature-time annealing sequence. A relative comparison of Fe-16Cr-10Ni alloy with 301LN and 301 austenitic stainless steels indicated that phase reversion in Fe-16Cr-10Ni occurred by shear mechanism, which is similar to that observed for 301, but is different from the diffusional mechanism in 301LN steel. While the phase reversion in the experimental Fe-16Cr-10Ni alloy and 301 austenitic stainless steel occurred by shear mechanism, there were fundamental differences between these two alloys. The reversed strain-free austenite grains in Fe-16Cr-10Ni alloy were characterized by nearly same crystallographic orientation, where as in 301 steel there was evidence of break-up of martensite laths during reversion annealing resulting in several regions of misoriented austenite grains in 301 steel. Furthermore, a higher phase reversion annealing temperature range (800-900 deg. C) was required to obtain a fully NG/UFG structure of grain size 200-600 nm. The difference in the phase reversion and the temperature-time sequence in the three stages is explained in terms of Gibbs free energy change that

  15. 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.

  16. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-01

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

  17. Neutron irradiation 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

  18. Post-irradiation annealing of coarse-grained model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ray, P H.N.; Wilson, C; McElroy, R J [AEA Reactor Services, Harwell (United Kingdom)

    1994-12-31

    Thermal ageing and irradiation studies have been carried out on three model alloys (JPC, JPB, JPG) that have identical compositions except for different levels of phosphorus and/or copper. They have been irradiated in three conditions, as-received, heat treated to produce a coarse grained microstructure (similar to heat-affected-zone), and in this condition further aged at 450 C to produce a temper embrittled condition. One of the alloy have been subject to a post-irradiation anneal. The effect of these treatments on mechanical property changes has been characterized by Charpy testing and Vickers hardness measurements; the phosphorus segregation has been studied by a combination of STEM and Auger techniques.

  19. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

    Loelgen, R.H.; Cundy, M.R.; Schuele, W.

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

  20. 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

  1. 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)

  2. 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.

  3. Structural transformations in austenitic stainless steel induced by deuterium implantation: irradiation at 100 K.

    Science.gov (United States)

    Morozov, Oleksandr; Zhurba, Volodymyr; Neklyudov, Ivan; Mats, Oleksandr; Rud, Aleksandr; Chernyak, Nikolay; Progolaieva, Viktoria

    2015-01-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic stainless steel 18Cr10NiTi preimplanted at 100 K with deuterium ions in the dose range from 3 × 10(15) to 5 × 10(18) D/cm(2). The kinetics of structural transformation development in the implantation steel layer was traced from deuterium thermodesorption spectra as a function of implanted deuterium concentration. At saturation of austenitic stainless steel 18Cr10NiTi with deuterium by means of ion implantation, structural-phase changes take place, depending on the dose of implanted deuterium. The maximum attainable concentration of deuterium in steel is C = 1 (at.D/at.met. = 1/1). The increase in the implanted dose of deuterium is accompanied by the increase in the retained deuterium content, and as soon as the deuterium concentration attains C ≈ 0.5 the process of shear martensitic structural transformation in steel takes place. It includes the formation of bands, body-centered cubic (bcc) crystal structure, and the ferromagnetic phase. Upon reaching the deuterium concentration C > 0.5, the presence of these molecules causes shear martensitic structural transformations in the steel, which include the formation of characteristic bands, bcc crystal structure, and the ferromagnetic phase. At C ≥ 0.5, two hydride phases are formed in the steel, the decay temperatures of which are 240 and 275 K. The hydride phases are formed in the bcc structure resulting from the martensitic structural transformation in steel.

  4. 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.

  5. 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])

  6. 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

  7. Irradiation induced precipitation in tungsten based, W-Re alloys

    Science.gov (United States)

    Williams, R. K.; Wiffen, F. W.; Bentley, J.; Stiegler, J. O.

    1983-03-01

    Tungsten-base alloys containing 5, 11, and 25 pct Re were irradiated in the EBR-II reactor. Irradiation temperatures ranged from 600 to 1500 °C. All compositions were irradiated to fluences in the range 4.3 to 6.1 X 1025 n/m2 (E > 0.1 MeV), and three 25 pct Re samples were also irradiated to 3.7 X 1026 n/m2 at temperatures 700 to 900 °C. Postirradiation examination included measurement of electrical resistivity at room temperature and lower temperatures, X-ray diffraction, optical metallography, microprobe analysis, and transmission electron microscopy. Irradiation induced resistivity decreases observed in most of the samples suggested second-phase precipitation. Complete results confirmed the precipitate formation in all samples, in disagreement with existing phase diagrams for the W-Re system. Electron diffraction showed the precipitates to be consistent with the cubic, Re-rich X-phase and inconsistent with the σ-phase. Large variations in precipitate morphology and distribution were observed between the different compositions and irradiation conditions. For the 5 and 11 pct Re-alloys, spherically symmetric strain fields surrounded the equiaxed precipitate particles, and were observed even where no particles were visible. These strain fields are believed to arise from local Re enrichment. Thermoelectric data show that the precipitation can lead to decalibration of W/Re thermocouples.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gussev, Maxim N. [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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  10. Measurement of carbon activity in sodium by Fe-Mn 20% alloy, and by strainless austenitic steel 304L and 316L

    International Nuclear Information System (INIS)

    Oberlin, C.; Saint Paul, P.; Baque, P.; Champeix, L.

    1980-01-01

    Precise knowledge of carbon activity in sodium used as coolant in fast breeder reactors, is essential for continuous survey of carburization-decarburization processes. Carbon activity can be periodically surveyed by measuring the carbon concentration or by hot trap like metal alloy strip placed in sodium loop. In fact, in equilibrium, activity of carbon in sodium is equal to the activity in metal alloy. Thus if the relation between concentration of carbon and it activity in the alloy is known, it is possible to estimate the activity of carbon in sodium. Materials to be used should have high solubility in carbon at the needed temperature. They should quickly attain equilibrium with sodium and they should not contain impurities that can affect the results. Materials chosen according to these criteria were Fe-Mn 20%, stainless austenitic steel AISI 304L and 316L

  11. Swelling in neutron irradiated nickel-base alloys

    International Nuclear Information System (INIS)

    Brager, H.R.; Bell, W.L.

    1972-01-01

    Inconel 625, Incoloy 800 and Hastelloy X were neutron irradiated at 500 to 700 0 C. It was found that of the three alloys investigated, Inconel 625 offers the greatest swelling resistance. The superior swelling resistance of Inconel 625 relative to that of Hastelloy-X is probably related to differences in the concentrations of the minor rather than major alloy constituents, and can involve (a) enhanced recombination of defects in the Inconel 625 and (b) preferential attraction of vacancies to incoherent precipitates. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhijie, E-mail: zjiao@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Was, Gary [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Miura, Terumitsu; Fukuya, Koji [Institute of Nuclear Safety System (INSS) (Japan)

    2014-09-15

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

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

    Science.gov (United States)

    2012-03-01

    stainless steel and ferritic/ martensitic steel can vary from structural and support components in the reactor core to reactor fuel...of ferritic/ martensitic steels compared to type 316 stainless steel after irradiation in Experimental Breeder Reactor-II at 420 ºC to ~80dpa (From...ferritic martensitic steel at Sandia National Laboratories. The 316 stainless steel had a certified composition of:

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Shigeki, E-mail: kasahara.shigeki@jaea.go.jp [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kitsunai, Yuji [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Chimi, Yasuhiro [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Chatani, Kazuhiro; Koshiishi, Masato [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Nishiyama, Yutaka [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2016-11-15

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

  17. 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

  18. Irradiation response in titanium modified austenitic stainless steels prepared by rapid solidification processing. Pt. 3

    International Nuclear Information System (INIS)

    Imeson, D.; Tong, C.H.; Parker, C.A.; Vander Sande, J.B.; Grant, N.J.; Harling, O.K.; Massachusetts Inst. of Tech., Cambridge

    1984-01-01

    Titanium carbide precipitation on dislocations during irradiation and recoil-induced particle dissolution are considered. The outline analysis given indicates that complete swelling suppression may occur in favorable conditions due to a counterbalancing of the effective dislocation interstitial bias. The behavior is, however, not stable against a return to normal swelling levels for type 316 steels. A model is presented which may serve as a basis for the interpretation of some aspects of the irradiation response in this system. (orig.)

  19. 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

  20. Oxidation behavior of austenitic iron-base ODS alloy in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Dong, Z.; Zahiri, R.; Kohandehghan, A.; Mitlin, D., E-mail: behnamia@ualberta.ca, E-mail: zdong@ualberta.ca, E-mail: kohandeh@ualberta.ca, E-mail: rzahiris@ualberta.ca, E-mail: dave.mitlin@ualberta.ca [Univ. of Alberta, Edmondon, AB (Canada); Zhou, Z., E-mail: zhouzhj@mater.ustb.edu.cn [Univ. of Science and Tech. Beijing, Beijing (China); Chen, W.; Luo, J., E-mail: weixing.chen@ualberta.ca, E-mail: Jingli.luo@ualberta.ca [Univ. of Alberta, Edmonton, AB (Canada); Zheng, W., E-mail: wenyue@nrcan.gc.ca [Natural Resources Canada, Canmet MATERIALS, Hamilton, ON (Canada); Guzonas, D. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    In this study, the effect of exposure time on the corrosion of the 304 stainless steel based oxide dispersion strengthened alloy, SS304ODS, in supercritical water was investigated at 650 {sup o}C with constant dissolved oxygen concentration. The results show that the oxidation of SS304ODS in supercritical water followed a parabolic law at 650 {sup o}C. Discontinuous oxide scale with two distinct layers has formed after 550 hours. The inner layer was chromium-rich while the outer layer was iron-rich (Magnetite). The oxide islands grow with increasing the exposure time. With increasing exposure time, the quantity of oxide islands increased in which major preferential growth along oxide-substrate interface was observed. The possible mechanism of SS304ODS oxidation in supercritical water was also discussed. (author)

  1. Segregation of a copper-nickel alloy after electron irradiation

    International Nuclear Information System (INIS)

    Wagner, W.

    1979-09-01

    In the present work measurement of diffuse neutron scattering are used to determine short range segregation effects of the alloy Cu 0 sub(.) 414 Ni 0 sub(.) 586 after thermal annealing and 3 MeV electron irradiation in the temperature range between 370 K and 600 K. In addition neutron small angle scattering measurement are performed after irradiation to study possible long range segregation effects. Residual resistivity measurements are performed in parallel in order tp orientate the relatively expensive neutron scattering measurements with respect to the residual changes (orig./KBI) [de

  2. Fracture toughness of ferritic alloys irradiated at FFTF

    International Nuclear Information System (INIS)

    Huang, F.H.

    1986-05-01

    Ferritic compact tension specimens loaded in the Material Open Test Assembly (MOTA) for irradiation during FFTF Cycle 4 were tested at temperatures ranging from room temperature to 428/degree/C. The electrical potential single specimen method was used to measure the fracture toughness of the specimens. Results showed that the fracture toughness of both HT-9 and 9Cr-1Mo decreases with increasing test temperature and that the toughness of HT-9 was about 30% higher than that of 9Cr-1Mo. In addition, increasing irradiation temperature resulted in an increase in tearing modulus for both alloys. 4 refs., 5 figs., 1 tab

  3. 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.

  4. 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

  5. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging*1

    Science.gov (United States)

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-07-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6C in JPCA-2.

  6. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

    International Nuclear Information System (INIS)

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-01-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1x10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6 C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6 C in JPCA-2. (orig.)

  7. Development plan of austenitic Fe and Ni based alloys with improved corrosion resistance to sulfuric acid and HI fluids of industrial processes

    International Nuclear Information System (INIS)

    Hirota, Noriaki; Iwatsuki, Jin; Imai, Yoshiyuki; Yan, Xing L.

