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Sample records for irradiated ferritic steels

  1. Comminuting irradiated ferritic steel

    DOEpatents

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

    1985-01-01

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

  2. Characterization of Irradiated Nanostructured Ferritic Steels

    SciTech Connect

    Bentley, James; Hoelzer, David T; Tanigawa, H.; Yamamoto, T.; Odette, George R.

    2007-01-01

    The past decade has seen the development of a new class of mechanically alloyed (MA) ferritic steels with outstanding mechanical properties that come, at least in part, from the presence of high concentrations (>10{sup 23} m{sup -3}) of Ti-, Y-, and O-enriched nanoclusters (NC). From the outset, there has been much interest in their potential use for applications to fission and proposed fusion reactors, not only because of their attractive high-temperature strength, but also because the presence of NC may result in a highly radiation-resistant material by efficiently trapping point defects to enhance recombination. Of special interest for fusion applications is the potential of NC to trap transmutation-produced He in high concentrations of small cavities, rather than in fewer but larger cavities that lead to greater radiation-induced swelling and other degraded properties.

  3. Prediction of yield stress in highly irradiated ferritic steels

    NASA Astrophysics Data System (ADS)

    Windsor, Colin G.; Cottrell, Geoff; Kemp, Richard

    2008-03-01

    The design of any fusion power plant requires information on the irradiation hardening of low-activation ferritic/martensitic steels beyond the range of most present measurements. Neural networks have been used by Kemp et al (J. Nucl. Mater. 348 311-28) to model the yield stress of some 1811 irradiated alloys. The same dataset has been used in this study, but has been divided into a training set containing the majority of the dataset with low irradiation levels, and a test set which contains just those alloys which have been irradiated above a given level. For example some 4.5% of the alloys were irradiated above 30 displacements per atom. For this 'prediction' problem it is found that simpler networks with fewer inputs are advantageous. By using target-driven dimensionality reduction, linear combinations of the atomic inputs reduce the test residual below that achievable by adding inputs from single atoms. It is postulated that these combinations represent 'mechanisms' for the prediction of irradiated yield stress.

  4. Impact toughness of irradiated reduced-activation ferritic steels*1

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

    Eight chromium-tungsten steels ranging from 2.25 to 12 wt% Cr were irradiated at 365°C to 13-14 dpa in the Fast Flux Test Facility. Post irradiation Charpy impact tests showed a loss of toughness for all steels, as measured by an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The most irradiation-resistant steels were two 9% Cr steels: the DBTT of a 9Cr-2W-0.25V-0.1C steel increased 29°C, and for the same composition with an addition of 0.07% Ta the DBTT increased only 15°C. This is the smallest shift ever observed for such a steel irradiated to these levels. The other steels developed shifts in DBTT of 100 to 300°C. A 2.25% Cr steel with 2% W, 0.25% V, and 0.1% C was less severely affected by irradiation than 2.25% Cr steels with 0.25% V and no tungsten, 2% W and no vanadium, and with 1% W and 0.25% V. Irradiation resistance appears to be associated with microstructure, and microstructural manipulation may lead to improved properties.

  5. Irradiation hardening of ODS ferritic steels under helium implantation and heavy-ion irradiation

    NASA Astrophysics Data System (ADS)

    Zhang, Hengqing; Zhang, Chonghong; Yang, Yitao; Meng, Yancheng; Jang, Jinsung; Kimura, Akihiko

    2014-12-01

    Irradiation hardening of ODS ferritic steels after multi-energy He-ion implantation, or after irradiation with energetic heavy ions including Xe and Bi-ions was investigated with nano-indentation technique. Three kinds of high-Cr ODS ferritic steels including the commercial MA956 (19Cr-3.5Al), the 16Cr-0.1Ti and the 16Cr-3.5Al-0.1Zr were used. Data of nano-hardness were analyzed with an approach based on Nix-Gao model. The depth profiles of nano-hardness can be understood by the indentation size effect (ISE) in specimens of MA956 implanted with multi-energy He-ions or irradiated with 328 MeV Xe ions, which produced a plateau damage profile in the near-surface region. However, the damage gradient overlaps the ISE in the specimens irradiated with 9.45 Bi ions. The dose dependence of the nano-hardness shows a rapid increase at low doses and a slowdown at higher doses. An 1/2-power law dependence on dpa level is obtained. The discrepancy in nano-hardness between the helium implantation and Xe-ion irradiation can be understood by using the average damage level instead of the peak dpa level. Helium-implantation to a high dose (7400 appm/0.5 dpa) causes an additional hardening, which is possibly attributed to the impediment of motion dislocations by helium bubbles formed in high concentration in specimens.

  6. Reduction method of DBTT shift due to irradiation for reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Okubo, N.; Ando, M.; Yamamoto, T.; Takada, F.

    2010-03-01

    The method for reducing irradiation-induced DBTT shift of reduced-activation ferritic/martensitic steels was examined. F82H-LN (low nitrogen, 20 ppm), F82H+60 ppm 11B+200 ppmN and F82H+60 ppm 10B+200 ppmN steels tempered at 780 °C for 0.5 h were irradiated at 250 °C to 2 dpa, and the results for Charpy impact tests were analyzed. The upper shelf energy of F82H+ 11B+N steel was hardly changed by the irradiation, and DBTT shift was very small. From our research, DBTT shift due to irradiation can be reduced by the control of tempered conditions before irradiation, and it is found to be furthermore reduced by impurity doping with 60 ppm 11B and 200 ppmN to F82H steel.

  7. Effect of heat treatment and irradiation temperature on impact behavior of irradiated reduced-activation ferritic steels

    SciTech Connect

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

    1998-03-01

    Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to {approx}14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with {approx}25% {delta}-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

  8. Mechanical property changes of low activation ferritic/martensitic steels after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Kohno, Y.; Kohyama, A.; Hirose, T.; Hamilton, M. L.; Narui, M.

    Mechanical property changes of Fe- XCr-2W-0.2V,Ta ( X: 2.25-12) low activation ferritic/martensitic steels including Japanese Low Activation Ferritic/martensitic (JLF) steels and F82H after neutron irradiation were investigated with emphasis on Charpy impact property, tensile property and irradiation creep properties. Dose dependence of ductile-to-brittle transition temperature (DBTT) in JLF-1 (9Cr steel) irradiated at 646-700 K increased with irradiation up to 20 dpa and then decreased with further irradiation showing highest DBTT of 260 K at 20 dpa. F82H showed similar dose dependence in DBTT to JLF-1 with higher transition temperature than that of JLF-1 at the same displacement damage. Yield strength in JLF steels and F82H showed similar dose dependence to that of DBTT. Yield strength increased with irradiation up to 15-20 dpa and then decreased to saturate above about 40 dpa. Irradiation hardening in 7-9%Cr steels (JLF-1, JLF-3, F82H) were observed to be smaller than those in steels with 2.25%Cr (JLF-4) or 12%Cr (JLF-5). Dependences of creep strain on applied hoop stress and neutron fluence were measured to be 1.5 and 1, respectively. Temperature dependence of creep coefficient showed a maximum at about 700 K which was caused by irradiation induced void formation or irradiation enhanced creep deformation. Creep coefficient of F82H was larger than those of JLF steels above 750 K. This was considered to be caused by the differences in N and Ta concentration between F82H and JLF steels.

  9. Effects of neutron irradiation on microstructural evolution in candidate low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kohno, Yutaka; Kohyama, Akira; Yoshino, Masahiko; Asakura, Kentaro

    1994-09-01

    Fe-(2.25-12)Cr-2W-V, Ta low activation ferritic steels (JLF series steels) were developed in the fusion materials development program of Japanese universities. Microstructural observations, including precipitation response, were performed after neutron irradiation in the FFTF/MOTA. The preirradiation microstructure was stable after irradiation at low temperature (< 683 K). Recovery of martensitic lath structure and coarsening of precipitates took place above 733 K. Precipitates observed after irradiation were the same as those in unirradiated materials in 7-9Cr steels, and no irradiation induced phase was identified. The irradiation induced shift in DBTT in the 9Cr-2W steel proved to be very small which is a reflection of stable precipitation response in these steels. A high density of fine α' precipitates was observed in the 12Cr steel which might be responsible for the large irradiation hardening found in the 12Cr steel. Void formation was observed in 7-9Cr steels irradiated at 683 K, but the amount of void swelling was very small.

  10. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    SciTech Connect

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; Pakarinen, Janne; Wu, Yaqiao; Allen, Todd; Yang, Yong

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

  11. Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel

    DOE PAGES

    Li, Zhangbo; Lo, Wei-Yang; Chen, Yiren; ...

    2015-08-08

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019more » n/cm2 E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinod,al decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. Lastly, The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.« less

  12. Mechanical properties and microstructure of advanced ferritic-martensitic steels used under high dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Golovanov, V. N.; Bulanova, T. M.; Povstianko, A. V.; Fedoseev, A. E.; Goncharenko, Yu. D.; Ostrovsky, Z. E.

    Some results of the study of mechanical properties and structure of ferritic-martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ˜10 dpa at 400°C 0.1C-9Cr-1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

  13. Helium effects on the mechanical properties of neutron-irradiated Cr-Mo ferritic steels

    SciTech Connect

    Klueh, R.L.