    2017-12-01

    In this study, austenitic Fe based alloys and Ni based alloys was developed as candidate structural materials for equipment operated in sulfuric acid and hydrogen iodide (HI) environment, which exists in various industrial processes including iodine-sulfur (IS) hydrogen production process and geothermal power generation process. The objectives of the study are to achieve the corrosion resistance performance sufficient under the working condition of these processes and to overcome the practical scale-up difficulty of the ceramic (SiC) material that is presently used in the processes due to the manufacturing size limitation of the ceramic. The chemical composition development plan for the austenitic Fe based alloys is threefold: reinforcement of matrix by addition of Cu and Ta, strength compensation of the surface film by addition of Si and Ti, and prevention of peeling of surface oxide by addition of rare earth elements. Because addition of Cu and Si is known to reduce the ductility of the material and thus manufacturability of the component, it is important to determine the allowable amount of each element to be added. On the other hand, the chemical composition development plan for the Ni based alloys is reinforcement of matrix by addition of Mo, W and Ta, strength compensation of the surface film by addition of Ti, and prevention of peeling of surface oxide by addition of rare earth elements. In particular, the addition of Mo and W to the Ni based alloy is expected to be effective in preventing dimensional deviation of structures from increasing during heating and cooling of process equipment. Various material specimens will be fabricated based on the above chemical composition development plans and tests on these specimens will then be carried out to confirm the corrosion resistance performance under the fluid conditions simulating each industrial process. (author)

  8. Irradiation testing of high density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

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

    1997-10-01

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

  9. 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)

  10. Microstructural examination of irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  11. 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.)

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  14. 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

  15. 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

  16. Radiation damage of austenitic stainless steels and zirconium alloys; Pregled radijacionog ostecenja austenitnih nerdjajucih celika i legura cirkonijuma

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This review contains analyses of available data concerning texture deformations and radiation damage of zirconium and zircaloy-2; radiation damage, influence of neutron radiation on the mechanical properties of austenitic, ferritic and other types of stainless steels.

  17. 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

  18. Enhanced low-temperature oxidation of zirconium alloys under irradiation

    International Nuclear Information System (INIS)

    Cox, B.; Fidleris, V.

    1989-01-01

    The linear growth of relatively thick (>300 nm) interference-colored oxide films on zirconium alloy specimens exposed in the Advanced Test Reactor (ATR) coolant at ≤55 o C was unexpected. Initial ideas were that this was a photoconduction effect. Experiments to study photoconduction in thin anodic zirconium oxide (ZrO 2 ) films in the laboratory were initiated to provide background data. It was found that, in the laboratory, provided a high electric field was maintained across the oxide during ultraviolet (UV) irradiation, enhanced growth of oxide occurred in the irradiated area. Similarly enhanced growth could be obtained on thin thermally formed oxide films that were immersed in an electrolyte with a high electric field superimposed. This enhanced growth was found to be caused by the development of porosity in the barrier oxide layer by an enhanced local dissolution and reprecipitation process during UV irradiation. Similar porosity was observed in the oxide films on the ATR specimens. Since it is not thought that a high electric field could have been present in this instance, localized dissolution of fast-neutron primary recoil tracks may be the operative mechanism. In all instances, the specimens attempt to maintain the normal barrier-layer oxide thickness, which causes the additional oxide growth. Similar mechanisms may have operated during the formation of thick loosely adherent, porous oxides in homogeneous reactor solutions under irradiation, and may be the cause of enhanced oxidation of zirconium alloys in high-temperature water-cooled reactors in some water chemistries. (author)

  19. Pre-irradiation tests on U-Si alloys

    International Nuclear Information System (INIS)

    Howe, L.M.; Bell, L.G.

    1958-05-01

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

  20. Ordering of equiatomic Fe-Ni alloy under irradiation

    International Nuclear Information System (INIS)

    Penisson, J.M.; Bourret, A.

    1975-01-01

    The interpretations of many kinetic studies on the ordering of alloys by heat treatments use models based on chemical kinetic equations or statistical models. In these models it is assumed that ordering takes place by a hole process. When ordering occurs under irradiation a strong supersaturation in interstitials and holes exists and the question is to know what defects take part in the ordering process. Thus in the order-disorder transformation in Fe-Ni demonstrated after neutron irradiation it is assumed that only the holes participate in the ordering. To check this hypothesis the above transformation was observed during electron irradiation in an electron microscope at 1MeV. Owing to the strong flux available the ordering takes place fairly quickly ( [fr

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

    International Nuclear Information System (INIS)

    Pettersson, K.; Hallstadius, L.; Bergqvist, H.; Nylund, A.; Wikstroem, C.

    1992-01-01

    Ten different alloys of zirconium have been tested with regard to the effect of irradiation on their mechanical properties and their sensitivity to environmentally induced failure. Two different environments were used: iodine vapour and liquid cesium with an addition of 2% cadmium. The neutron dose was 10 21 n/cm 2 (E>1MeV) and the irradiation temperature was about 300 degrees C. All alloy additions increased the irradiation hardening. Especially notable was the large effect of titanium and tin on irradiation hardening. A limited amount of transmission electron microscopy was carried out in order to find an explanation to the effects. The testing in different environments showed that there is no clear correlation between environmental sensitivity and yield stress. For materials of similar yield stress an alloyed material tends to be more sensitive to environmental cracking than a material which only contains oxygen as an impurity. There also seems to be an effect of oxygen on the environmental cracking sensitivity. A material with 910 ppm oxygen was considerably more sensitive to cracking than a material with 470 ppm oxygen despite the fact that the yield stress values differed by only 90 MPa

  2. 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.

  3. 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

  4. Fusion neutron irradiation of Ni-Si alloys at high temperature*1

    Science.gov (United States)

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

    1988-07-01

    Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been 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 6-7% 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 6-7% for the fusion neutron irradiated sample.

  5. 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

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

    International Nuclear Information System (INIS)

    Hofmann, P.

    1994-08-01

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

  7. 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

  8. Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

    Science.gov (United States)

    Mansur, L. K.; Grossbeck, M. L.

    1988-07-01

    Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

  9. 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

  10. Microstructural development for copper alloys irradiated in RTNS-11

    International Nuclear Information System (INIS)

    Doran, D.G.

    1979-01-01

    Microscopy and microhardness measurements were performed on pure Cu and Cu alloyed with 5% of either Al, Ni, or Mn, all irradiated in RTNS-II up to 7.5 x 10 17 n/cm 2 . Results show that a substantial fraction of the defects are below the microscope's resolution limit and account for a large amount of the radiation hardening. The solute Al appears to lead to substantial differences in clustering of point defects within the cascade and thus affect the visibility of the clusters. The fraction of defects surviving the original cascade event is at least 9%

  11. Nucleation of dislocation loops during irradiation in binary FCC alloys with different alloy compositions

    International Nuclear Information System (INIS)

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

    1992-01-01

    Dislocation loop nucleation is analyzed using a rate theory based model for face-centered cubic (fcc) binary alloys containing A- and B-atoms. In order to calculate the nucleation process in concentrated alloys, the model considers three types of interstitial dumbbells composed of A- and B-atoms, AA-, BB-, and AB-type dumbbells. Conversions between these interstitial dumbbells are newly introduced in the formulation in consideration of dumbbell configurations and movements. The model also includes reactions, such as point defect production by irradiation, mutual recombination of an interstitial and a vacancy, and dislocation loop nucleation and growth. Parameter values are chosen based on the atom size of the alloy component elements, and dislocation loop nucleation kinetics are investigated while varying alloy compositions. Two different types of kinetics are obtained in accordance with the dominant loop nucleus type. The migration energy difference of AA- and BB-type interstitial dumbbells is important in the determination of the dominant loop nucleus type. The present model predicts that the dislocation loop concentration decrease with increasing under sized atoms content, but defect production rate and temperature dependences of loop concentration are insensitive to alloy compositions. (author)

  12. 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

  13. Irradiation-induced segregation in multi-component alloys

    International Nuclear Information System (INIS)

    Chen, I.W.

    1983-01-01

    A unified analysis of irradiation-induced segregation in multi-component alloys is developed using the formulation of irreversible thermodynamics. Three distinct mechanisms for segregation, namely the inverse Kirkendall effect, the vacancy-wind effect, and the solute drag of interstitials, are identified. In particular, the inverse Kirkendall effect due to interstitials arises only if a solute-interstitial interaction or a mutual conversion among interstitials via lattice atom intermediaries operates simultaneously. In the limit of fast conversion a para-equilibrium state may be reached between interstitials and lattice atoms, and the interstitial mechanism becomes formally analogous to the vacancy mechanism. Although the past treatment of rate phenomena in this field was apparently limited to the latter case, the importance of the consideration of separate chemical potentials for interstitials of different species, in segregation and other irradiation effects, is emphasized. (orig.)

  14. Precipitate resolution in an electron irradiated ni-si alloy

    Science.gov (United States)

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

    1988-09-01

    Precipitate resolution processes in a Ni-12.6 at% Si alloy under electron irradiation have been observed by means of HVEM. Above 400°C, growth and resolution of Ni 3Si precipitates were observed simultaneously. The detail stereoscopic observation showed that the precipitates close to free surfaces grew, while those in the middle of a specimen dissolved. The critical dose when the precipitates start to shrink increases with increasing the depth. This depth dependence of the precipitate behavior under irradiation has a close relation with the formation of surface precipitates and the growth of solute depleted zone beneath them. The temperature and dose dependence of the resolution rate showed that the precipitates in the solute depleted zone dissolved by the interface controlled process of radiation-enhanced diffusion.

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

    International Nuclear Information System (INIS)

    Nogaret, Th.

    2007-12-01

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

  16. Simulation of radiation induced segregation and PWSCC susceptibility for austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto Koji; Yonezawa, Toshio; Iwamura, Toshihiko [Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago R and D Center; Ajiki, Kazuhide [Mitsubishi Heavy Industries Ltd., Kobe (Japan). Kobe Shipyard and Machinery Works; Urata, Sigeru [General Office of Nuclear and Fossil Power Production, Kansai Electric Power Co., Inc., Osaka (Japan)

    2000-08-01

    Recently, irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internal components materials become a subject of discussion in light water reactors (LWRs). IASCC has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC of austenitic stainless steels for core internal materials so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, in order to verify the hypothetical that the IASCC in PWRs shall be caused by the primary water stress corrosion cracking (PWSCC) as a result of radiation induced segregation (RIS) at grain boundaries, the authors simulated RIS at grain boundaries of austenitic stainless steels based on previous study and estimated RIS tendency after long time operation. And the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated austenitic stainless steels and made clear chromium-nickel-silicon compositions for PWSCC susceptibility area in austenitic alloys by slow strain rate tensile (SSRT) test. (author)

  17. Simulation of radiation induced segregation and PWSCC susceptibility for austenitic stainless steels

    International Nuclear Information System (INIS)

    Fujimoto Koji; Yonezawa, Toshio; Iwamura, Toshihiko

    2000-01-01

    Recently, irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internal components materials become a subject of discussion in light water reactors (LWRs). IASCC has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC of austenitic stainless steels for core internal materials so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, in order to verify the hypothetical that the IASCC in PWRs shall be caused by the primary water stress corrosion cracking (PWSCC) as a result of radiation induced segregation (RIS) at grain boundaries, the authors simulated RIS at grain boundaries of austenitic stainless steels based on previous study and estimated RIS tendency after long time operation. And the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated austenitic stainless steels and made clear chromium-nickel-silicon compositions for PWSCC susceptibility area in austenitic alloys by slow strain rate tensile (SSRT) test. (author)

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

    International Nuclear Information System (INIS)

    Mills, W.J.; James, L.A.

    The effect of fast-neutron irradiation on the fatigue-crack propagation and fracture toughness behaviour of Alloy A-286 was characterized using fracture mechanics techniques. The fracture toughness was found to decrease continuously with increasing irradiation damage at both 24 deg. C and 427 deg. C. In the unirradiated and low fluence conditions, specimens displayed appreciable plasticity prior to fracture, and equivalent Ksub(Ic) values were determined from Jsub(Ic) fracture toughness results. At high irradiation exposure levels, specimens exhibited a brittle Ksub(Ic) fracture mode. The 427 deg. C fracture toughness fell from 129 MPa√m in the unirradiated condition to 35 MPa√m at an exposure of 16.2 dpa (total fluence of 5.2x10 22 n/cm 2 ). Room temperature fracture toughness values were consistently 40 to 60 percent higher than the 427 deg. C values. Electron fractography revealed that the reduction in fracture resistance was attributed to a fracture mechanism transition from ductile microvoid coalescence to channel fracture. Fatigue-crack propagation tests were conducted at 427 deg. C on specimens irradiated at 2.4 dpa and 16.2 dpa. Crack growth rates at the lower exposure level were comparable to those in unirradiated material, while those at the higher exposure were slightly higher than in unirradiated material. (author)

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

    International Nuclear Information System (INIS)

    Stumphy, B.; Chee, S.W.; Vo, N.Q.; Averback, R.S.; Bellon, P.; Ghafari, M.

    2014-01-01

    Compositional patterning in dilute Cu 1−x Fe x (x ≈ 12%) induced by 1.8 MeV Kr + irradiation was studied as a function of temperature using atom probe tomography. Irradiation near room temperature led to homogenization of the sample, whereas irradiation at 300 °C and above led to precipitation and macroscopic coarsening. Between these two temperatures the irradiated alloys formed steady state patterns of composition where precipitates grew to a fixed size. The size in this regime increased somewhat with temperature. It was also observed that the steady state concentrations of Fe in Cu matrix and Cu in the Fe precipitates both greatly exceeded their equilibrium solubilities, with the degree of supersaturation in each phase decreasing with increasing temperature. In the macroscopic coarsening regime, the Fe-rich precipitates showed indications of a “cherry-pit” structure, with Cu precipitates forming within the Fe precipitates. In the patterning regime, interfaces between Fe-rich precipitates and the Cu-rich matrix were irregular and diffuse

  20. 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

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

    DEFF Research Database (Denmark)

    Barlow, P.; Leffers, Torben

    1977-01-01

    Climb sources emitting dislocation loops are observed in Cu-Ni alloys during irradiation with 1 MeV electrons in a high voltage electron microscope. High source densities are found in alloys containing 5, 10 and 20% Ni, but sources are also observed in alloys containing 1 and 2% Ni. The range of ...

  2. 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.)

  3. 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

  4. Alloys of nickel-iron and nickel-silicon do not swell under fast neutron irradiation

    International Nuclear Information System (INIS)

    Silvestre, G.; Silvent, A.; Regnard, C.; Sainfort, G.

    1975-01-01

    This research is concerned with the effect of fast-neutron irradiation on the swelling of nickel and nickel alloys. Ni-Fe (0-60at%Fe) and Ni-Si (0-8at%Si) were studied, and the fluences were in the range 10 20 -4.3x10 22 n/cm 2 . In dilute alloys, the added elements are dissolved and reduce swelling, silicon being particularly effective. In more concentrated alloys, irradiation of Ni-Fe and Ni-Si alloys brings about the formation of plate-shaped precipitates of Ni 3 X and these alloys do not swell. (Auth.)

  5. Microstructure and mechanical properties of 2024-T3 and 7075-T6 aluminum alloys and austenitic stainless steel 304 after being exposed to hydrogen peroxide

    Science.gov (United States)

    Sofyan, Nofrijon Bin Imam

    The effect of hydrogen peroxide used as a decontaminant agent on selected aircraft metallic materials has been investigated. The work is divided into three sections; bacterial attachment behavior onto an austenitic stainless steel 304 surface; effect of decontamination process on the microstructure and mechanical properties of aircraft metallic structural materials of two aluminum alloys, i.e. 2024-T3 and 7075-T6, and an austenitic stainless steel 304 as used in galley and lavatory surfaces; and copper dissolution rate into hydrogen peroxide. With respect to bacterial attachment, the results show that surface roughness plays a role in the attachment of bacteria onto metallic surfaces at certain extent. However, when the contact angle of the liquid on a surface increased to a certain degree, detachment of bacteria on that surface became more difficult. In its relation to the decontamination process, the results show that a corrosion site, especially on the austenitic stainless steel 304 weld and its surrounding HAZ area, needs more attention because it could become a source or a harborage of bio-contaminant agent after either incidental or intentional bio-contaminant delivery. On the effect of the decontamination process on the microstructure and mechanical properties of aircraft metallic structural materials, the results show that microstructural effects are both relatively small in magnitude and confined to a region immediately adjacent to the exposed surface. No systematic effect is found on the tensile properties of the three alloys under the conditions examined. The results of this investigation are promising with respect to the application of vapor phase hydrogen peroxide as a decontaminant agent to civilian aircraft, in that even under the most severe circumstances that could occur; only very limited damage was observed. The results from the dissolution of copper by concentrated liquid hydrogen peroxide showed that the rate of copper dissolution increased for

  6. Mechanisms of irradiation growth of alpha-zirconium alloys

    International Nuclear Information System (INIS)

    Holt, R.A.

    1988-01-01

    Experimental observations in the last few years have shown that the range of irradiation growth behaviour of alpha-zirconium alloys is more varied, that a wider variety of sinks must be considered, and that there are more potential sources of anisotropy than was previously recognized. The important new experimental observations which influence our preception of the growth phenomenon in zirconium alloys include the growth of single crystals, accelerating growth in annealed material with the coincident appearance of vacancy loops on the basal planes, the occurrence of 'negative' growth, i.e., contractions along prism directions, the absence of a pronounced effect of grain size on the long term growth rate at low temperatures, and the presence of intergranular constraints prior to irradiation. With the greater complexity of behaviour now being observed, it is necessary to apply new theoretical concepts to assist in understanding growth, e.g., the potential role of anisotropic diffusion in segregation point defects to different sinks and 'growth' caused by the anisotropic relaxation of intergranular constrains. These can be combined with earlier ideas to predict a variety of growth behaviours, including 'negative growth'. Because the most important physical information required for theoretical treatments of growth, i.e, the characteristics of vacancies and self interstitial atoms, are still poorly understood, it is almost impossible to test rigorously any particular theoretical concept and a complete picture of growth has yet to emerge. (orig./MM)

  7. Study of retained austenite and nano-scale precipitation and their effects on properties of a low alloyed multi-phase steel by the two-step intercritical treatment

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Z.J.; Han, G., E-mail: hangang@mater.ustb.edu.cn; Zhou, W.H.; Zeng, C.Y.; Shang, C.J., E-mail: cjshang@ustb.edu.cn

    2016-03-15

    Microstructure evolution and properties were studied in a low carbon low alloyed hot-rolled bainitic steel by annealing and annealing plus tempering. Microstructure of the hot-rolled steel consists of lath bainite and martensite. By annealing at 720 °C for 30 min and water quenching, multi-phase microstructure consisting of intercritical ferrite, tempered bainite/martensite, retained austenite and fresh martensite was obtained. With increasing annealing temperature to 760 °C, microstructure of the steel consisted of intercritical ferrite, fresh martensite without retained austenite. After the second step of tempering at 680 °C for samples annealed both at 720 °C and 760 °C, ~ 8–9% volume fraction of retained austenite was obtained in the multi-phase microstructure. Moreover, fine precipitates of VC with size smaller than 10 nm and copper precipitates with size of ~ 10–50 nm were obtained after tempering. Results from scanning transmission electron microscopy (STEM) give evidence to support that the partitioning of Mn, Ni and Cu is of significance for retained austenite stabilization. Due to the combined contribution of multiphase microstructure, the transformation-induced-plasticity effect of retained austenite and strengthening effect of nanometer-sized precipitates, yield strength greater than 800 MPa, yield to tensile ratio of 0.9, uniform elongation of ~ 9% and good low temperature impact toughness of 147 J at − 40 °C were achieved. - Highlights: • Stable retained austenite was produced in a low alloyed steel. • Partition of Mn, Ni and Cu was confirmed by STEM for austenite stabilization. • Nano-sized VC and Cu precipitates were achieved by second tempering. • High strength–high toughness with low Y/T ratio was obtained.

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

    International Nuclear Information System (INIS)

    Liu, P.P.; Yu, R.; Zhu, Y.M.; Zhao, M.Z.; Bai, J.W.; Wan, F.R.; Zhan, Q.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  10. 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.

  11. Alloying effects on dissolution rate of crevice corrosion for austenitic stainless steels in 3% NaCl solution at 80 C

    International Nuclear Information System (INIS)

    Chen, P.; Shinohara, Tadashi; Tsujikawa, Shigeo

    1996-01-01

    Chloride stress corrosion cracking (SCC) has been a problem for austenitic stainless steel in aqueous environments containing chlorides. Studies have found that SCC initiates only from a dissolving surface and under the condition that the crack growth rate is higher than the dissolution rate of the dissolving surface. Research conducted to improve the resistance to SCC for Type 304 steels (UNS S30400) have revealed that while molybdenum and phosphorus are unfavored, the combined alloying of 3% aluminum with 2% copper can almost nullify their detrimental effect. Based on the mentioned criteria, this study was dedicated to clarify the mechanism behind these alloying effects by examining the relationship between the measured enhancements on SCC resistance and the dissolution rate observed via the moire technique. It was found that the addition of both molybdenum and phosphorus reduces the dissolution rate and therefore impaired SCC resistance; the addition of copper increases the dissolution rate of steady growth stage where crevice corrosion proceeds at a constant rate. Moreover this dissolution rate could further be increased when combined with the alloying of aluminum. These observed results correspond well to that of the measured behavior of the SCC critical temperature, T c , suggesting that the SCC susceptibility is influenced by anodic dissolution

  12. Microstructure and tribologic behaviour of metastable austenitic FeMn alloys as a function of chromium content; Gefuegeausbildung und Triboverhalten metastabiler austenitischer FeMn-Legierungen in Abhaengigkeit vom Chromgehalt

    Energy Technology Data Exchange (ETDEWEB)

    Roethig, J. [Magdeburg Univ. (Germany). Inst. fuer Stroemungstechnik und Thermodynamik; Veit, P.; Strassburger, G.; Blaesing, J. [Magdeburg Univ. (Germany). Inst. fuer Experimentelle Physik; Heyse, H. [Magdeburg Univ. (Germany). Inst. fuer Werkstofftechnik und Werkstoffpruefung

    1997-12-31

    In FeMn20Cr alloys with chromium contents of up to 20%, the solidification process is primarily an eutectic process. The {delta}-ferrite becomes increasingly instable below a temperature of 900 C and gradually disintegrates during slow cooling into austenite and a sigma phase. Tempering of these microstructures at T=450 C (6hours) leads to formation of {epsilon}-martensite in the austenite. Fast quenching starting above 900 C freezes the {delta}-ferrite, so that in the case of chromium contents between 13 and 18%, austenitic-hexagonal-ferritic microstructures form and above 18%, austenitic-ferritic microstructures. Tempering does not remove the {delta}-ferrite, but induces formation of {epsilon}-martensite in the austenite. Trobologic examinations with solutionized and water-quenched alloys showed, as compared to an FeMn20Cr18 alloy, for various types of wear, a very good tribologic performance (except for the alloy FeMn20Cr18 and cavitation). As to abrasion or hot wear, the formation of a sigma-phase or intercalation of metalloid hard phases should be considered. (orig./CB) [Deutsch] FeMn20Cr-Legierungen mit Chromgehalten bis zu 20% erstarren primaer ferritisch. Der {delta}-Ferrit ist unterhalb 900 C nicht mehr stabil und zerfaellt bei langsamer Abkuehlung in Austenit und Sigmaphase. Ein Anlassen dieser Gefuege T=450 C (6 Stunden) fuehrt zur {epsilon}-Martensitbildung im Austenit. Schnelles Abschrecken von oberhalb 900 C friert den {delta}-Ferrit ein, so dass bei Chromgehalten zwischen 13 und 18% austenitisch-hexagonal-ferritische und >18% austenitisch-ferritische Gefuege entstehen. Durch Anlassen kann der {delta}-Ferrit nicht beseitigt werden. Im Austenit kommt es aber zur {epsilon}-Martensitbildung. Tribologische Untersuchungen mit loesungsgegluehten und in Wasser abgeschreckten Legierungen zeigten im Vergleich zu einer FeCrNi-Legierung bei verschiedenen Verschleissarten (mit Ausnahme FeMn20Cr18 bei Kavitation) ein sehr gutes Triboverhalten. Gegenueber Abrasion

  13. 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)

  14. Optimization of HIP bonding conditions for ITER shielding blanket/first wall made from austenitic stainless steel and dispersion strengthened copper alloy

    International Nuclear Information System (INIS)

    Sato, S.; Hatano, T.; Kuroda, T.; Furuya, K.; Hara, S.; Enoeda, M.; Takatsu, H.

    1998-01-01

    Optimum bonding conditions were studied on the hot isostatic pressing (HIP) bonded joints of type 316L austenitic stainless steel and dispersion strengthened copper alloy (DSCu) for application to the ITER shielding blanket / first wall. HIP bonded joints were fabricated at temperatures in a 980-1050 C range, and a series of mechanical tests and metallurgical observations were conducted on the joints. Also, bondability of two grades of DSCu (Glidcop Al-25 trademark and Al-15 trademark ) with SS316L was examined in terms of mechanical properties of the HIP bonded joints. From those studies it was concluded that the HIP temperature of 1050 C was an optimal condition for obtaining higher ductility, impact values and fatigue strength. Also, SS316L/Al-15 joints showed better results in terms of ductility and impact values compared with SS316L/Al-25 joints. (orig.)