    1990-01-01

    In the first wall of a fusion rector, large amounts of transmutation helium will be produced simultaneously with the displacement damage caused by high-energy neutrons from the fusion reaction. One method used to simulate irradiation effects for ferritic steels is to add nickel to the steels and irradiate them in a mixed-spectrum reactor. Fast neutrons in the spectrum produce displacement damage, while transmutation helium is produced by a two-step reaction of {sup 58}Ni with thermal neutrons. This technique has been used to investigate the effect of helium on tensile properties and toughness. Results from these studies are summarized.

  14. Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

    SciTech Connect

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

    1997-08-01

    Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.

  15. IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

    SciTech Connect

    Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

    2002-09-01

    Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

  16. Radiation hardening and deformation behavior of irradiated ferritic-martensitic steels

    SciTech Connect

    Robertson, J.P.; Klueh, R.L.; Rowcliffe, A.F.; Shiba, K.

    1998-03-01

    Tensile data from several 8--12% Cr alloys irradiated in the High Flux Isotope Reactor (HFIR) to doses up to 34 dpa at temperatures ranging from 90 to 600 C are discussed in this paper. One of the critical questions surrounding the use of ferritic-martensitic steels in a fusion environment concerns the loss of uniform elongation after irradiation at low temperatures. Irradiation and testing at temperatures below 200--300 C results in uniform elongations less than 1% and stress-strain curves in which plastic instability immediately follows yielding, implying dislocation channeling and flow localization. Reductions in area and total elongations, however, remain high.

  17. Low-temperature irradiation effects on tensile and Charpy properties of low-activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Shiba, Kiyoyuki; Hishinuma, Akimichi

    2000-12-01

    Tensile and Charpy properties of low-activation ferritic steel, F82H irradiated up to 0.8 dpa at low temperature below 300°C were investigated. The helium effect on these properties was also investigated using the boron isotope doping method. Neutron irradiation increased yield stress accompanied with ductility loss, and it also shifted the ductile-to-brittle transition temperature (DBTT) from -50°C to 0°C. Boron-doped F82H showed larger degradation in DBTT and ductility than boron-free F82H, while they had the same yield stress before and after irradiation.

  18. Irradiation-induced grain growth in nanocrystalline reduced activation ferrite/martensite steel

    SciTech Connect

    Liu, W. B.; Chen, L. Q.; Zhang, C. Yang, Z. G.; Ji, Y. Z.; Zang, H.; Shen, T. L.

    2014-09-22

    In this work, we investigate the microstructure evolution of surface-nanocrystallized reduced activation ferrite/martensite steels upon high-dose helium ion irradiation (24.3 dpa). We report a significant irradiation-induced grain growth in the irradiated buried layer at a depth of 300–500 nm, rather than at the peak damage region (at a depth of ∼840 nm). This phenomenon can be explained by the thermal spike model: minimization of the grain boundary (GB) curvature resulting from atomic diffusion in the cascade center near GBs.

  19. Metallography studies and hardness measurements on ferritic/martensitic steels irradiated in STIP

    NASA Astrophysics Data System (ADS)

    Zhang, H.; Long, B.; Dai, Y.

    2008-06-01

    In this work metallography investigations and microhardness measurements have been performed on 15 ferritic/martensitic (FM) steels and 6 weld metals irradiated in the SINQ Target Irradiation Program (STIP). The results demonstrate that all the steels have quite similar martensite lath structures. However, the sizes of the prior austenite grain (PAG) of these steels are quite different and vary from 10 to 86 μm. The microstructure in the fusion zones (FZ) of electron-beam welds (EBWs) of 5 steels (T91, EM10, MANET-II, F82H and Optifer-IX) is similar in respect to the martensite lath structure and PAG size. The FZ of the inert-gas-tungsten weld (TIGW) of the T91 steel shows a duplex structure of large ferrite gains and martensite laths. The microhardness measurements indicate that the normalized and tempered FM steels have rather close hardness values. The unusual high hardness values of the EBW and TIGW of the T91 steel were detected, which suggests that these materials are without proper tempering or post-welding heat treatment.

  20. Neutron irradiation effects on the ductile-brittle transition of ferritic/martensitic steels

    SciTech Connect

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

    1997-06-01

    Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement will be reviewed. Below 425-450{degrees}C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture. In addition to irradiation hardening, neutrons from the fusion reaction will produce large amounts of helium in the steels used to construct fusion power plant components. Tests to simulate the fusion environment indicate that helium can also affect the toughness. Steels are being developed for fusion applications that have a low DBTT prior to irradiation and then show only a small shift after irradiation. A martensitic 9Cr-2WVTa (nominally Fe-9Cr-2W-0.25V-0.07Ta-0.1C) steel had a much lower DBTT than the conventional 9Cr-1MoVNb steel prior to neutron irradiation and showed a much smaller increase in DBTT after irradiation. 27 refs., 5 figs., 1 tab.

  1. Technical Letter Report on the Cracking of Irradiated Cast Stainless Steels with Low Ferrite Content

    SciTech Connect

    Chen, Y.; Alexandreanu, B.; Natesan, K.

    2014-11-01

    Crack growth rate and fracture toughness J-R curve tests were performed on CF-3 and CF-8 cast austenite stainless steels (CASS) with 13-14% of ferrite. The tests were conducted at ~320°C in either high-purity water with low dissolved oxygen or in simulated PWR water. The cyclic crack growth rates of CF-8 were higher than that of CF-3, and the differences between the aged and unaged specimens were small. No elevated SCC susceptibility was observed among these samples, and the SCC CGRs of these materials were comparable to those of CASS alloys with >23% ferrite. The fracture toughness values of unirradiated CF-3 were similar between unaged and aged specimens, and neutron irradiation decreased the fracture toughness significantly. The fracture toughness of CF-8 was reduced after thermal aging, and declined further after irradiation. It appears that while lowering ferrite content may help reduce the tendency of thermal aging embrittlement, it is not very effective to mitigate irradiation-induced embrittlement. Under a combined condition of thermal aging and irradiation, neutron irradiation plays a dominant role in causing embrittlement in CASS alloys.

  2. Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tanigawa, H.; Hashimoto, N.; Sakasegawa, H.; Klueh, R. L.; Sokolov, M. A.; Shiba, K.; Jitsukawa, S.; Kohyama, A.

    2004-08-01

    The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile-brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr-2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr-2WVTa).

  3. Charpy impact tests on martensitic/ferritic steels after irradiation in SINQ target-3

    NASA Astrophysics Data System (ADS)

    Dai, Yong; Marmy, Pierre

    2005-08-01

    Charpy impact tests were performed on martensitic/ferritic (MF) steels T91, F82H, Optifer-V and Optimax-A/-C irradiated in SINQ Target-3 up to 7.5 dpa and 500 appm He in a temperature range of 120-195 °C. Results demonstrate that for all the four kinds of steels, the ductile-to-brittle transition temperature (DBTT) increases with irradiation dose. The difference in the DBTT shifts (ΔDBTT) of the different steels is not significant after irradiation in the SINQ target. The ΔDBTT data from the previous small punch (Δ DBTT SP) and the present Charpy impact (ΔDBTT CVN) tests can be correlated with the expression: Δ DBTT SP = 0.4ΔDBTT CVN. All the ΔDBTT data fall into a linear band when they are plotted versus helium concentration. The results indicate that helium effects on the embrittlement of MF steels are significant, particularly at higher concentrations. It suggests that MF steels may not be very suitable for applications at low temperatures in spallation irradiation environments where helium production is high.

  4. Microstructural analysis of ferritic-martensitic steels irradiated at low temperature in HFIR

    SciTech Connect

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

    1998-09-01

    Disk specimens of ferritic-martensitic steel, HT9 and F82H, irradiated to damage levels of {approximately}3 dpa at irradiation temperatures of either {approximately}90 C or {approximately}250 C have been investigated by using transmission electron microscopy. Before irradiation, tempered HT9 contained only M{sub 23}C{sub 6} carbide. Irradiation at 90 C and 250 C induced a dislocation loop density of 1 {times} 10{sup 22} m{sup {minus}3} and 8 {times} 10{sup 21} m{sup {minus}3}, respectively. in the HT9 irradiated at 250 C, a radiation-induced phase, tentatively identified as {alpha}{prime}, was observed with a number density of less than 1 {times} 10{sup 20} m{sup {minus}3}. On the other hand, the tempered F82H contained M{sub 23}C{sub 6} and a few MC carbides; irradiation at 250 C to 3 dpa caused minor changes in these precipitates and induced a dislocation loop density of 2 {times} 10{sup 22} m{sup {minus}3}. Difference in the radiation-induced phase and the loop microstructure may be related to differences in the post-yield deformation behavior of the two steels.

  5. Irradiation-induced impurity segregation and ductile-to-brittle transition temperature shift in high chromium ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Lu, Z.; Faulkner, R. G.; Flewitt, P. E. J.

    2007-08-01

    A model is presented to predict irradiation-induced impurity segregation and its contribution to the ductile-to-brittle transition temperature (DBTT) shift in high chromium ferritic steels. The hardening contribution (dislocation loops, voids and precipitates) is also considered in this study. The predicted results are compared with the experimental DBTT shifts data for irradiated 9Cr1MoVNb and 12Cr1MoVW steels with different grain sizes.

  6. On the (in)adequacy of the Charpy impact test to monitor irradiation effects of ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Chaouadi, R.