  15. Reactor irradiation and helium-3 effects on mechanical properties of alpha-titanium alloys

    International Nuclear Information System (INIS)

    Tebus, V.N.; Alekseev, Eh.F.; Golikov, I.V.

    1990-01-01

    Dependence of α-titanium alloy mechanical properties on test temperature and neutron fluence is investigated. Irradiation is shown to result in material hardening and in their plasticity reduction, but residual plasticity remains rather high. Additional reduction of plasticity results in helium-3 introduced in materials under irradiation. Restoration of properties is observed at test temperature higher 500 deg C. Irradiation by fast neutrons up to high fluences (1.4·10 23 cm -2 ) results in essential alloy softening

  16. Structure and properties of UV-irradiated LLDPE and alloy of PA66 and the irradiated LLDPE

    International Nuclear Information System (INIS)

    Ran Qianping; Zou Hua; Wu Shishan; Shen Jian

    2006-01-01

    Some oxygen-containing groups such as C=O, C-O and -OH were introduced onto linear low-density polyethylene (LLDPE) chains by UV irradiation in air. Their concentration increased with the irradiation time. Crystal shape of the irradiated LLDPE remained an orthorhombic structure, while space of the crystalline plane kept unchanged. The melting temperature and crystallinity decreased due to the LLDPE chain scission into small molecules compound and crystalline defects caused by UV irradiation. Compared with pristine LLDPE, hydrophilicity of the irradiated LLDPE increased due to the introduction of polar oxygen-containing groups, but the tensile strength decreased due to the LLDPE chain degradation and reduction of crystallinity. The temperature of initial weight loss of the irradiated LLDPE was lower than that of pristine LLDPE. An alloy of PA66 and the irradiated LLDPE (irradiated PA66/LLDPE) was prepared by melting blend at 260-270 degree C. Compared to non-irradiated PA66/LLDPE alloy, dispersion of LLDPE particles in the irradiated PA66/LLDPE alloy and interfacial interactions between the components were markedly improved. Therefore, tensile strength and impact strength of the irradiated PA66/ LLDPE were higher than those of the control. (authors)

  17. Unirradiated UO2 in irradiated zirconium alloy sheathing

    International Nuclear Information System (INIS)

    MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

    1979-07-01

    Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

  18. 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

  19. Effects of irradiation on ferritic alloys and implications for fusion reactor applications

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1986-07-01

    This paper reviews the ADIP irradiation effects data base on ferritic (martensitic) alloys to provide reactor teams with an understanding of how such alloys will behave for fusion reactor first wall applications. Irradiation affects dimensional stability, strength and toughness. Dimensional stability is altered by precipitation and void swelling. Swelling as high as 25% may occur in some ferritic alloys at 500 dpa. Irradiation alters strength both during and following irradiation. Irradiation at low temperatures leads to hardening whereas at higher temperatures and high exposures, precipitate coarsening can result in softening. Toughness can also be adversely affected by irradiation. Failure can occur in ferritic in a brittle manner and irradiation induced hardening causes brittle failure at higher temperatures. Even at high test temperatures, toughness is reduced due to reduced failure initiation stresses. 39 refs

  20. 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

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

    International Nuclear Information System (INIS)

    Ames, M.; Kohse, G.; Lee, T.S.; Grant, N.J.; Harling, O.K.

    1986-01-01

    A scoping experiment in which 25 different copper materials of 17 alloy compositions were irradiated to approx.13.5 dpa approx.400 0 C in a fast reactor is described. The materials include rapidly solidified (RS) alloys, with and without oxide dispersion strengthening, as well as conventionally processed alloys. Immersion density (swelling), electrical conductivity (which can be related to thermal conductivity), and yield stress and ductility by miniature disk bend testing have been measured before and after irradiation. It was found, in general, that the Rs alloys are stable under irradiation to 13.5 dpa, showing small conductivity changes and little or no swelling. Reduction of strength and ductility, in post-irradiation tests at the irradiation temperature, are not generally observed. Some conventionally processed alloys also performed well, although irradiation softening and swelling of several percent were observed in some cases, and pure copper swelled in excess of 5%. It is concluded that a number of copper alloys should receive further study, and that higher dose irradiations will be required to establish the limits of swelling suppression in these alloys

  2. Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Miya, Naoyuki

    2002-01-01

    In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

  3. Effect of irradiation on the tensile properties of niobium-base alloys

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Heestand, R.L.; Atkin, S.D.

    1986-11-01

    The alloys Nb-1Zr and PWC-11 (Nb-1Zr-0.1C) were selected as prime candidate alloys for the SP-100 reactor. Since the mechanical properties of niobium alloys irradiated to end-of-life exposure levels of about 2 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures above 1300 K were not available, an irradiation experiment (B-350) in EBR-II was conducted. Irradiation creep, impact properties, bending fatigue, and tensile properties were investigated; however, only tensile properties will be reported in this paper. The tensile properties were studied since they easily reveal the common irradiation phenomena of hardening and embrittlement. Most attention was directed to testing at the irradiation temperature. Further testing was conducted at lower temperatures in order to scope the behavior of the alloys in cooldown conditions

  4. Effects of neutron irradiation on microstructure in experimental and commercial ferritic alloys

    International Nuclear Information System (INIS)

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

    1983-05-01

    A series of microstructural studies have been undertaken on fast-reactor-irradiated specimens of experimental ferritic alloys and ferritic/martensitic commercial alloys covering a broad range of compositions and starting microstructures. It is found that voids do indeed form in ferritic alloys and that dislocation loops and tangles are created during irradiation at temperatures below 500 0 C. Swelling rates as high as 0.25% per 10 22 n/cm 2 have been measured. However, the major effect of irradiation is precipitation and precipitation can suppress void swelling completely and/or be responsible for degradation of mechanical properties

  5. Influence of grain structure on the deformation mechanism in martensitic shear reversion-induced Fe-16Cr-10Ni model austenitic alloy with low interstitial content: Coarse-grained versus nano-grained/ultrafine-grained structure

    Energy Technology Data Exchange (ETDEWEB)

    Challa, V.S.A. [Laboratory for Excellence in Advanced Steel Research, Department of Metallurgical, Materials Engineering, and Biomedical Engineering, University of Texas at El Paso, El Paso, TX 79968 (United States); Misra, R.D.K., E-mail: dmisra2@utep.edu [Laboratory for Excellence in Advanced Steel Research, Department of Metallurgical, Materials Engineering, and Biomedical Engineering, University of Texas at El Paso, El Paso, TX 79968 (United States); Somani, M.C. [Center for Advanced Steels Research, The University of Oulu, P.O. Box 4200, 90014 Oulu (Finland); Wang, Z.D. [State Key Laboratory for Rolling and Automation, Northeastern University, 3-11 Wenhua Road, Shenyang 110819 (China)

    2016-04-20

    Nanograined/ultrafine-grained (NG/UFG) materials characterized by high strength-high ductility combination are excellent vehicles to obtain an unambiguous understanding of deformation mechanisms vis-à-vis their coarse-grained counterparts. In this context, the innovative concept of phase reversion-induced NG/UFG structure enabled achieving high strength besides comparable ductility, for instance, in metastable austenitic stainless steels. In the phase reversion process, severe deformation of austenite at room temperature (typically ~60–80%) transforms face-centered cubic austenite (γ) to body centered cubic martensite (α′). Upon annealing, martensite reverts to austenite leading to extensive grain refinement. The objective of the present study to fundamentally understand the deformation mechanisms in NG/UFG structure in relation to that of the coarse-grained (CG) structure was accomplished by combining depth-sensing nanoscale experiments on an Fe-16Cr-10Ni model austenitic alloy conducted at different strain rates, followed by the study of structural evolution in the deformed zone using transmission electron microscopy (TEM). In the high strength NG/UFG steel (YS~585 MPa), stacking faults and nanotwins contributed to the enhanced ductility (El~35%), while in the case of low strength (YS~260 MPa) coarse-grained (CG) counterpart, ductility was also high (El~40%), but chiefly due to strain-induced martensite, which points to a clear case of grain size effect (and the corresponding level of strength). The distinct change in the deformation mechanism from stacking faults and twinning-induced plasticity (TWIP) in the NG structure to transformation-induced plasticity (TRIP) in the CG structure is elucidated in terms of austenite stability-strain energy relationship. The insights on the relationship between grain structure (and strength) and deformation mechanisms are envisaged to be important in providing a new direction for the futuristic design of high strength

  6. Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

    Science.gov (United States)

    Yamakawa, K.; Shimomura, Y.

    1999-01-01

    The defects formed in Ni based (Ni-Si, Ni-Cu and Ni-Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni-Si alloys was much higher than those in Ni-Cu and Ni-Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni-Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni-Cu alloys but scarcely formed in Ni-Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni-Cu alloys. Possible formation processes of interstitial loops, SFT dislocation lines and voids are discussed.

  7. Damage structures in fission-neutron irradiated Ni-based alloys at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Yamakawa, K.; Shimomura, Y. [Hiroshima Univ. (Japan). Faculty of Engineering

    1999-01-01

    The defects formed in Ni based (Ni-Si, Ni-Cu and Ni-Fe) alloys which were irradiated with fission-neutrons were examined by electron microscopy. Irradiations were carried out at 473 K and 573 K. In the 473 K irradiated specimens, a high density of large interstitial loops and small vacancy clusters with stacking fault tetrahedra (SFT) were observed. The number densities of these two types of defects did not strongly depend on the amount of solute atoms in each alloy. The density of the loops in Ni-Si alloys was much higher than those in Ni-Cu and Ni-Fe alloys, while the density of SFT only slightly depended on the kind of solute. Also, the size of the loops depended on the kinds and amounts of solute. In 573 K irradiated Ni-Cu specimens, a high density of dislocation lines developed during the growth of interstitial loops. In Ni-Si alloys, the number density and size of the interstitial loops changed as a function of the amount of solute. Voids were formed in Ni-Cu alloys but scarcely formed in Ni-Si alloys. The number density of voids was one hundredth of that of SFT observed in 473 K irradiated Ni-Cu alloys. Possible formation processes of interstitial loops, SFT, dislocation lines and voids are discussed. (orig.) 8 refs.

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

    Science.gov (United States)

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

    2010-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2010-12-01

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

  10. The Use of Austenitic Stainless Steel versus Monel (Ni-Cu) Alloy in Pressurized Gaseous Oxygen (GOX) Life Support Systems.

    Science.gov (United States)

    1985-03-01

    Carbon Steel AISI 1025 2. AISI 4140 3. Ductile Iron 4. 304 Stainless Steel 5. 17-4 PH Stainless Steel 6. 410 Stainless Steel 7. Lead Babbit 8. Tin Babbit...9. Inconel 718 i0. Aluminum 1100 30 6- AISI 4140 steel, all the results were negative (no ignitions). The single exception was with a sample of 4140 ...rates for austenitic stainless steel ( AISI 316), Monel (63% Ni - 34% Cu) and carbon steel (AMS 5050) tubing in this environment. 12 - 14-660 A 7

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

    Science.gov (United States)

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

    2012-11-27

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

  12. 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

  13. Interdiffusion between U-Mo alloys and Al or Al alloys at 340 deg. C. Irradiation plan

    International Nuclear Information System (INIS)

    Fortis, A.M.; Mirandou, M.; Ortiz, M.; Balart, S.; Denis, A.; Moglioni, A.; Cabot, P.

    2005-01-01

    Out of reactor interdiffusion experiments between U-Mo alloys and Al alloys made close to fuel operation temperature are needed to validate the results obtained above 500 deg. C. A study of interdiffusion between U-Mo and Al or Al alloys, out and in reactor, has been initiated. The objective is to characterize the interdiffusion layer around 250 deg. C and study the influence of neutron irradiation. Irradiation experiments will be performed in the Argentine RA3 reactor and chemical diffusion couples will be fabricated by Friction Stir Welding (FSW) technique. In this work out-of-pile diffusion experiments performed at 340 deg. C are presented. Friction Stir Welding (FSW) was used to fabricate some of the samples. One of the results is the presence of Si, in the interaction layer, coming from the Al alloy. This is promising in the sense that the absence of Al rich phases may also be expected at low temperature. (author)

  14. Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, Samuel A., E-mail: sabriggs2@wisc.edu [University of Wisconsin-Madison, 1415 Engineering Drive, Madison, WI 53706 (United States); Barr, Christopher M. [Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Pakarinen, Janne [University of Wisconsin-Madison, 1415 Engineering Drive, Madison, WI 53706 (United States); SKC-CEN Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Mamivand, Mahmood [University of Wisconsin-Madison, 1415 Engineering Drive, Madison, WI 53706 (United States); Hattar, Khalid [Sandia National Laboratories, PO Box 5800, Albuquerque, NM 87185 (United States); Morgan, Dane D. [University of Wisconsin-Madison, 1415 Engineering Drive, Madison, WI 53706 (United States); Taheri, Mitra [Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Sridharan, Kumar [University of Wisconsin-Madison, 1415 Engineering Drive, Madison, WI 53706 (United States)

    2016-10-15

    Two binary Ni-Cr model alloys with 5 wt% Cr and 18 wt% Cr were irradiated using 2 MeV protons at 400 and 500 °C and 20 MeV Ni{sup 4+} ions at 500 °C to investigate microstructural evolution as a function of composition, irradiation temperature, and irradiating ion species. Transmission electron microscopy (TEM) was applied to study irradiation-induced void and faulted Frank loops microstructures. Irradiations at 500 °C were shown to generate decreased densities of larger defects, likely due to increased barriers to defect nucleation as compared to 400 °C irradiations. Heavy ion irradiation resulted in a larger density of smaller voids when compared to proton irradiations, indicating in-cascade clustering of point defects. Cluster dynamics simulations were in good agreement with the experimental findings, suggesting that increases in Cr content lead to an increase in interstitial binding energy, leading to higher densities of smaller dislocation loops in the Ni-18Cr alloy as compared to the Ni-5Cr alloy. - Highlights: • Binary Ni-Cr alloys were irradiated with protons or Ni ions at 400 and 500 °C. • Higher irradiation temperatures yield increased size, decreased density of defects. • Hypothesize that varying Cr content affects interstitial binding energy. • Fitting CD models for loop nucleation to data supports this hypothesis.

  15. 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

  16. An overview of microstructural and experimental factors that affect the irradiation growth behavior of zirconium alloys

    International Nuclear Information System (INIS)

    Fidleris, V.; Tucker, R.P.; Adamson, R.B.

    1987-01-01

    This paper presents an overview of factors affecting irradiation growth of zirconium alloys. Recent data obtained from irradiation programs in EBR-II, ATR, and NRU reactors are used to illustrate the effects of various microstructural and experimental factors on the growth of Zircaloy, zirconium, and zirconium-biobium alloys irradiated to fluences up to 2 X 10 26 nm -2 (E > 1 MeV) over the temperature range 330 to 720 K. Open literature results are also used to confirm or illustrate various effects. Important factors are texture, grain boundary parameters, residual stresses, original dislocation density, microstructure evolution, temperature during irradiation, solute effects, and fluence

  17. 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.

  18. 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

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

    International Nuclear Information System (INIS)

    Decours, Jacques; Cadalbert, Robert; Sidhom, Habib.

    1982-06-01

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

  20. 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

  1. Irradiation performance of oxide dispersion strengthened copper alloys to 150 dpa at 415 degree C

    International Nuclear Information System (INIS)

    Edwards, D.J.; Kumar, A.S.; Anderson, K.R.; Stubbins, J.F.; Garner, F.A.; Hamilton, M.L.

    1991-11-01

    Results have been obtained on the post-irradiation properties of various oxide dispersion strengthened copper alloys irradiated from 34 to 150 dpa at 415 degrees C in the Fast Flux Test Facility. The GlidCop alloys strengthened by Al 2 O 3 continue to outperform other alloys with respect to swelling resistance, and retention of both electrical conductivity and yield strength. Several castable ODS alloys and a Cr 2 O 3 -strengthened alloy show increasingly poor resistance to radiation, especially in their swelling behavior. A HfO 2 -strengthened alloy retains most of its strength and its electrical conductivity reaches a constant level after 50 dpa, but it exhibits a higher residual radioactivity

  2. 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.

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

    International Nuclear Information System (INIS)

    Cox, B.; Ishigure, K.; Johnson, A.B.; Lemalgnan, J.C.; Nechaev, A.F.; Petrik, N.G.; Reznichenko, E.A.

    1990-01-01

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

  4. Amorphous-to-crystalline phase transformation by neutron irradiation of the alloy Fe83B17

    International Nuclear Information System (INIS)

    Weis, J.; Gabris, F.; Cerven, I.; Sitek, J.

    1984-01-01

    The purpose of the present work is to investigate the structural changes of amorphous Fe 83 B 17 alloy after irradiation with fast neutrons ( > 1 MeV) and to compare with the crystallization behaviour of the amorphous Fe 83 B 17 alloy after annealing. The structural changes were studied by Moessbauer spectroscopy and X-ray diffraction with the usual Fourier analysis. (author)

  5. 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

  6. Evaluation of critical resolved shear strength and deformation mode in proton-irradiated austenitic stainless steel using micro-compression tests

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung-Ha; Ko, Eunsol; Kwon, Junhyun; Hwang, Seong Sik [Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Shin, Chansun, E-mail: c.shin@mju.ac.kr [Department of Materials Science and Engineering, Myongji University, 116 Myongji-ro, Cheoin-gu, Youngin, Gyeonggi-do, 449-728 (Korea, Republic of)

    2016-03-15

    Micro-compression tests were applied to evaluate the changes in the strength and deformation mode of proton-irradiated commercial austenitic stainless steel. Proton irradiation generated small dots at low dose levels and Frank loops at high dose levels. The increase in critical resolved shear stresses (CRSS) was measured from micro-compression of pillars and the Schmid factor calculated from the measured loading direction. The magnitudes of the CRSS increase were in good agreement with the values calculated from the barrier hardening model using the measured size and density of radiation defects. The deformation mode changed upon increasing the irradiation dose level. At a low radiation dose level, work hardening and smooth flow behavior were observed. Increasing the dose level resulted in the flow behavior changing to a distinct heterogeneous flow, yielding a few large strain bursts in the stress–strain curves. The change in the deformation mode was related to the formation and propagation of defect-free slip bands. The effect of the orientation of the pillar or loading direction on the strengths is discussed.

  7. Effect of neutron irradiation on the cellular stage of Ni-Be alloy decomposition

    International Nuclear Information System (INIS)

    Larikov, L.N.; Borimskaya, S.T.

    1981-01-01

    Effects of neutron irradiation on the cellular stage of decomposition are investigated in deformed supersaturated solid solution Ni-1.92%Be by the X-ray structural and metallographic analyses. Radiation-initiated stimulation of the recovery properties in the deformed alloy and a lower rate of the cellular decomposition in irradiated samples are discovered [ru

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

    Alloy 718 is a nickel-base superalloy whose strength derives from γ′(Ni 3 (Al,Ti)) and γ″(Ni 3 Nb) precipitates. The evolution of the precipitates in alloy 718 irradiated with argon ions at elevated temperature were examined via transmission electron microscopy. Selected-area electron diffraction indicated superlattice spots disappeared after argon ion irradiation, which showing that the ordered structure of the γ′ and γ″ precipitates became disordered. The size of the precipitates became smaller with the irradiation dose increasing at 290 °C

  11. Effects of post-irradiation annealing on the transformation behavior of Ti-Ni alloys

    International Nuclear Information System (INIS)

    Kimura, A.; Tsuruga, H.; Morimura, T.; Misawa, T.; Miyazaki, S.

    1993-01-01

    Recovery processes of martensitic transformation of neutron irradiated Ti-50.0, 50.5 and 51.0 at.%Ni alloys during post-irradiation annealing were investigated by means of differential scanning calorimetry (DSC), tensile tests and transmission electron microscope (TEM) observations. Neutron irradiation up to a fluence of 1.2x10 24 n/cm 2 at 333 K suppressed the martensitic transformation as well as the stress-induced martensitic transformation of these alloys above 150 K. The TEM observations revealed that the disordered zones containing small defect clusters in high density were formed in the neutron irradiated Ti-Ni alloys. The DSC measurements also showed that the post-irradiation annealing caused recovery of the transformation of which the progress depended on the annealing temperature and period. A significant retardation of the recovery was recognized in the Ti-51.0 at.%Ni alloy in comparison with the Ti-50.0 at.%Ni alloy. From the shifts in the transformation temperature upon isothermal annealing at various annealing temperatures, the activation energies of the recovery process of the transformation in the neutron irradiated Ti-50.0 and 51.0 at.%Ni alloys were evaluated by a cross-cut method to be 1.2 eV and 1.5 eV, respectively. The recovery of the transformation was ascribed to the re-ordering resulting from decomposition of vacancy clusters, and those obtained values of the activation energy were considered to be the sum of the migration energy of vacancy and the binding energy of vacancy-vacancy cluster. The retardation of the recovery in the Ti-51.0 at%Ni alloy was interpreted in terms of large binding energy in this alloy due to the off-stoichiometry. (author)

  12. Fine structure characterization of martensite/austenite constituent in low-carbon low-alloy steel by transmission electron forward scatter diffraction.

    Science.gov (United States)

    Li, C W; Han, L Z; Luo, X M; Liu, Q D; Gu, J F

    2016-11-01

    Transmission electron forward scatter diffraction and other characterization techniques were used to investigate the fine structure and the variant relationship of the martensite/austenite (M/A) constituent of the granular bainite in low-carbon low-alloy steel. The results demonstrated that the M/A constituents were distributed in clusters throughout the bainitic ferrite. Lath martensite was the main component of the M/A constituent, where the relationship between the martensite variants was consistent with the Nishiyama-Wassermann orientation relationship and only three variants were found in the M/A constituent, suggesting that the variants had formed in the M/A constituent according to a specific mechanism. Furthermore, the Σ3 boundaries in the M/A constituent were much longer than their counterparts in the bainitic ferrite region. The results indicate that transmission electron forward scatter diffraction is an effective method of crystallographic analysis for nanolaths in M/A constituents. © 2016 The Authors Journal of Microscopy © 2016 Royal Microscopical Society.

  13. 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

  14. Hydrogen release from irradiated vanadium alloy V-4Cr-4Ti

    Energy Technology Data Exchange (ETDEWEB)

    Klepikov, A.Kh. E-mail: klepikov@ietp.alma-ata.su; Romanenko, O.G.; Chikhray, Y.V.; Tazhibaeva, I.L.; Shestakov, V.P.; Longhurst, G.R. E-mail: gxl@inel.gov

    2000-11-01

    The present work is an attempt to obtain data concerning the influence of neutron and {gamma} irradiation upon hydrogen retention in V-4Cr-4Ti vanadium alloy. The experiments on in-pile loading of vanadium alloy specimens at the neutron flux density 10{sup 14} n/cm{sup 2} s, hydrogen pressure of 80 Pa, and temperatures of 563, 613 and 773 K were carried out using the IVG.1M reactor of the Kazakhstan National Nuclear Center. A preliminary set of loading/degassing experiments with non-irradiated material has been carried out to obtain data on hydrogen interaction with vanadium alloy. The, data presented in this work are related both to non-irradiated and irradiated samples.

  15. Hydrogen release from irradiated vanadium alloy V-4Cr-4Ti

    International Nuclear Information System (INIS)

    Klepikov, A.Kh.; Romanenko, O.G.; Chikhray, Y.V.; Tazhibaeva, I.L.; Shestakov, V.P.; Longhurst, G.R.

    2000-01-01

    The present work is an attempt to obtain data concerning the influence of neutron and γ irradiation upon hydrogen retention in V-4Cr-4Ti vanadium alloy. The experiments on in-pile loading of vanadium alloy specimens at the neutron flux density 10 14 n/cm 2 s, hydrogen pressure of 80 Pa, and temperatures of 563, 613 and 773 K were carried out using the IVG.1M reactor of the Kazakhstan National Nuclear Center. A preliminary set of loading/degassing experiments with non-irradiated material has been carried out to obtain data on hydrogen interaction with vanadium alloy. The, data presented in this work are related both to non-irradiated and irradiated samples

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

    International Nuclear Information System (INIS)

    Ardell, A.J.

    1982-04-01

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

  17. Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

    International Nuclear Information System (INIS)

    Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

    2008-01-01

    As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

  18. Carbon transfer between 2 1/4 Cr 1 Mo alloy and austenitic steels (experiments in anisothermal loops)

    International Nuclear Information System (INIS)

    Baque, P.; Besson, M.; Champeix, L.; Donati, J.R.; Oberlin, C.; Saint-Paul, P.