    2007-02-01

    Irradiation embrittlement studies rely very often on Charpy impact data, in particular the ductile-to-brittle transition temperature (DBTT). However, while the DBTT-shift is equivalent to the increase of the fracture toughness transition temperature of ferritic steels, it is not the case for ferritic/martensitic steels. The aim of this study is to critically assess experimental data obtained on a 9%Cr-ferritic/martensitic steel, Eurofer-97, to better understand the underlying mechanisms involved during the fracture process. More specifically, a dedicated analysis using the load diagram approach allows to unambiguously reveal the actual effects of irradiation on physically rather than empirically based parameters. A comparison is made between a ferritic and ferritic/martensitic steel to better identify the possible similarities and differences. Tensile, Charpy impact and fracture toughness tests data are examined in a global approach to assess the actual rather than apparent irradiation effects. The adequacy or inadequacy of the Charpy impact test to monitor irradiation effects is extensively discussed.

  7. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    SciTech Connect

    Marquis, Emmanuelle; Wirth, Brian; Was, Gary

    2016-03-28

    Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

  8. Effect of heat treatment and irradiation temperature on impact properties of Cr-W-V ferritic steels

    NASA Astrophysics Data System (ADS)

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

    Charpy impact tests were conducted on eight normalized-and-tempered ferritic and martensitic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility (FFTF) at 393°C to ≈14 dpa on eight steels with 2.25%, 5%, 9%, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5% and 9% Cr steels, and martensite with ≈25% δ-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy (USE). The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5Cr steel was affected by heat treatment. When the results at 393°C were compared with previous results at 365°C, all but a 5Cr and a 9Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

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

    SciTech Connect

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

    1996-04-01

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

  10. Effects of proton irradiation on nanocluster precipitation in ferritic steel containing fcc alloying additions

    SciTech Connect

    Zhang, Zhongwu; Liu, C T; Wang, Xun-Li; Miller, Michael K; Ma, Dong; Chen, Guang; Williams, J R; Chin, Bryan

    2012-01-01

    Newly-developed precipitate-strengthened ferritic steels with and without pre-existing nanoscale precipitates were irradiated with 4 MeV protons to a dose of ~5 mdpa at 50 C and subsequently examined by nanoindentation and atom probe tomography (APT). Irradiation-enhanced precipitation and coarsening of pre-existing nanoscale precipitates were observed. Copper partitions to the precipitate core along with a segregation of Ni, Al and Mn to the precipitate/matrix interface after both thermal aging and proton irradiation. Proton irradiation induces the precipitation reaction and coarsening of pre-existing nanoscale precipitates, and these results are similar to a thermal aging process. The precipitation and coarsening of nanoscale precipitates are responsible for the changes in hardness. The observation of the radiation-induced softening is essentially due to the coarsening of the pre-existing Cu-rich nanoscale precipitates. The implication of the precipitation on the embrittlement of reactor-pressure-vessel steels after irradiation is discussed.

  11. Microstructural stability of a self-ion irradiated lanthana-bearing nanostructured ferritic steel

    SciTech Connect

    Pasebani, Somayeh; Charit, Indrajit; Burns, Jatuporn; Alsagabi, Sultan; Butt, Darryl P.; Cole, James I.; Price, Lloyd M.; Shao, Lin

    2015-07-01

    Thermally stable nanofeatures with high number density are expected to impart excellent high temperature strength and irradiation stability in nanostructured ferritic steels (NFSs) which have potential applications in advanced nuclear reactors. A lanthana-bearing NFS (14LMT) developed via mechanical alloying and spark plasma sintering was used in this study. The sintered samples were irradiated by Fe2+ ions to 10, 50 and 100 dpa at 30 °C and 500 °C. Microstructural and mechanical characteristics of the irradiated samples were studied using different microscopy techniques and nanoindentation, respectively. Overall morphology and number density of the nanofeatures remained unchanged after irradiation. Average radius of nanofeatures in the irradiated sample (100 dpa at 500 °C) was slightly reduced. A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30 °C and ≥50 dpa. Other microstructural features like grain boundaries and high density of dislocations also provided defect sinks to assist in defect removal.

  12. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    SciTech Connect

    Gelles, D.S.; Shibayama, T.

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  13. Void swelling in high dose ion-irradiated reduced activation ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Wang, Xu; Monterrosa, Anthony M.; Zhang, Feifei; Huang, Hao; Yan, Qingzhi; Jiao, Zhijie; Was, Gary S.; Wang, Lumin

    2015-07-01

    To determine the void swelling resistance of reduced-activation ferritic-martensitic steels CNS I and CNS II at high doses, ion irradiation was performed up to 188 dpa (4.6 × 1017 ion/cm2) at 460 °C using 5 MeV Fe++ ions. Helium was pre-implanted at levels of 10 and 100 appm at room temperature to investigate the role of helium on void swelling. Commercial FM steel T91 was also irradiated in this condition and the swelling results are of included in this paper as a reference. Voids were observed in all conditions. The 9Cr CNS I samples implanted with 10 appm helium exhibited lower swelling than 9Cr T91 irradiated at the same condition. The 12Cr CNS II with 10 and 100 appm helium showed significantly lower swelling than CNS I and T91. The swelling rate for CNS I and CNS II were determined to be 0.02%/dpa and 0.003%/dpa respectively. Increasing the helium content from 10 to 100 appm shortened the incubation region and increased the void density but had no effect on the swelling rates.

  14. Prediction of yield stress and Charpy transition temperature in highly neutron irradiated ferritic steels

    NASA Astrophysics Data System (ADS)

    Windsor, Colin; Cottrell, Geoff; Kemp, Richard

    2010-07-01

    Recent predictions have been made of metallurgical properties of low-activation ferritic/martensitic steels alloys at the high irradiation levels (displacements per atom or dpa) needed for a fusion power plant as based on measurements at low irradiation levels where more data are available. These predictions have been published for the yield stress and for the Charpy ductile to brittle transition temperature shift. The neural network model predictions use training data up to a certain dpa level to predict metallurgical properties above this level. This 'extrapolation' mode of neural networks is explored in some detail. Our studies revealed an increasing accuracy of predictions as the test dpa level is increased for both yield stress and Charpy shift predictions. This result suggests that a model exists for these metallurgical properties as a function of dpa level which becomes more accurate as the available irradiation range in the training data is increased. The explanation suggested is that the metallurgical annealing, which occurs as the irradiation level is increased, simplifies the microstructure and makes prediction more reliable.

  15. High Strain Fatigue Properties of the F82H Ferritic-Martensitic Steel under Proton Irradiation.

    SciTech Connect

    Marmy, P; Oliver, Brian M. )

    2003-05-15

    During the up and down cycles of a fusion reactor, the first wall is exposed concomitantly to a flux of energetic neutrons that generates radiation defects and to a neutron thermal flux that induces thermal stresses. The resulting strains may exceed the elastic limit and induce a plastic deformation in the material. A similar situation occurs in the window of a spallation liquid source target and results in the same type of damage. This particular loading has been simulated in F82H martensitic ferritic steel, using a device allowing a fatigue test to be carried out during irradiation with 590 MeV protons. All fatigue tests were carried out at 300?C, in a strain controlled test at strain levels around 0.8%. Two different signals have been used: a fully symmetrical triangle wave signal (R=-1) and a triangle ramp with 2 min tension holds. The fatigue was investigated under three different conditions: unirradiated , irradiated and post irradiation tested, and finally in beam tested. The main result is that the in beam tested specimens have the lowest life as compared to the post irradiation tested specimens and unirradiated specimens. Hydrogen is suspected to be the main contributor to the observed embrittlement.

  16. High strain fatigue properties of F82H ferritic martensitic steel under proton irradiation

    NASA Astrophysics Data System (ADS)

    Marmy, P.; Oliver, B. M.

    2003-05-01

    During the up and down cycles of a fusion reactor, the first wall is exposed concomitantly to a flux of energetic neutrons that generates radiation defects and to a thermal flux that induces thermal stresses. The resulting strains may exceed the elastic limit and induce plastic deformation in the material. A similar situation occurs in the window of a spallation liquid source target and results in the same type of damage. This particular loading has been simulated in F82H ferritic-martensitic steel, using a device allowing a fatigue test to be carried out during irradiation with 590 MeV protons. All fatigue tests were carried out in a strain controlled test at strain levels around 0.8% and at 300 °C. Two different signals have been used: a fully symmetrical triangle wave signal ( R=-1) and a triangle ramp with 2 min tension holds. The fatigue was investigated under three different conditions: unirradiated, irradiated and post-irradiation tested, and finally in-beam tested. The main result is that the in-beam tested specimens have the lowest life as compared to the post-irradiation tested specimen and unirradiated specimen. Hydrogen is suspected to be the main contributor to the observed embrittlement.

  17. Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Chakraborty, Pritam; Biner, S. Bulent

    2015-10-01

    Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

  18. Irradiation effects on precipitation and its impact on the mechanical properties of reduced-activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tanigawa, H.; Sakasegawa, H.; Hashimoto, N.; Klueh, R. L.; Ando, M.; Sokolov, M. A.

    2007-08-01

    It was previously reported that reduced-activation ferritic/martensitic steels (RAFs) showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573 K up to 5 dpa. The precipitation behavior of the irradiated steels was examined and the presence of irradiation induced precipitation which works as if it was forced to reach the thermal equilibrium state at irradiation temperature 573 K. In this study, transmission electron microscopy was performed on extraction replica specimens to analyze the size distribution of precipitates. It turned out that the hardening level multiplied by the square root of the average block size showed a linear dependence on the extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by different precipitation behavior on block, packet and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable for interpreting this dependence.

  19. Microstructural evolution of ferritic-martensitic steels under heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Topbasi, Cem

    Ferritic-martensitic steels are primary candidate materials for fuel cladding and internal applications in the Sodium Fast Reactor, as well as first-wall and blanket materials in future fusion concepts because of their favorable mechanical properties and resistance to radiation damage. Since microstructure evolution under irradiation is amongst the key issues for these materials in these applications, developing a fundamental understanding of the irradiation-induced microstructure in these alloys is crucial in modeling and designing new alloys with improved properties. The goal of this project was to investigate the evolution of microstructure of two commercial ferritic-martensitic steels, NF616 and HCM12A, under heavy ion irradiation at a broad temperature range. An in situ heavy ion irradiation technique was used to create irradiation damage in the alloy; while it was being examined in a transmission electron microscope. Electron-transparent samples of NF616 and HCM12A were irradiated in situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with 1 MeV Kr ions to ˜10 dpa at temperatures ranging from 20 to 773 K. The microstructure evolution of NF616 and HCM12A was followed in situ by systematically recording micrographs and diffraction patterns as well as capturing videos during irradiation. In these irradiations, there was a period during which no changes are visible in the microstructure. After a threshold dose (˜0.1 dpa between 20 and 573 K, and ˜2.5 dpa at 673 K) black dots started to become visible under the ion beam. These black dots appeared suddenly (from one frame to the next) and are thought to be small defect clusters (2-5 nm in diameter), possibly small dislocation loops with Burgers vectors of either ½ or . The overall density of these defect clusters increased with dose and saturated around 6 dpa. At saturation, a steady-state is reached in which defects are eliminated and created at the same rates so that the

  20. Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

    SciTech Connect

    Ahmad Alsabbagh; Apu Sarkar; Brandon Miller; Jatuporn Burns; Leah Squires; Douglas Porter; James I. Cole; K. L. Murty

    2014-10-01

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

  1. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    SciTech Connect

    Jones, R.H.; Gelles, D.S.

    1993-08-01

    Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry.

  2. Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

    SciTech Connect

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

    1996-10-01

    Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

  3. Effect of neutron irradiation at low temperature on the embrittlement of the reduced-activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Rybin, V. V.; Kursevich, I. P.; Lapin, A. N.

    1998-10-01

    Effects of neutron irradiation to fluence of 2.0 × 10 24 n/m 2 ( E > 0.5 MeV) in temperature range 70-300°C on mechanical properties and structure of the experimental reduced-activation ferritic 0.1%C-(2.5-12)%Cr-(1-2)%W-(0.2-0.7)%V alloys were investigated. The steels were studied in different initial structural conditions obtained by changing the modes of heat treatments. Effect of neutron irradiation estimated by a shift in ductile-brittle transition temperature (ΔDBTT) and reduction of upper shelf energy (ΔUSE) highly depends on both irradiation condition and steel chemical composition and structure. For the steel with optimum chemical composition (9Cr-1.5WV) after irradiation to 2 × 10 24 n/m 2 ( E ⩾ 0.5 MeV) at 280°C the ΔDBTT does not exceed 25°C. The shift in DBTT increased from 35°C to 110°C for the 8Cr-1.5WV steel at a decrease in irradiation temperature from 300°C to 70°C. The CCT diagrams are presented for several reduced-activated steels.

  4. Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation

    SciTech Connect

    Klueh, R.L.; Vitek, J.M.; Corwin, W.R.; Alexander, D.J.

    1987-01-01

    Miniature Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 1 and 2% Ni were irradiated in the High-Flux Isotope Reactor (HFIR) at 50/sup 0/C to displacement damage levels of up to 9 dpa. Nickel was added to study the effect of transmutation helium. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT). The 9Cr-1MoVNb steels, with and without nickel, showed a larger shift than the 12Cr-1MoVW steels, with and without nickel. The results indicated that helium also increased the DBTT. The same steels were previously irradiated at higher temperatures. From the present and past tests, the effect of irradiation temperature on the DBTT behavior can be evaluated. For the 9Cr-1MoVNb steel, there is a continuous decrease in the magnitude of the DBTT increase up to an irradiation temperature of about 400/sup 0/C, after which the shift drops rapidly to zero at about 450/sup 0/C. The DBTT of the 12Cr-1MoVW steel shows a maximum increase at an irradiation temperature of about 400/sup 0/C and less of an increase at either higher or lower irradiation temperatures.

  5. Tensile and charpy impact properties of irradiated reduced-activation ferritic steels

    SciTech Connect

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

    1996-10-01

    Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365{degrees}C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365{degrees}C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After {approx}7 dpa, the strength of the steels increased and then remained relatively unchanged through 26-29 dpa (i.e., the strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels were the most irradiation resistant.

  6. Evolution of the mechanical properties and microstructure of ferritic-martensitic steels irradiated in the BOR-60 reactor

    NASA Astrophysics Data System (ADS)

    Shamardin, V. K.; Golovanov, V. N.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.; Ostrovsky, Z. E.; Goncharenko, Yu. D.

    2002-12-01

    The effect of neutron irradiation on mechanical properties of low-activation ferritic-martensitic (FM) steels 0.1C-9Cr-1W, V, Ta, B and 0.1C-12Cr-2W, V, Ti, B is studied under tension at temperatures of 330-540 °C and doses of 50 dpa. Steel 0.1C-13Cr-Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330-340 °C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 °C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model ɛ¯/ σ¯=B 0+D Ṡ, when B0 and D have the values typical for steels of FM type.

  7. Charpy Impact Properties of Reduced-Activation Ferritic/Martensitic Steels Irradiated in HFIR up to 20 dpa

    SciTech Connect

    Tanigawa, H.; Shiba, K.; Sokolov, M.A.; Klueh, R.L.

    2003-07-15

    The effects of irradiation up to 20 dpa on the Charpy impact properties of reduced-activation ferritic/martensitic steels (RAFs) were investigated. The ductile-brittle transition temperature (DBTT) of F82H-IEA shifted up to around 323K. TIG weldments of F82H showed a fairly small variation on their impact properties. A finer prior austenite grain size in F82H-IEA after a different heat treatment resulted in a 20K lower DBTT compared to F82H-IEA after the standard heat treatment, and that effect was maintained even after irradiation. Helium effects were investigated utilizing Ni-doped F82H, but no obvious evidence of helium effects was obtained. ORNL9Cr-2WVTa and JLF-1 steels showed smaller DBTT shifts compared to F82H-IEA.

  8. Tensile and Charpy impact properties of irradiated reduced-activation ferritic steels

    SciTech Connect

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

    1996-10-01

    Tensile tests were conducted on 8 reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on steels irradiated to 26-29 dpa. Irradiation was in Fast Flux Test Facility at 365 C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15- 17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20,000 h at 365 C. Thermal aging had little effect on tensile properties or ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in upper-shelf energy (USE). After 7 dpa, strength increased (hardened) and then remained relatively unchanged through 26-29 dpa (ie, strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness (increased DBTT, decreased USE) remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels had the most irradiation resistance.

  9. MICROSTRUCTURAL EXAMINATION OF LOW ACTIVATION FERRITIC STEELS FOLLOWING IRRADIATION IN ORR

    SciTech Connect

    Gelles, David S.

    2002-09-01

    Microstructural examinations are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 degrees C to approximately 10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5 percent and W to 1.0 percent. Results include compositional changes in precipitates and microstructural changes as a function of composition and irradiation temperature. It is concluded that temperatures in ORR are on the order of 50 degrees C higher than anticipated.

  10. Articles comprising ferritic stainless steels

    SciTech Connect

    Rakowski, James M.

    2016-06-28

    An article of manufacture comprises a ferritic stainless steel that includes a near-surface region depleted of silicon relative to a remainder of the ferritic stainless steel. The article has a reduced tendency to form an electrically resistive silica layer including silicon derived from the steel when the article is subjected to high temperature oxidizing conditions. The ferritic stainless steel is selected from the group comprising AISI Type 430 stainless steel, AISI Type 439 stainless steel, AISI Type 441 stainless steel, AISI Type 444 stainless steel, and E-BRITE.RTM. alloy, also known as UNS 44627 stainless steel. In certain embodiments, the article of manufacture is a fuel cell interconnect for a solid oxide fuel cell.

  11. Influence of helium on deuterium retention in reduced activation ferritic martensitic steel (F82H) under simultaneous deuterium and helium irradiation

    NASA Astrophysics Data System (ADS)

    Yakushiji, K.; Lee, H. T.; Oya, M.; Hamaji, Y.; Ibano, K.; Ueda, Y.