    1976-01-01

    Studies on carbon transfer between the ferritic steel 2 1/4 Cr 1 Mo and the austenitic steels 316L and 321H have shown that there is not any measurable carbon transfer in the operating conditions of the secondary circuit of PHENIX (475 deg C was the maximal temperature of the 2 1/4 Cr 1 Mo steel). A significant carbon transfer has been observed between the ferritic steel and the 316L steel when the 321H was replaced by the 2 1/4 Cr 1 Mo steel in the same thermohydraulic conditions (the ferritic steel was then used up to 545 deg C). This experiment has demonstrated the importance of the temperature and the initial carbon content of the ferritic steel as parameters in the decarburization process. It appears that decarburization may not be sensitive to the thermohydraulic conditions at least in the range investigated in those experiments. In the other hand the 316L steel is observed to have been carburized, the degree of carburization remaining appreciably constant and independent on the temperature between 400 deg C and 550 deg C [fr

  19. 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

  20. Ion irradiation-induced swelling and hardening effect of Hastelloy N alloy

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, S.J. [Key Laboratory of Artificial Micro-and Nano-structures of Ministry of Education, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Li, D.H.; Chen, H.C.; Lei, G.H.; Huang, H.F.; Zhang, W.; Wang, C.B. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Yan, L., E-mail: yanlong@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Fu, D.J. [Key Laboratory of Artificial Micro-and Nano-structures of Ministry of Education, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Tang, M. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2017-06-15

    The volumetric swelling and hardening effect of irradiated Hastelloy N alloy were investigated in this paper. 7 MeV and 1 MeV Xe ions irradiations were performed at room temperature (RT) with irradiation dose ranging from 0.5 to 27 dpa. The volumetric swelling increases with increasing irradiation dose, and reaches up to 3.2% at 27 dpa. And the irradiation induced lattice expansion is also observed. The irradiation induced hardening initiates at low ion dose (≤1dpa) then saturates with higher ion dose. The irradiation induced volumetric swelling may be ascribed to excess atomic volume of defects. The irradiation induced hardening may be explained by the pinning effect where the defects can act as obstacles for the free movement of dislocation lines. And the evolution of the defects' size and number density could be responsible for the saturation of hardness. - Highlights: •Irradiation Swelling: The irradiation induced volumetric swelling increases with ion dose. •Irradiation Hardening: The irradiation hardening initiates below 1 dpa, then saturates with higher ion dose (1–10 dpa). •Irradiation Mechanism: The irradiation phenomena are ascribed to the microstructural evolution of the irradiation defects.

  1. 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.

  2. Study of neutron irradiation on F82H alloys by Mössbauer spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Huang, S.S., E-mail: h.shaosong@ht8.ecs.kyoto-u.ac.jp; Kitao, S.; Kobayashi, Y.; Yoshiie, T.; Xu, Q.; Sato, K.; Seto, M.

    2015-01-15

    The effects of neutron irradiation on F82H ferritic/martensitic stainless steel and its model alloys were studied by Mössbauer spectroscopy. The microstructural damage mechanisms of these alloys, during the void incubation period were interpreted using the short range order (SRO) parameters. Results show that within Fe–8Cr alloy, the atoms in the nearest neighbor (NN) of the Fe nuclei were inhomogeneous, prior to irradiation. A configuration trapping model of Cr supported the negative average SRO observed for the NN shells in our Fe–Cr alloys. We found that irradiation also accelerated the SRO in Fe–8Cr through a diffusion mechanism, where Cr atom repulsion was concentration dependent. Finally, comparative studies were conducted on F82H model alloys using the present Mössbauer measurements and our previously reported work on positron annihilation spectroscopy, which further established that irradiation of the alloys promoted the growth of a M{sub 23}C{sub 6} complex.

  3. Compositional redistribution in alloy films under high-voltage electron microscope irradiation

    Science.gov (United States)

    Lam, Nghi Q.; Leaf, O. K.; Minkoff, M.

    1983-10-01

    The problem of nonequilibrium segregation in alloy films under high-voltage electron microscope (HVEM) irradiation at elevated temperatures is re-examined in the present work, taking into account the damage-rate gradients caused by radial variation in the electron flux. Axial and radial compositional redistributions in model solid solutions, representative of concentrated Ni-Cu, Ni-Al and Ni-Si alloys, were calculated as a function of time, temperature, and film thickness, using a kinetic theory of segregation in binary alloys. The numerical results were achieved by means of a new software package (DISPL2) for solving convection-diffusion-kinetics problems with general orthogonal geometries. It was found that HVEM irradiation-induced segregation in thin films consists of two stages. Initially, due to the proximity of the film surfaces as sinks for point defects, the usual axial segregation (to surfaces) occurs at relatively short irradiation times, and rapidly attains quasi-steady state. Then, radial segregation becomes more and more competitive, gradually affecting the kinetics of axial segregation. At a given temperature, the buildup time to steady state is much longer in the present situation than in the simple case of one-dimensional segregation with uniform defect production. Changes in the alloy composition occur in a much larger zone than the irradiated volume. As a result, the average alloy composition within the irradiated region can differ greatly from that of the unirradiated alloy. The present calculations may be useful in the interpretation of the kinetics of certain HVEM irradiation-induced processes in alloys.

  4. Effects of neutron irradiation to 63 dpa on the properties of various commercial copper alloys

    International Nuclear Information System (INIS)

    Brager, H.R.

    1985-04-01

    High purity copper and six commercial copper alloys were neutron irradiated to 47 and 63 dpa at about 450 0 C in the FFTF. Immersion density measurements showed a wide range of swelling behavior after irradiation to 63 dpa. At one extreme was CuBe in the aged and tempered (AT) condition which had densified slightly. At the other extreme was 20% CW Cu-0.1% Ag which swelled over 45%. Electrical resistivity measurements followed trends similar to previously published results for the same alloys irradiated to 16 dpa: a continued change in conductivity with fluence which appears to relate to void formation, transmutation products and coarsening of second phase precipitates. These results were compared with electrical conductivity of unirradiated alloys examined after aging for 10,000 hours. The most irradiation resistant high-conductivity copper alloys examined after 63 dpa are A125 and MZC. Cu-2.0Be, only a moderate-conductivity alloy, exhibits very consistent irradiation resistant properties

  5. 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.

  6. Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

    International Nuclear Information System (INIS)

    Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

    2007-01-01

    Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

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

    Science.gov (United States)

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

    2015-04-01

    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.

  8. 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)

  9. Concentration dependence of solute atoms on vacancy cluster formation in neutron irradiated Ni alloy

    International Nuclear Information System (INIS)

    Sato, K.; Itoh, D.; Yoshiie, T.; Xu, Q.

    2007-01-01

    Full text of publication follows: One dimensional (1-D) motion of interstitial clusters is important for the microstructural evolution in metals. The movement of interstitial clusters was often observed in neutron irradiated metals by transmission electron microscopy (TEM). Alloying elements are expected to affect the motion of interstitial clusters. Yoshiie et al. have studied the effect of alloying elements in Ni. For example, in neutron irradiated pure Ni, well-developed dislocation networks and voids were observed at 573 K at a dose of 0.026 dpa by TEM. After the addition of 2at.%Si (-5.81% volume size factor to Ni) and Sn (74.08% volume size factor), no voids were detected by TEM observation and positron lifetime measurement. Alloying elements of Si and Sn were expected to prevent the 1-D motion of the interstitial clusters. In this study, the concentration dependence of alloying elements on the 1-D motion of the interstitial clusters was investigated by positron annihilation lifetime measurements, and the microstructural evolution was discussed. Specimens irradiated were 99.99 pure Ni (Johnson Matthey) and Ni based binary alloys, which contain Si, Cu, Ge and Sn as solute atoms. The concentration of solute atoms was 0.05at.%o, 0.3at.% and 2at.%. Neutron irradiation was performed with the Kyoto University Reactor (KUR) and Japan materials testing reactor (JMTR) at Japan Atomic Energy Agency. Neutron dose was 6x10 -5 -1x10 -2 dpa at KUR, and 8x10 -3 -0.3 dpa at JMTR. Irradiation temperature was 573 K at KUR and 563 K at JMTR. After the neutron irradiation, positron annihilation lifetime measurements were performed at room temperature. Microvoids were detected in pure Ni, Ni-0.05%Si, Ni-0.05%Sn, Ni-Cu and Ni-Ge alloys. In Ni-Si and Ni-Sn alloys, the size of microvoids decreased as the concentration of solute atoms increased. This is because the frequency of 1-D motion of the interstitial clusters depends on the alloy concentration. High concentration of alloying

  10. Concentration dependence of solute atoms on vacancy cluster formation in neutron irradiated Ni alloy

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K.; Itoh, D.; Yoshiie, T.; Xu, Q. [Kyoto Univ., Research Reactor Institute, Osaka (Japan)

    2007-07-01

    Full text of publication follows: One dimensional (1-D) motion of interstitial clusters is important for the microstructural evolution in metals. The movement of interstitial clusters was often observed in neutron irradiated metals by transmission electron microscopy (TEM). Alloying elements are expected to affect the motion of interstitial clusters. Yoshiie et al. have studied the effect of alloying elements in Ni. For example, in neutron irradiated pure Ni, well-developed dislocation networks and voids were observed at 573 K at a dose of 0.026 dpa by TEM. After the addition of 2at.%Si (-5.81% volume size factor to Ni) and Sn (74.08% volume size factor), no voids were detected by TEM observation and positron lifetime measurement. Alloying elements of Si and Sn were expected to prevent the 1-D motion of the interstitial clusters. In this study, the concentration dependence of alloying elements on the 1-D motion of the interstitial clusters was investigated by positron annihilation lifetime measurements, and the microstructural evolution was discussed. Specimens irradiated were 99.99 pure Ni (Johnson Matthey) and Ni based binary alloys, which contain Si, Cu, Ge and Sn as solute atoms. The concentration of solute atoms was 0.05at.%o, 0.3at.% and 2at.%. Neutron irradiation was performed with the Kyoto University Reactor (KUR) and Japan materials testing reactor (JMTR) at Japan Atomic Energy Agency. Neutron dose was 6x10{sup -5}-1x10{sup -2} dpa at KUR, and 8x10{sup -3} -0.3 dpa at JMTR. Irradiation temperature was 573 K at KUR and 563 K at JMTR. After the neutron irradiation, positron annihilation lifetime measurements were performed at room temperature. Microvoids were detected in pure Ni, Ni-0.05%Si, Ni-0.05%Sn, Ni-Cu and Ni-Ge alloys. In Ni-Si and Ni-Sn alloys, the size of microvoids decreased as the concentration of solute atoms increased. This is because the frequency of 1-D motion of the interstitial clusters depends on the alloy concentration. High

  11. Deformation mechanism maps for pure iron, corrosion resistant austenitic steels and a low-alloy carbon steel

    International Nuclear Information System (INIS)

    Frost, H.Y.; Ashby, M.F.

    1980-01-01

    Principles of construction of deformation mechanisms charts for iron base alloys are presented. Deformation mechanisms charts for pure iron, 316 and 314 stainless steels, a ferritic steel with 1% Cr, Mo, V are given, examples of the charts application being provided. The charts construction is based, when it is possible, on the state equations, deduced from theoretical models and satisfying experimental data. The charts presented should be considered as an attempt to unite the main regularities of the theory of dislocations and diffusion with the observed experimental picture of plastic deformation and creep of commercial steels [ru

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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)

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

    International Nuclear Information System (INIS)

    Kamyshendo, V.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Wiffen, F.W.

    1984-01-01

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

  19. 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)

  20. 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...

  1. α′ precipitation in neutron-irradiated Fe–Cr alloys

    International Nuclear Information System (INIS)

    Bachhav, Mukesh; Robert Odette, G.; Marquis, Emmanuelle A.

    2014-01-01

    Graphical abstract: -- A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed α′ precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of α′ precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested

  2. Data collection on the effect of irradiation on the mechanical properties of austenitic stainless steels and weld metals

    International Nuclear Information System (INIS)

    Picker, C.; Wareing, J.; Tavassoli, A.A.