    2016-02-01

    Deuterium and helium retention in Japanese reduced activation ferritic martensitic (RAFM) steel (F82H) under simultaneous D-He irradiation at 500, 625, 750, and 818 K was studied. This study aims to clarify tritium retention behavior in RAFM steels to assess their use as plasma facing materials. The irradiation fluence was kept constant at 1 × 1024 D m-2. Four He desorption peaks were observed with He retention greatest at 625 K. At T > 625 K a monotonic decrease in He retention was observed. At all temperatures a systematic reduction in D retention was observed for the simultaneous D-He case in comparison to D-only case. This suggests that He implanted at the near surface in RAFM steels may reduce the inward penetration of tritium in RAFM steels that would result in lower tritium inventory for a given fluence.

  12. Small punch tests on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3

    NASA Astrophysics Data System (ADS)

    Jia, X.; Dai, Y.

    2003-12-01

    Small punch (SP) tests were conducted in a temperature range from -190 to 80 °C on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3 up to 9.4 dpa in a irradiation temperature range of 90-275 °C. Results demonstrate: (a) the irradiation hardening deduced from SP tests is reasonably consistent with the results obtained by tensile tests; (b) with increasing irradiation dose, the SP yield load increases at all test temperatures, while the displacement at the maximum load and the total displacement at failure decrease; (c) the ductile-to-brittle transition temperature (DBTT SP) increases with increasing irradiation dose, and does so more quickly at irradiation doses above ˜6-7 dpa; in addition, the ΔDBTT SP increases linearly with helium content.

  13. Synthesis of atom probe experiments on irradiation-induced solute segregation in French ferritic pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Auger, P.; Pareige, P.; Welzel, S.; Van Duysen, J.-C.

    2000-08-01

    Microstructural changes due to neutron irradiation cause an evolution of the mechanical properties of reactor pressure vessels (RPV) steels. This paper aims at identifying and characterising the microstructural changes which have been found to be responsible in part for the observed embrittlement. This intensive work relies principally on an atom probe (AP) study of a low Cu-level French RPV steel (Chooz A). This material has been irradiated in in-service conditions for 0-16 years in the frame of the surveillance program. Under this aging condition, solute clustering occurs (Cu, Ni, Mn, Si, P, …). In order to identify the role of copper, experiments were also carried out on Fe-Cu model alloys submitted to different types of irradiations (neutron, electron, ion). Cu-cluster nucleation appears to be directly related to the presence of displacement cascades during neutron (ion) irradiation. The operating basic physical process is not clearly identified yet. A recovery of the mechanical properties of the irradiated material can be achieved by annealing treatments (20 h at 450°C in the case of the RPV steel under study, following microhardness measurements). It has been shown that the corresponding microstructural evolution was a rapid dissolution of the high number density of irradiation-induced solute clusters and the precipitation of a very low number density of Cu-rich particles.

  14. SANS and TEM of ferritic-martensitic steel T91 irradiated in FFTF up to 184 dpa at 413 °C

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Anderoglu, O.; Dickerson, R.; Hartl, M.; Dickerson, P.; Aguiar, J. A.; Hosemann, P.; Toloczko, M. B.; Maloy, S. A.

    2013-09-01

    Ferritic-martensitic steel T91 was previously irradiated in the Materials Open Test Assembly (MOTA) program of the Fast Flux Test Reactor Facility (FFTF) at 413 °C up to 184 dpa. The microstructure was analyzed by small angle neutron scattering (SANS) and transmission electron microscopy (TEM). Both SANS and TEM revealed a large fraction of voids with an average size of 29-32 nm leading to a calculated void swelling of 1.2-1.6% based on the volume fraction of the voids in the sample. SANS gave no indication of second phase particles having formed under irradiation in the material. Using TEM, one zone was found where a few G-phase particles were analyzed. Quantities were however too low to state reliable particle densities. No alpha prime (α') or Laves phase were observed in any of the investigated zones.

  15. The effect of low dose irradiation on the impact fracture energy and tensile properties of pure iron and two ferritic martensitic steels

    NASA Astrophysics Data System (ADS)

    Belianov, I.; Marmy, P.

    1998-10-01

    Two batches of subsize V-notched impact bend specimens and subsize tensile specimens have been irradiated in the Saphir test reactor of the Paul Scherrer Institute (PSI). The first batch of specimen has been irradiated at 250°C to a dose of 2.65 × 10 19 n/cm 2 (0.042 dpa) and the second batch has been irradiated at 400°C to a dose of 8.12 × 10 19 n/cm 2 (0.13 dpa). Three different materials in three different microstructures were irradiated: pure iron and two ferritic steels, the alloy MANET 2 and a low activation composition CETA. The results of the impact tests and of the corresponding tensile tests are presented. Despite the very low neutron dose, a significant shift of the ductile to brittle transition temperature (DBTT) is observed. The influence of the test temperature on the impact energy is discussed for the irradiated and unirradiated conditions, with special emphasis on the microstructure.

  16. Effect of ferrite on cast stainless steels

    SciTech Connect

    Nadezhdin, A.; Cooper, K. ); Timbers, G. . Kraft Pulp Division)

    1994-09-01

    Premature failure of stainless steel castings in bleach washing service is attributed to poor casting quality high porosity and to a high ferrite content, which makes the castings susceptible to corrosion by hot acid chloride solutions. A survey of the chemical compositions and ferrite contents of corrosion-resistant castings in bleach plants at three pulp mills found high [delta]-ferrite levels in the austenitic matrix due to the improper balance between austenite and ferrite stabilizers.

  17. Ferritic steel melt and FLiBe/steel experiment : melting ferritic steel.

    SciTech Connect

    Troncosa, Kenneth P.; Smith, Brandon M.; Tanaka, Tina Joan

    2004-11-01

    In preparation for developing a Z-pinch IFE power plant, the interaction of ferritic steel with the coolant, FLiBe, must be explored. Sandia National Laboratories Fusion Technology Department was asked to drop molten ferritic steel and FLiBe in a vacuum system and determine the gas byproducts and ability to recycle the steel. We tried various methods of resistive heating of ferritic steel using available power supplies and easily obtained heaters. Although we could melt the steel, we could not cause a drop to fall. This report describes the various experiments that were performed and includes some suggestions and materials needed to be successful. Although the steel was easily melted, it was not possible to drip the molten steel into a FLiBe pool Levitation melting of the drop is likely to be more successful.

  18. Effect of heat treatment and irradiation temperature on mechanical properties and structure of reduced-activation Cr-W-V steels of bainitic, martensitic, and martensitic-ferritic classes

    NASA Astrophysics Data System (ADS)

    Gorynin, I. V.; Rybin, V. V.; Kursevich, I. P.; Lapin, A. N.; Nesterova, E. V.; Klepikov, E. Yu

    2000-12-01

    Effects of molybdenum replacement by tungsten in steels of the bainitic, martensitic, and martensitic-ferritic classes containing 2.5%, 8% and 11% Cr, respectively, were investigated. The phase composition and structure of the bainitic steels were varied by changing the cooling rates from the austenitization temperature (from values typical for normalization up to V=3.3 × 10-2°C/s) and then tempering. The steels were irradiated to a fluence of 4×1023 n/m2 (⩾0.5 MeV) at 270°C and to fluences of 1.3×1023 and 1.2×1024 n/m2 (⩾0.5 MeV) at 70°C. The 2.5Cr-1.4WV and 8Cr-1.5WV steels have shown lower values of the shifts in ductile-brittle transition temperature (DBTT) under irradiation in comparison with corresponding Cr-Mo steels. Radiation embrittlement at elevated irradiation temperature was lowest in bainitic 2.5Cr-1.4WV steel and martensitic-ferritic 11Cr-1.5WV steel. The positive effect of molybdenum replacement by tungsten at irradiation temperature ∼300°C is reversed at Tirr=70∘C.

  19. R&D of low activation ferritic steels for fusion in japanese universities*1

    NASA Astrophysics Data System (ADS)

    Kohyama, Akira; Kohno, Yutaka; Asakura, Kentaro; Kayano, Hideo

    1994-09-01

    Following the brief review of the R&D of low activation ferritic steels in Japanese universities, the status of 9Cr-2W type ferritic steels development is presented. The main emphasis is on mechanical property changes by fast neutron irradiation in FFTF. Bend test, tensile test, CVN test and in-reactor creep results are provided including some data about low activation ferritic steels with Cr variation from 2.25 to 12%. The 9Cr-2W ferritic steel, denoted as JLF-1, showed excellent mechanical properties under fast neutron irradiation as high as 60 dpa. As potential materials for DEMO and beyond, innovative oxide dispersion strengthened (ODS) quasi-amorphous low activation ferritic steels are introduced. The baseline properties, microstructural evolution under ion irradiation and the recent progress of new processes are provided.

  20. Welding irradiated stainless steel

    SciTech Connect

    Kanne, W.R. Jr.; Chandler, G.T.; Nelson, D.Z.; Franco-Ferreira, E.A.

    1993-12-31

    Conventional welding processes produced severe underbead cracking in irradiated stainless steel containing 1 to 33 appm helium from n,a reactions. A shallow penetration overlay technique was successfully demonstrated for welding irradiated stainless steel. The technique was applied to irradiated 304 stainless steel that contained 10 appm helium. Surface cracking, present in conventional welds made on the same steel at the same and lower helium concentrations, was eliminated. Underbead cracking was minimal compared to conventional welding methods. However, cracking in the irradiated material was greater than in tritium charged and aged material at the same helium concentrations. The overlay technique provides a potential method for repair or modification of irradiated reactor materials.

  1. Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

    SciTech Connect

    Hirose, Takanori; Sokolov, Mikhail A; Ando, M.; Tanigawa, H.; Shiba, K.; Stoller, Roger E; Odette, G.R.

    2013-11-01

    This work investigates irradiation response in the joints of F82H employed for a fusion breeding blanket. The joints, which were prepared using welding and diffusion welding, were irradiated up to 6 dpa in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Post-irradiation tests revealed hardening in weldment (WM) and base metal (BM) greater than 300 MPa. However, the heat affected zones (HAZ) exhibit about half that of WM and BM. Therefore, neutron irradiation decreased the strength of the HAZ, leaving it in danger of local deformation in this region. Further the hardening in WM made with an electron beam was larger than that in WM made with tungsten inert gas welding. However the mechanical properties of the diffusion-welded joint were very similar to those of BM even after the irradiation.

  2. Use of double and triple-ion irradiation to study the influence of high levels of helium and hydrogen on void swelling of 8-12% Cr ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Kupriiyanova, Y. E.; Bryk, V. V.; Borodin, O. V.; Kalchenko, A. S.; Voyevodin, V. N.; Tolstolutskaya, G. D.; Garner, F. A.

    2016-01-01

    In accelerator-driven spallation (ADS) devices, some of the structural materials will be exposed to intense fluxes of very high energy protons and neutrons, producing not only displacement damage, but very high levels of helium and hydrogen. Unlike fission flux-spectra where most helium and hydrogen are generated by transmutation in nickel and only secondarily in iron or chromium, gas production in ADS flux-spectra are rather insensitive to alloy composition, such that Fe-Cr base ferritic alloys also generate very large gas levels. While ferritic alloys are known to swell less than austenitic alloys in fission spectra, there is a concern that high gas levels in fusion and especially ADS facilities may strongly accelerate void swelling in ferritic alloys. In this study of void swelling in response to helium and hydrogen generation, irradiation was conducted on three ferritic-martensitic steels using the Electrostatic Accelerator with External Injector (ESUVI) facility that can easily produce any combination of helium to dpa and/or hydrogen to dpa ratios. Irradiation was conducted under single, dual and triple beam modes using 1.8 MeV Cr+3, 40 keV He+, and 20 keV H+. In the first part of this study we investigated the response of dual-phase EP-450 to variations in He/dpa and H/dpa ratio, focusing first on dual ion studies and then triple ion studies, showing that there is a diminishing influence on swelling with increasing total gas content. In the second part we investigated the relative response of three alloys spanning a range of starting microstructure and composition. In addition to observing various synergisms between He and H, the most important conclusion was that the tempered martensite phase, known to lag behind the ferrite phase in swelling in the absence of gases, loses much of its resistance to void nucleation when irradiated at large gas/dpa levels.

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

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1998-03-01

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

  4. Martensitic/ferritic steels as container materials for liquid mercury target of ESS

    SciTech Connect

    Dai, Y.

    1996-06-01

    In the previous report, the suitability of steels as the ESS liquid mercury target container material was discussed on the basis of the existing database on conventional austenitic and martensitic/ferritic steels, especially on their representatives, solution annealed 316 stainless steel (SA 316) and Sandvik HT-9 martensitic steel (HT-9). Compared to solution annealed austenitic stainless steels, martensitic/ferritic steels have superior properties in terms of strength, thermal conductivity, thermal expansion, mercury corrosion resistance, void swelling and irradiation creep resistance. The main limitation for conventional martensitic/ferritic steels (CMFS) is embrittlement after low temperature ({le}380{degrees}C) irradiation. The ductile-brittle transition temperature (DBTT) can increase as much as 250 to 300{degrees}C and the upper-shelf energy (USE), at the same time, reduce more than 50%. This makes the application temperature range of CMFS is likely between 300{degrees}C to 500{degrees}C. For the present target design concept, the temperature at the container will be likely controlled in a temperature range between 180{degrees}C to 330{degrees}C. Hence, CMFS seem to be difficult to apply. However, solution annealed austenitic stainless steels are also difficult to apply as the maximum stress level at the container will be higher than the design stress. The solution to the problem is very likely to use advanced low-activation martensitic/ferritic steels (LAMS) developed by the fusion materials community though the present database on the materials is still very limited.

  5. Residual ferrite formation in 12CrODS steels

    NASA Astrophysics Data System (ADS)

    Ukai, S.; Kudo, Y.; Wu, X.; Oono, N.; Hayashi, S.; Ohtsuka, S.; Kaito, T.

    2014-12-01

    Increasing Cr content from 9 to 12 mass% leads to superior corrosion and high-temperature oxidation resistances, and usually changes microstructure from martensite to a ferrite. To make transformable martensitic type of 12CrODS steels that have superior processing capability by using α/γ phase transformation, alloy design was conducted through varying nickel content. The structure of 12CrODS steels was successfully modified from full ferrite to a transformable martensite-base matrix containing ferrite. This ferrite consists of both equilibrium ferrite and a metastable residual ferrite. It was shown that the fraction of the equilibrium ferrite is predictable by computed phase diagram and formation of the residual ferrite was successfully evaluated through pinning of α/γ interfacial boundaries by oxide particles.

  6. Cast Stainless Steel Ferrite and Grain Structure

    SciTech Connect

    Ruud, Clayton O.; Ramuhalli, Pradeep; Meyer, Ryan M.; Mathews, Royce; Diaz, Aaron A.; Anderson, Michael T.

    2012-09-01

    In-service inspection requirements dictate that piping welds in the primary pressure boundary of light-water reactors be subject to a volumetric examination based on the rules contained within the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI. The purpose of the inspection is the reliable detection and accurate sizing of service-induced degradation and/or material flaws introduced during fabrication. The volumetric inspection is usually carried out using ultrasonic testing (UT) methods. However, the varied metallurgical macrostructures and microstructures of cast austenitic stainless steel piping and fittings, including statically cast stainless steel and centrifugally cast stainless steel (CCSS), introduce significant variations in the propagation and attenuation of ultrasonic energy. These variations complicate interpretation of the UT responses and may compromise the reliability of UT inspection. A review of the literature indicated that a correlation may exist between the microstructure and the delta ferrite content of the casting alloy. This paper discusses the results of a recent study where the goal was to determine if a correlation existed between measured and/or calculated ferrite content and grain structure in CCSS pipe.

  7. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

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

  8. Investigations of low-temperature neutron embrittlement of ferritic steels

    SciTech Connect

    Farrell, K.; Mahmood, S.T.; Stoller, R.E.; Mansur, L.K.

    1992-12-31

    Investigations were made into reasons for accelerated embrittlement of surveillance specimens of ferritic steels irradiated at 50C at the High Flux Isotope Reactor (HFIR) pressure vessel. Major suspects for the precocious embrittlement were a highly thermalized neutron spectrum,a low displacement rate, and the impurities boron and copper. None of these were found guilty. A dosimetry measurement shows that the spectrum at a major surveillance site is not thermalized. A new model of matrix hardening due to point defect clusters indicates little effect of displacement rate at low irradiation temperature. Boron levels are measured at 1 wt ppM or less, inadequate for embrittlement. Copper at 0.3 wt % and nickel at 0.7 wt % are shown to promote radiation strengthening in iron binary alloys irradiated at 50 to 60C, but no dependence on copper and nickel was found in steels with 0.05 to 0.22% Cu and 0.07 to 3.3% Ni. It is argued that copper impurity is not responsible for the accelerated embrittlement of the HFIR surveillance specimens. The dosimetry experiment has revealed the possibility that the fast fluence for the surveillance specimens may be underestimated because the stainless steel monitors in the surveillance packages do not record an unexpected component of neutrons in the spectrum at energies just below their measurement thresholds of 2 to 3 MeV.

  9. Must we use ferritic steel in TBM?

    SciTech Connect

    Salavy, Jean-Francois; Boccaccini, Lorenzo V.; Chaudhuri, Paritosh; Cho, Seungyon; Enoeda, Mikio; Giancarli, Luciano; Kurtz, Richard J.; Luo, Tian Y.; Rao, K. Bhanu Sankara; Wong, Clement

    2010-12-13

    Mock-ups of DEMO breeding blankets, called Test Blanket Modules (TBMs), inserted and tested in ITER in dedicated equatorial ports directly facing the plasma, are expected to provide the first experimental answers on the necessary performance of the corresponding DEMO breeding blankets. Several DEMO breeding blanket designs have been studied and assessed in the last 20 years. At present, after considering various coolant and breeder combinations, all the TBM concepts proposed by the seven ITER Parties use Reduced-Activation Ferritic/Martensitic (RAFM) steel as the structural material. In order to perform valuable tests in ITER, the TBMs are expected to use the same structural material as corresponding DEMO blankets. However, due to the fact that this family of steels is ferromagnetic, their presence in the ITER vacuum vessel will create perturbations of the ITER magnetic fields that could reduce the quality of the plasma confinement during H-mode. As a consequence, a legitimate question has been raised on the necessity of using RAFM steel for TBMs structural material in ITER. By giving a short description of the main TBM testing objectives in ITER and assessing the consequences of not using such a material, this paper gives a comprehensive answer to this question. According to the working group author of the study, the use of RAFM steel as structural material for TBM is judged mandatory.

  10. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    SciTech Connect

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

  11. Controlled ferrite content improves weldability of corrosion-resistant steel

    NASA Technical Reports Server (NTRS)

    Malin, C. O.