    1995-01-01

    Data on the influence of low dose irradiation on the mechanical properties of structural steels (Types 304, 316, 316L, 316H and 316L(N) and associated weld metals) at temperatures from 20 deg. C to 750 deg. C, have been compiled from published literature and the results of British, Dutch, French and German Laboratories. The preliminary results, which cover the dose range from 0 to 5 displacements per atom (and/or up to 2 appm helium) are presented as comparisons between irradiated and unirradiated control data, covering a range of strength and cyclic properties. The results show that low dose irradiation can have a significant influence on the properties ranging from increases in 0.2% proof stress to decreases in stress rupture strength and ductility. More detailed investigations of the significant factors on the individual properties will be completed in the future. (author). 13 figs, 1 tab

  3. Nanocluster irradiation evolution in Fe-9%Cr ODS and ferritic-martensitic alloys

    Science.gov (United States)

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

    2017-12-01

    The objective of this study is to evaluate the influence of dose rate and cascade morphology on nanocluster evolution in a model Fe-9%Cr oxide dispersion strengthened steel and the commercial ferritic/martensitic (F/M) alloys HCM12A and HT9. We present a large, systematic data set spanning the three alloys, three irradiating particle types, four orders of magnitude in dose rate, and doses ranging 1-100 displacements per atom over 400-500 °C. Nanoclusters are characterized using atom probe tomography. ODS oxide nanoclusters experience partial dissolution after irradiation due to inverse Ostwald ripening, while F/M nanoclusters undergo Ostwald ripening. Damage cascade morphology is indicative of nanocluster number density evolution. Finally, the effects of dose rate on nanocluster morphology provide evidence for a temperature dilation theory, which purports that a negative temperature shift is necessary for higher dose rate irradiations to emulate nanocluster evolution in lower dose rate irradiations.

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  6. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B N; Edwards, D J; Horsewell, A; Toft, P

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

  7. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    International Nuclear Information System (INIS)

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

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe 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, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

  8. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  9. Relation between the punctual defects dynamics and irradiation growth in Zr and its alloys

    International Nuclear Information System (INIS)

    Sarce, A.; Monti, A.M.; Smetniansky-De Grande, N.

    1990-01-01

    The knowledge of the static properties punctual defects and their interaction with the drains present in the material, such as dislocations and grain borders, is of fundamental importance for the correct prediction of the reactor's components behaviour under irradiation. This work calculates the Zr alloys growth under irradiation and results are compared with the stretching values of the cooling channels measured at the Atucha I nuclear power plant. (Author) [es

  10. Short-range order of amorphous FeNiB alloy after neutron irradiation

    International Nuclear Information System (INIS)

    Miglierini, M.; Sitek, J.; Baluch, S.; Cirak, J.; Lipka, J.

    1990-01-01

    Transmission Moessbauer spectroscopy was used to study irradiation-induced changes in the short-range order of an amorphous Fe 80-x Ni x B 20 alloy. Neutron irradiation led to an increase of the width of a hyperfine field distribution implying atomic rearrangement towards disordering. Changes in a mean value of a HFD and Moessbauer line areas can be associated with a reorientation of spins due to radiation damage. (orig.)

  11. 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

  12. The amorphous phase transition in irradiated NiTi alloy

    International Nuclear Information System (INIS)

    Brimhall, J.L.; Kissinger, H.E.; Pelton, A.R.

    1985-01-01

    Observed supralinear dose dependence for the amorphous transformation during irradiation of NiTi is compatible with a cascade overlap model for heavy ion (2.5 MeV Ni + , 6 MeV Ta +++ ) irradiations. A model based on total defect build-up, however, is necessary to explain the amorphous transition induced by electron irradiation and can also be applied to heavy ion irradiation. The cascade effects in this latter model are manifested by non-uniform defect distribution in the lattice. The defect build-up model requires a high activation energy for interstitial migration which is not incompatible with recent findings. The form of the temperature dependence can also be rationalized using a defect build-up model (amorphous phase transition, heavy-ion irradiation, electron irradiation, NiTi, defect build-up, cascade overlap). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-01

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

  14. 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

  15. Laser irradiation effects on the surface, structural and mechanical properties of Al-Cu alloy 2024

    Science.gov (United States)

    Yousaf, Daniel; Bashir, Shazia; Akram, Mahreen; kalsoom, Umm-i.-; Ali, Nisar

    2014-02-01

    Laser irradiation effects on surface, structural and mechanical properties of Al-Cu-Mg alloy (Al-Cu alloy 2024) have been investigated. The specimens were irradiated for various fluences ranging from 3.8 to 5.5 J/cm2 using an Excimer (KrF) laser (248 nm, 18 ns, 30 Hz) under vacuum environment. The surface and structural modifications of the irradiated targets have been investigated by scanning electron microscope (SEM) and X-ray diffractometer (XRD), respectively. SEM analysis reveals the formation of micro-sized craters along the growth of periodic surface structures (ripples) at their peripheries. The size of the craters initially increases and then decreases by increasing the laser fluence. XRD analysis shows an anomalous trend in the peak intensity and crystallite size of the specimen irradiated for various fluences. A universal tensile testing machine and Vickers microhardness tester were employed in order to investigate the mechanical properties of the irradiated targets. The changes in yield strength, ultimate tensile strength and microhardness were found to be anomalous with increasing laser fluences. The changes in the surface and structural properties of Al-Cu alloy 2024 after laser irradiation have been associated with the changes in mechanical properties.

  16. 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

  17. 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.

  18. Short distance ordering kinetics and vacancy and autointerstitial characteristics in gamma Fe Ni Cr alloys

    International Nuclear Information System (INIS)

    Berroudji, S.A.

    1988-01-01

    Characteristics of vacancies formation and migration are studied in 3 austenitic steels with 16% of chromium and respectively 25, 45 and 75% of nickel. Influence of impurities is examined. The 3 alloys are irradiated with a Van de Graaff from 508 K to 830 K defect migration is studied and discussed in relationship with swelling [fr

  19. Effect of irradiation damage and helium on the swelling and structure of vanadium-base alloys

    International Nuclear Information System (INIS)

    Chung, H.M.; Loomis, B.A.; Smith, D.L.

    1993-12-01

    Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420--600C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti 5 Si 3 promotes good resistance to swelling of the Ti-containing alloys and it was concluded that Ti of >3 wt.% and 400--1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. V-20Ti doped with B exhibited somewhat higher swelling because of He generation. Lithium atoms, generated from transmutation of 10 B, formed γ-LiV 2 O 5 precipitates and did not seem to produce undesirable effects on mechanical properties

  20. 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.

  1. Zirconium hydrides and Fe redistribution in Zr-2.5%Nb alloy under ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Idrees, Y.; Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada); Cui, J.; Shek, G.K. [Kinetrics, Mississauga, ON (Canada); Daymond, M.R., E-mail: daymond@queensu.ca [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada, K7L 3N6 (Canada)

    2016-11-15

    Zr-2.5%Nb alloy is used to fabricate the pressure tubes of the CANDU reactor. The pressure tube is the primary pressure boundary for coolant in the CANDU design and is susceptible to delayed hydride cracking, reduction in fracture toughness upon hydride precipitation and potentially hydride blister formation. The morphology and nature of hydrides in Zr-2.5%Nb with 100 wppm hydrogen has been investigated using transmission electron microscopy. The effect of hydrides on heavy ion irradiation induced decomposition of the β phase has been reported. STEM-EDX mapping was employed to investigate the distribution of alloying elements. The results show that hydrides are present in the form of stacks of different sizes, with length scales from nano- to micro-meters. Heavy ion irradiation experiments at 250 °C on as-received and hydrided Zr-2.5%Nb alloy, show interesting effects of hydrogen on the irradiation induced redistribution of Fe. It was found that Fe is widely redistributed from the β phase into the α phase in the as-received material, however, the loss of Fe from the β phase and subsequent precipitation is retarded in the hydrided material. This preliminary work will further the current understanding of microstructural evolution of Zr based alloys in the presence of hydrogen. - Graphical abstract: STEM HAADF micrographs at low magnification showing the hydride structure in Zr-2.5Nb alloy.

  2. Work hardening characteristics of gamma-ray irradiated Al-5356 alloy

    International Nuclear Information System (INIS)

    Saad, G.; Fayek, S.A.; Fawzy, A.; Soliman, H.N.; Nassr, E.

    2014-01-01

    Effects of γ-irradiation and deformation temperatures on the hardening behavior of Al-5356 alloy have been investigated by means of stress–strain measurements. Wire samples irradiated with different doses (ranging from 500 to 2000 kGy) were strained at different deformation temperatures T w (ranging from 303 to 523 K) and a constant strain rate of 1.5×10 −3 s −1 . The effect of γ-irradiation on the work-hardening parameters (WHP): yield stress σ y , fracture stress σ f , total strain ε T and work-hardening coefficient χ p of the given alloy was studied at the applied deformation temperature range. The obtained results showed that γ-irradiation exhibited an increase in the WHP of the given alloy while the increase in its deformation temperature showed a reverse effect. The mean activation energy of the deformation process was calculated using an Arrhenius-type relation, and was found to be ∼80 kJ/mole, which is close to that of grain boundary diffusion in aluminum alloys

  3. Effect of solute atom concentration on vacancy cluster formation in neutron-irradiated Ni alloys

    Science.gov (United States)

    Sato, Koichi; Itoh, Daiki; Yoshiie, Toshimasa; Xu, Qiu; Taniguchi, Akihiro; Toyama, Takeshi

    2011-10-01

    The dependence of microstructural evolution on solute atom concentration in Ni alloys was investigated by positron annihilation lifetime measurements. The positron annihilation lifetimes in pure Ni, Ni-0.05 at.%Si, Ni-0.05 at.%Sn, Ni-Cu, and Ni-Ge alloys were about 400 ps even at a low irradiation dose of 3 × 10 -4 dpa, indicating the presence of microvoids in these alloys. The size of vacancy clusters in Ni-Si and Ni-Sn alloys decreased with an increase in the solute atom concentration at irradiation doses less than 0.1 dpa; vacancy clusters started to grow at an irradiation dose of about 0.1 dpa. In Ni-2 at.%Si, irradiation-induced segregation was detected by positron annihilation coincidence Doppler broadening measurements. This segregation suppressed one-dimensional (1-D) motion of the interstitial clusters and promoted mutual annihilation of point defects. The frequency and mean free path of the 1-D motion depended on the solute atom concentration and the amount of segregation.

  4. Neutronic irradiation effect in FeNi alloys, observed by magnetic measurements

    International Nuclear Information System (INIS)

    Sciani, V.; Lucki, G.

    1986-01-01

    In this work some aspects of radiation damage are analysed through the influence of neutron irradiation on magnetic properties of FeNi alloys. The main points emphasized are: radiation enhanced diffusion, determination of the activation energy for diffusion process and vacancies supersaturation, which is an important parameter from technological point of view and a necessary condition for the void formation. (Author) [pt

  5. Postirradiation tensile properties of Mo and Mo alloys irradiated with 600 MeV protons

    International Nuclear Information System (INIS)

    Mueller, G.V.; Gavillet, D.; Victoria, M.; Martin, J.L.

    1994-01-01

    Tensile specimens of pure Mo and Mo-5 Re, Mo-41 Re and TZM alloys have been irradiated with 600 MeV protons in the PIREX facility at 300 and 660 K to 0.5 dpa. Results of the postirradiation tensile testing show a strong radiation hardening and a severe loss of ductility for all the materials tested at room temperature. ((orig.))

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

    International Nuclear Information System (INIS)

    Soukieh, Mohamad; Al-Mohamad, Ali

    1993-12-01

    The use of metallic thin films for studying the processes which take place during ion irradiation has recently increased. For example, ion implantation is widely used to study the structural defects in transition metallic thin films such as (Fe, Ni, Co), because it can simulate the effects occurring in nuclear reactors during neutron irradiation especially the swelling of reactor materials. The swelling of metals and alloys is strongly related to the material structure and to the irradiation conditions. The general feature of formation of structural defects as a function of irradiation dosage and annealing temperature is well known. However, the detailed mechanisms are still not well understood. For example, the swelling of iron alloy with 30-35% nickel is very small in comparison with other Ni concentrations, and there is no clear information on the possibility of phase transitions in fe-Ni alloys during irradiation. The aim of this work is to study the phase-structural changes in Fe-30% Ni implanted by high dose of argon ions. The effect of irradiation with low energy argon ions (40 KeV, and fluences of 10.E15 to 10.E17 ions/cm) on the deposited thin films of Fe-30% Ni alloy was investigated using RBS and TEM techniques. The thicknesses of these films were about 65+-10 nm deposited on ceramic, KBr, and Be fiols substrates. Gas bubble formation and profile distribution of the implanted argon ions were investigated. Formation of an ordered phase Fe 3 Ni during irradiation appears to inhibit gas bubble formations in the film structure. (author). 17 refs., 15 figs., 7 tabs

  7. 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.