    1967-01-01

    Corrosion-resistant steel that adds restrictions on chemical composition to ensure sufficient ferrite content decreases the tendency of CRES to develop cracks during welding. The equations restricting composition are based on the Schaeffler constitution diagram.

  12. Performance of ferritic stainless steels for automobile muffler corrosion

    SciTech Connect

    Tarutani, Y.; Hashizume, T.

    1995-11-01

    Corrosion behavior of ferritic stainless steels was studied in artificial exhaust gas condensates containing corrosive ions such as Cl{sup {minus}} and SO{sub 3}{sup 2{minus}}. Continuous immersion tests in flasks and Dip and Dry tests by using the alternate corrosion tester with a heating system clarified the effects of chromium and molybdenum additions on the corrosion resistance of a ferritic stainless steel in the artificial exhaust gas condensates. Effects of surface oxidation on the corrosion behavior were investigated in a temperature range of 573K to 673K. Oxidation of 673K reduced the corrosion resistance of the ferritic stainless steels in the artificial environment of the automobile muffler. Particulate matter deposited on the muffler inner shell from the automobile exhaust gas was also examined. Deposited particulate matter increased the corrosion rate of the ferritic stainless steel. Finally, the authors also investigated the corrosion of the automobile mufflers made of Type 436L ferritic stainless steel with 18% chromium-1.2% molybdenum after 24 months, in Japan. The sets of results clarified that Type 436L ferritic stainless steel as the material for the automobile muffler exhibited acceptable corrosion resistance.

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

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    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.

  14. Tritium retention in reduced-activation ferritic/martensitic steels

    SciTech Connect

    Hatano, Y.; Abe, S.; Matsuyama, M.; Alimov, V.K.; Spitsyn, A.V.; Bobyr, N.P.; Cherkez, D.I.; Khripunov, B.I.; Golubeva, A.V.; Ogorodnikova, O.V.; Klimov, N.S.; Chernov, V.M.; Oyaidzu, M.; Yamanishi, T.

    2015-03-15

    Reduced-activation ferritic/martensitic (RAFM) steels are structural material candidates for breeding blankets of future fusion reactors. Therefore, tritium (T) retention in RAFM steels is an important problem in assessing the T inventory of blankets. In this study, specimens of RAFM steels were subjected to irradiation of 20 MeV W ions to 0.54 displacements per atom (dpa), exposure to high flux D plasmas at 400 and 600 K and that to pulsed heat loads. The specimens thus prepared were exposed to DT gas at 473 K. Despite severe modification in the surface morphology, heat loads had negligible effects on T retention. Significant increase in T retention at the surface and/or subsurface was observed after D plasma exposure. However, T trapped at the surface/subsurface layer was easily removed by maintaining the specimens in the air at about 300 K. Displacement damage led to increase in T retention in the bulk due to the trapping effects of defects, and T trapped was stable at 300 K. It was therefore concluded that displacement damages had the largest influence on T retention under the present conditions.

  15. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  16. 77 FR 60478 - Control of Ferrite Content in Stainless Steel Weld Metal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-03

    ... COMMISSION Control of Ferrite Content in Stainless Steel Weld Metal AGENCY: Nuclear Regulatory Commission... Ferrite Content in Stainless Steel Weld Metal.'' This guide describes a method that the NRC staff considers acceptable for controlling ferrite content in stainless steel weld metal. Revision 4 updates...

  17. Current status and recent research achievements in ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Tavassoli, A.-A. F.; Diegele, E.; Lindau, R.; Luzginova, N.; Tanigawa, H.

    2014-12-01

    When the austenitic stainless steel 316L(N) was selected for ITER, it was well known that it would not be suitable for DEMO and fusion reactors due to its irradiation swelling at high doses. A parallel programme to ITER collaboration already had been put in place, under an IEA fusion materials implementing agreement for the development of a low activation ferritic/martensitic steel, known for their excellent high dose irradiation swelling resistance. After extensive screening tests on different compositions of Fe-Cr alloys, the chromium range was narrowed to 7-9% and the first RAFM was industrially produced in Japan (F82H: Fe-8%Cr-2%W-TaV). All IEA partners tested this steel and contributed to its maturity. In parallel several other RAFM steels were produced in other countries. From those experiences and also for improving neutron efficiency and corrosion resistance, European Union opted for a higher chromium lower tungsten grade, Fe-9%Cr-1%W-TaV steel (Eurofer), and in 1997 ordered the first industrial heats. Other industrial heats have been produced since and characterised in different states, including irradiated up to 80 dpa. China, India, Russia, Korea and US have also produced their grades of RAFM steels, contributing to overall maturity of these steels. This paper reviews the work done on RAFM steels by the fusion materials community over the past 30 years, in particular on the Eurofer steel and its design code qualification for RCC-MRx.

  18. Subcascade formation ratio in neutron-irradiated stainless steels

    NASA Astrophysics Data System (ADS)

    Yoshiie, T.; Satoh, Y.; Huang, S. S.; Horiki, M.; Sato, K.; Xu, Q.

    2016-01-01

    High-energy-particle irradiation in metals produces cascade damage. If the particle energy is high enough, a cascade is divided into subcascades. In each subcascade, a vacancy rich area is surrounded by an interstitial area. Vacancy clusters are expected to form directly in the vacancy rich area. In this study, the vacancy cluster formation ratio in subcascades was estimated by positron annihilation lifetime spectroscopy and transmission electron microscopy in commercial stainless steels and their model alloys. The vacancy cluster formation ratio was 1.7×10-3 and 9.1×10-5 in austenitic stainless steel and ferritic/martensitic stainless steel, respectively

  19. Low-chromium reduced-activation ferritic steels for fusion

    SciTech Connect

    Klueh, R.L.; Alexander, D.J.; Kenik, E.A.

    1996-04-01

    Development of reduced-activation ferritic steels has concentrated on high-chromium (8-10 wt% Cr) steels. However, there are advantages for a low-chromium steel, and initial ORNL studies on reduced-activation steels were on compositions with 2.25 to 12% Cr. Those studies showed an Fe-2.25Cr-2W-0.25V-0.1C (2 1/4Cr-2WV) steel to have the highest strenglth of the steels studied. Although this steel had the best strength, Charpy impact properties were inferior to those of an Fe-9Cr-2W-0.25V-0.07Ta-0.1C (9Cr-2WVTa) and an Fe-2.25Cr-2W-0.1C (2 1/4Cr-2W) steel. Therefore, further development of the low-chromium Cr-W steels was required. These results indicate that it is possible to develop low-chromium reduced-activation ferritic steels that have tensile and impact properties as good or better than those of high-chromium (7-9% Cr) steels. Further improvement of properties should be possible by optimizing the composition.

  20. Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

    2013-01-01

    Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3 T M ( T M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

  1. Recent improvements in size effects correlations for DBTT and upper shelf energy of ferritic steels

    SciTech Connect

    Kumar, A.S.; Louden, B.S. ); Garner, F.A.; Hamilton, M.L. )

    1992-01-01

    Currently available correlations for the effects of specimen size on the USE were developed for relatively ductile steels and will not serve as well when the steels become embrittled. Size effects correlations were developed recently for the impact properties of less ductile HT9 to be applied to other initially more ductile steels as they lose their ductility during irradiation. These new correlations successfully predict the ductile brittle transition temperature (DBTT) and the upper shelf energy (USE) of full size Charpy specimens based on subsize specimen data. The new DBTT and the USE correlations were tested against published experimental data on other ferritic steels and shown to perform successfully at lower USE particularly when both precracked and notched only specimens were employed.

  2. A comparative assessment of the fracture toughness behavior of ferritic-martensitic steels and nanostructured ferritic alloys

    NASA Astrophysics Data System (ADS)

    Byun, Thak Sang; Hoelzer, David T.; Kim, Jeoung Han; Maloy, Stuart A.

    2017-02-01

    The Fe-Cr alloys with ultrafine microstructures are primary candidate materials for advanced nuclear reactor components because of their excellent high temperature strength and high resistance to radiation-induced damage such as embrittlement and swelling. Mainly two types of Fe-Cr alloys have been developed for the high temperature reactor applications: the quenched and tempered ferritic-martensitic (FM) steels hardened primarily by ultrafine laths and carbonitrides and the powder metallurgy-based nanostructured ferritic alloys (NFAs) by nanograin structure and nanoclusters. This study aims at elucidating the differences and similarities in the temperature and strength dependences of fracture toughness in the Fe-Cr alloys to provide a comparative assessment of their high-temperature structural performance. The KJQ versus yield stress plots confirmed that the fracture toughness was inversely proportional to yield strength. It was found, however, that the toughness data for some NFAs were outside the band of the integrated dataset at given strength level, which indicates either a significant improvement or deterioration in mechanical properties due to fundamental changes in deformation and fracture mechanisms. When compared to the behavior of NFAs, the FM steels have shown much less strength dependence and formed narrow fracture toughness data bands at a significantly lower strength region. It appeared that at high temperatures ≥600 °C the NFAs cannot retain the nanostructure advantage of high strength and high toughness either by high-temperature embrittlement or by excessive loss of strength. Irradiation studies have revealed, however, that the NFAs have much stronger radiation resistance than tempered martensitic steels, such as lower radiation-induced swelling, finer helium bubble formation, lower irradiation creep rate and reduced low temperature embrittlement.