  8. Effects of solute elements on irradiation hardening and microstructural evolution in low alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Katsuhiko, E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System Inc., 64 Sata, Mihama 919-1205 (Japan); Ohkubo, Tadakatsu, E-mail: OHKUBO.Tadakatsu@nims.go.jp [National Institute for Materials Science, 1-2-1 Sengen, Tsukuba 305-0047 (Japan); Fukuya, Koji, E-mail: fukuya@inss.co.jp [Institute of Nuclear Safety System Inc., 64 Sata, Mihama 919-1205 (Japan)

    2011-10-01

    The effects of the elements Mn, Ni, Si and Cu on irradiation hardening and microstructural evolution in low alloy steels were investigated in ion irradiation experiments using five kinds of alloys prepared by removing Mn, Ni and Si from, and adding 0.05 wt.%Cu to, the base alloy (Fe-1.5Mn-0.5Ni-0.25Si). The alloy without Mn showed less hardening and the alloys without Ni or Si showed more hardening. The addition of Cu had hardly any influence on hardening. These facts indicated that Mn enhanced hardening and that Ni and Si had some synergetic effects. The formation of solute clusters was not confirmed by atom probe (AP) analysis, whereas small dislocation loops were identified by TEM observation. The difference in hardening between the alloys with and without Mn was qualitatively consistent with loop formation. However, microstructural components that were not detected by the AP and TEM were assumed to explain the hardening level quantitatively.

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

    CERN Document Server

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

    2003-01-01

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

  10. 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)

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

    International Nuclear Information System (INIS)

    Smith, D.L.

    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 ampersand D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  14. 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)

  15. Microstructure and Mechanical Properties of n-irradiated Fe-Cr Model Alloys

    International Nuclear Information System (INIS)

    Matijasevic, Milena; Al Mazouzi, Abderrahim

    2008-01-01

    High chromium ( 9-12 wt %) ferritic/martensitic steels are candidate structural materials for future fusion reactors and other advanced systems such as accelerator driven systems (ADS). Their use for these applications requires a careful assessment of their mechanical stability under high energy neutron irradiation and in aggressive environments. In particular, the Cr concentration has been shown to be a key parameter to be optimized in order to guarantee the best corrosion and swelling resistance, together with the least embrittlement. In this work, the characterization of the neutron irradiated Fe-Cr model alloys with different Cr % with respect to microstructure and mechanical tests will be presented. The behavior of Fe-Cr alloys have been studied using tensile tests at different temperature range ( from -160 deg. C to 300 deg. C). Irradiation-induced microstructure changes have been studied by TEM for two different irradiation doses at 300 deg. C. The density and the size distribution of the defects induced have been determined. The tensile test results indicate that Cr content affects the hardening behavior of Fe-Cr binary alloys. Hardening mechanisms are discussed in terms of Orowan type of approach by correlating TEM data to the measured irradiation hardening. (authors)

  16. 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

  17. 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.

  18. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  20. 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

  1. 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

  2. Effect of γ-IRRADIATION on the Mechanical Properties of Al-Cu Alloy

    Science.gov (United States)

    Abo-Elsoud, M.; Ismail, H.; Sobhy, Maged S.

    SEM observations and Vickers hardness tests were performed to identify the irradiation effects. γ-irradiation effect during the aging hardening process can be explained depending on the composition of the alloy and is used to derive quantitative information on the kinetics of the transformation precipitates. Increasing the Cu content of an Al-Cu alloy can improve the aging hardness. The present results of the hardness behavior, with SEM observations of surveillance specimens at different doses, suggest that the radiation-induced defects are probably complex valence-solute clusters. These clusters act as nuclei for the precipitation of θ-Al2Cu type. This can be effectively utilized to study the systematics of nucleation of precipitates at vacancy-type defects. γ-irradiation probably plays the key role in defects responsible for material strengthening and embrittlement.

  3. Surface Properties of a Nanocrystalline Fe-Ni-Nb-B Alloy After Neutron Irradiation

    Science.gov (United States)

    Pavùk, Milan; Sitek, Jozef; Sedlačková, Katarína

    2014-09-01

    The effect of neutron radiation on the surface properties of the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy was studied. Firstly, amorphous (Fe0.25Ni0.75)81Nb7B12 ribbon was brought by controlled annealing to the nanocrystalline state. After annealing, the samples of the nanocrystalline ribbon were irradiated in a nuclear reactor with neutron fluences of 1×1016cm-2 and 1 × 1017cm-2 . By utilizing the magnetic force microscopy (MFM), topography and a magnetic domain structure were recorded at the surface of the ribbon-shaped samples before and after irradiation with neutrons. The results indicate that in terms of surface the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy is radiation-resistant up to a neutron fluence of 1 × 1017cm-2 . The changes in topography observed for both irradiated samples are discussed

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    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.

  6. 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...

  7. 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).

  8. Microstructure and tensile properties of neutron-irradiated (FE061Ni039)3V ordered alloy

    International Nuclear Information System (INIS)

    Braski, D.N.

    1982-01-01

    Small tensile specimens of the (Fe 0 61 Ni 0 39 ) 3 V long-range-ordered alloy were irradiated in the ORR to 4 dpa at 523, 623, and 823 K and subsequently tested at the same respective temperatueres. The alloy remained ordered after irradiation at all three temperatures. Irradiation at 523 and 623 K increased the yield strength of the material by producing Frank loops in the microstructure and reduced the total elongation. The low strain hardening observed was attributed to planar slip and the absence of cross slip. Irradiation at 823 K embrittled the alloy. Premature failure was apparently initiated by helium bubbles on sigma phase boundaries which grew rapidly during the test to form microcracks. Fracture occurred after a microcrack propagated across grain boundaries that were weakened by helium and possible sulfur. New LRO alloys without sigma phase should perform better under neutron irradiation

  9. Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

    Science.gov (United States)

    Aydogan, E.; Weaver, J. S.; Maloy, S. A.; El-Atwani, O.; Wang, Y. Q.; Mara, N. A.

    2018-05-01

    FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α‧ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

  10. 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

  11. Extra spots in the electron diffraction patterns of neutron irradiated zirconium and its alloys

    International Nuclear Information System (INIS)

    Madden, P.K.

    1977-01-01

    Specimens of neutron irradiated zirconium and its alloys were examined in the transmission electron microscope. Groups of extra spots, often exhibiting four-fold symmetry, were observed in thin foil electron diffraction patterns of these specimens. The 'extra-spot' structure, like the expected black-dot/small scale dislocation loop neutron irradiated damage, is approximately 100 A in size. Its nature is uncertain. It may be related to irradiation damage or to some artefact introduced during specimen preparation. If it is the latter, then published irradiation damage defect size distributions and determined irradiation growth strains of other investigators, may require modification. The present inconclusive results indicate that extra-spot structure is likely to consist of oxide particles, but may correspond to hydride precipitation or decoration effects, or even, to electron beam effects. (author)

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  14. Effect of minor elements on microstructure evolution in Ni alloys irradiated with neutrons

    International Nuclear Information System (INIS)

    Xu, Q.; Yoshiie, T.

    2001-01-01

    The minor elements, Si (-5.81%), Cu (7.18%), Ge (14.76%) and Sn (74.08%) were chosen to investigate the effects of volume size factor as shown in the parentheses on void swelling in neutron irradiated Ni alloys. Neutron irradiation temperature and dose were changed widely from 473 K to 703 K, and 0.001 dpa to 1 dpa, respectively. Voids were observed by transmission electron microscopy (TEM) in Ni even after a very small irradiation dose of 0.026 dpa at 573 K. With increasing dose, the number density of voids was nearly constant while void size increased. The microstructure evolution in Ni-2 at%Cu and Ni-2 at%Ge alloys was similar to that in Ni. However, in Ni-2 at%Si and Ni-2 at%Sn alloys, no voids were observed by TEM even at 703 K to 1 dpa. The minor elements, Si and Sn, play an important role for the suppression of vacancy clusters. Vacancies are annihilated by mutual recombination with interstitials in Si and Sn added alloys. (orig.)

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

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

    1993-01-01

    A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ''superclean'' composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature (ΔTT4 41J or ΔTT 47J ) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest ΔTT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials

  17. 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

  18. 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

  19. Neutron irradiation effects of iron alloys and ceramics

    International Nuclear Information System (INIS)

    Kuramoto, Eiichi; Takenaka, Minoru; Hasegawa, Masayuki.

    1991-01-01

    Positron annihilation angular correlation measurements have been performed for the neutron irradiated various metals and ceramics in order to obtain the information of the microvoids and positronium formation in them. Positronium (Ps) formation was observed in Nb containing a small amount of oxygen and Fe-15%Cr-16%Ni-0.006%B 10 . In practical steels such as JPCA and JFMS no Ps formation was observed. High temperature deformation might induce microvoids into metals, but the positron annihilation angular correlation measurements could not confirm this. In non-metallic materials neutron irradiated no Ps formation has so far been observed. (author)

  20. 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.

  1. APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Blavette, D.

    1996-01-01

    Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

  2. Microstructural examination of several commercial ferritic alloys irradiated to high fluence

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1981-01-01

    Microstructural observations are reported for a series of five commercial ferritic alloys, 2 1/4 Cr-1 Mo, H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650 0 C and to a maximum fluence of 17.6 x 10 22 n/cm 2 (E > 0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400 0 C to 14.0 x 10 22 n/cm 2 . A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6 C, Mo 2 C, Chi, Laves, M 23 C 6 , α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation

  3. Microstructural examination of several commercial ferritic alloys irradiated to high fluence

    Science.gov (United States)

    Gelles, D. S.

    Microstructural observations are reported for a series of five commercial ferritic alloys, 2 {1}/{4}Cr-1Mo , H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650°C and to a maximum fluence of 1.76 × 10 23 n/cm 2 (E >0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400°C to 1.40 × 10 23 n/cm 2. A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M 6C, Mo 2C, Chi, Laves, M 23C 6, α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation.

  4. Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Wiffen, F.W.

    1985-01-01

    The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs

  5. Alloy development for irradiation performance. Quarterly progress report for period ending March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-06-01

    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.

  6. Effects of irradiation on low-activation ferritic alloys to 45 dpa

    International Nuclear Information System (INIS)

    Gelles, D.S.; Hamilton, M.L.

    1986-06-01

    Nine low activation ferritic alloys covering the range 2 to 12Cr with alloying additions of tungsten and/or vanadium have been irradiated to intermediate fluences of 30 to 45 dpa and tensile tested or examined by transmission electron microscopy in order to determine the effect of increasing neutron dose on properties and microstructure. Changes in properties and microstructure are for the most part completed within 10 dpa but swelling and dislocation evolution continue with increasing dose at 420/degree/C and subgrain coarsening occurs at 600/degree/C. 9 refs., 7 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    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

  8. Power ultrasound irradiation during the alkaline etching process of the 2024 aluminum alloy

    Science.gov (United States)

    Moutarlier, V.; Viennet, R.; Rolet, J.; Gigandet, M. P.; Hihn, J. Y.

    2015-11-01

    Prior to any surface treatment on an aluminum alloy, a surface preparation is necessary. This commonly consists in performing an alkaline etching followed by acid deoxidizing. In this work, the use of power ultrasound irradiation during the etching step on the 2024 aluminum alloy was studied. The etching rate was estimated by weight loss, and the alkaline film formed during the etching step was characterized by glow discharge optical emission spectrometry (GDOES) and scanning electron microscope (SEM). The benefit of power ultrasound during the etching step was confirmed by pitting potential measurement in NaCl solution after a post-treatment (anodizing).

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

    International Nuclear Information System (INIS)

    1981-07-01

    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

  10. 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

  11. 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.

  12. 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

  13. Austenitic stainless steels with cryogenic resistance

    International Nuclear Information System (INIS)

    Tarata, Daniela Florentina

    1999-01-01

    The most used austenitic stainless steels are alloyed with chromium and nickel and have a reduced carbon content, usually lower than 0.1 % what ensures corresponding properties for processing by plastic deformation at welding, corrosion resistance in aggressive environment and toughness at low temperatures. Steels of this kind alloyed with manganese are also used to reduce the nickel content. By alloying with manganese which is a gammageneous element one ensures the stability of austenites. Being cheaper these steels may be used extensively for components and equipment used in cryogenics field. The best results were obtained with steels of second group, AMnNi, in which the designed chemical composition was achieved, i.e. the partial replacement of nickel by manganese ensured the toughness at cryogenic temperatures. If these steels are supplementary alloyed, their strength properties may increase to the detriment of plasticity and toughness, although the cryogenic character is preserved

  14. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Parish, C.M., E-mail: parishcm@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Unocic, K.A.; Tan, L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zinkle, S.J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Kondo, S. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto, 611-0011 (Japan); Snead, L.L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Hoelzer, D.T.; Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)