  3. A comparative assessment of the fracture toughness behavior of ferritic-martensitic steels and nanostructured ferritic alloys

    DOE PAGES

    Byun, Thak Sang; Hoelzer, David T.; Kim, Jeoung Han; ...

    2016-12-07

    The Fe-Cr alloys with ultrafine microstructures are primary candidate materials for advanced nuclear reactor components because of their excellent high temperature strength and high resistance to radiation-induced damage such as embrittlement and swelling. Mainly two types of Fe-Cr alloys have been developed for the high temperature reactor applications: the quenched and tempered ferritic-martensitic (FM) steels hardened primarily by ultrafine laths and carbonitrides and the powder metallurgy-based nanostructured ferritic alloys (NFAs) by nanograin structure and nanoclusters. This paper aims at elucidating the differences and similarities in the temperature and strength dependences of fracture toughness in the Fe-Cr alloys to provide amore » comparative assessment of their high-temperature structural performance. The KJQ versus yield stress plots confirmed that the fracture toughness was inversely proportional to yield strength. It was found, however, that the toughness data for some NFAs were outside the band of the integrated dataset at given strength level, which indicates either a significant improvement or deterioration in mechanical properties due to fundamental changes in deformation and fracture mechanisms. When compared to the behavior of NFAs, the FM steels have shown much less strength dependence and formed narrow fracture toughness data bands at a significantly lower strength region. It appeared that at high temperatures ≥600 °C the NFAs cannot retain the nanostructure advantage of high strength and high toughness either by high-temperature embrittlement or by excessive loss of strength. Finally, irradiation studies have revealed, however, that the NFAs have much stronger radiation resistance than tempered martensitic steels, such as lower radiation-induced swelling, finer helium bubble formation, lower irradiation creep rate and reduced low temperature embrittlement.« less

  4. A comparative assessment of the fracture toughness behavior of ferritic-martensitic steels and nanostructured ferritic alloys

    SciTech Connect

    Byun, Thak Sang; Hoelzer, David T.; Kim, Jeoung Han; Maloy, Stuart A.

    2016-12-07

    The Fe-Cr alloys with ultrafine microstructures are primary candidate materials for advanced nuclear reactor components because of their excellent high temperature strength and high resistance to radiation-induced damage such as embrittlement and swelling. Mainly two types of Fe-Cr alloys have been developed for the high temperature reactor applications: the quenched and tempered ferritic-martensitic (FM) steels hardened primarily by ultrafine laths and carbonitrides and the powder metallurgy-based nanostructured ferritic alloys (NFAs) by nanograin structure and nanoclusters. This paper aims at elucidating the differences and similarities in the temperature and strength dependences of fracture toughness in the Fe-Cr alloys to provide a comparative assessment of their high-temperature structural performance. The KJQ versus yield stress plots confirmed that the fracture toughness was inversely proportional to yield strength. It was found, however, that the toughness data for some NFAs were outside the band of the integrated dataset at given strength level, which indicates either a significant improvement or deterioration in mechanical properties due to fundamental changes in deformation and fracture mechanisms. When compared to the behavior of NFAs, the FM steels have shown much less strength dependence and formed narrow fracture toughness data bands at a significantly lower strength region. It appeared that at high temperatures ≥600 °C the NFAs cannot retain the nanostructure advantage of high strength and high toughness either by high-temperature embrittlement or by excessive loss of strength. Finally, irradiation studies have revealed, however, that the NFAs have much stronger radiation resistance than tempered martensitic steels, such as lower radiation-induced swelling, finer helium bubble formation, lower irradiation creep rate and reduced low temperature embrittlement.

  5. Influence of structural-phase state of ferritic-martensitic steels on the helium porosity development

    NASA Astrophysics Data System (ADS)

    Chernov, I. I.; Staltsov, M. S.; Kalin, B. A.; Bogachev, I. A.; Guseva, L. Yu; Dzhumaev, P. S.; Emelyanova, O. V.; Drozhzhina, M. V.; Manukovsky, K. V.; Nikolaeva, I. D.

    2016-04-01

    Transmission electron microscopy (TEM) has been used to study the effect of the initial structural-phase state (SPhS) of ferritic-martensitic steels EK-181, EP-450 and EP-450- ODS (with 0.5 wt.% nanoparticles of Y2O3) on the of helium porosity formation and gas swelling. Different SPhS of steel EK-181 was produced by water quenching, annealing, normalizing plus tempered, intensive plastic deformation by torsion (HPDT). Irradiation was carried out by He+-40 keV ions at 923 K up to fluence of 5-1020 He+/m2. It is shown that the water quenching causes the formation of uniformly distributed small bubbles (d¯ ∼ 2 nm) of the highest density (ρ∼ 1025 m-3). After normalization followed by tempering as well as after annealing bubbles distribution is highly non-uniform both by volume and in size. Very large faceted bubbles (pre-equilibrium gas-filled voids) are formed in ferrite grains resulting in high level of gas swelling of the irradiated layer with S = 4,9 ± 1,2 and 3.8 ± 0.9% respectively. Nano- and microcrystalline structure created by HPDT completely degenerate at irradiation temperature and ion irradiation formed bubbles of the same parameters as in the annealed steel. Bubbles formed in EP-450-ODS steel are smaller in size and density, which led to a decrease of helium swelling by 4 times (S = 0.8 ± 0.2%) as compared to the swelling of the matrix steel EP-450 (S = 3.1 ± 0.7%).

  6. Thermal helium desorption behavior in advanced ferritic steels

    NASA Astrophysics Data System (ADS)

    Kimura, Akihiko; Sugano, R.; Matsushita, Y.; Ukai, S.

    2005-02-01

    Thermal helium desorption measurements were performed to investigate the difference in the helium trapping and accumulation behavior among a reduced activation ferritic (RAF) steel and oxide dispersion strengthening (ODS) steels after implantation of He+ ions at room temperature. Thermal helium desorption spectra (THDS) were obtained during annealing to 1200 °C at a heating rate of 1 °C/s. The THDS of the ODS steels are very similar to that of the RAF steel, except for the presence of the peak in the temperature range from 800 to 1000 °C, where the α γ transformation related helium desorption from the γ-phase is considered to occur in the 9Cr-ODS martensitic steels. The fraction of helium desorption becomes larger at higher temperatures, and this trend is increased with the amount of implanted helium. In the 9Cr-ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel. This suggests that the bubble formation was suppressed in the ODS steels. In the 12Cr-ODS steel, the fraction of helium desorption by bubble migration reached more than 90%, suggesting that the trapping capacity of martensite phase in the 9Cr-ODS steel is rather large.

  7. Mechanical alloying of lanthana-bearing nanostructured ferritic steels

    SciTech Connect

    Somayeh Paseban; Indrajit Charit; Yaqiao Q. Wu; Jatuporn Burns; Kerry N. Allahar; Darryl P. Butt; James I. Cole

    2013-09-01

    A novel nanostructured ferritic steel powder with the nominal composition Fe–14Cr–1Ti–0.3Mo–0.5La2O3 (wt.%) was developed via high energy ball milling. La2O3 was added to this alloy instead of the traditionally used Y2O3. The effects of varying the ball milling parameters, such as milling time, steel ball size and ball to powder ratio, on the mechanical properties and micro structural characteristics of the as-milled powder were investigated. Nanocrystallites of a body-centered cubic ferritic solid solution matrix with a mean size of approximately 20 nm were observed by transmission electron microscopy. Nanoscale characterization of the as-milled powder by local electrode atom probe tomography revealed the formation of Cr–Ti–La–O-enriched nanoclusters during mechanical alloying. The Cr:Ti:La:O ratio is considered “non-stoichiometric”. The average size (radius) of the nanoclusters was about 1 nm, with number density of 3.7 1024 m3. The mechanism for formation of nanoclusters in the as-milled powder is discussed. La2O3 appears to be a promising alternative rare earth oxide for future nanostructured ferritic steels.

  8. SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

    DOEpatents

    Beaver, R.J.; Cherubini, J.H.

    1963-05-14

    A process is described for separating uranium from a nuclear fuel element comprising a uranium-containing core and a ferritic stainless steel clad by heating said element in a non-carburizing atmosphere at a temperature in the range 850-1050 un. Concent 85% C, rapidly cooling the heated element through the temperature range 815 un. Concent 85% to 650 EC to avoid annealing said steel, and then contacting the cooled element with an aqueous solution of nitric acid to selectively dissolve the uranium. (AEC)

  9. Tensile properties of CLAM steel irradiated up to 20.1 dpa in STIP-V

    NASA Astrophysics Data System (ADS)

    Ge, Hongen; Peng, Lei; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic steel (CLAM) were irradiated in the fifth experiment of SINQ Target Irradiation Program (STIP-V) up to 20.1 dpa/1499 appm He/440 °C. Tensile tests were performed at room temperature (R.T) and irradiation temperatures (Tirr) in the range of 25-450 °C. The tensile results demonstrated strong effect of irradiation dose and irradiation temperature on hardening and embrittlement. With Tirr below ˜314 °C, CLAM steel specimens tested at R.T and Tirr showed similar evolution trend with irradiation dose, compared to other reduced activation ferritic/martensitic (RAFM) steels in similar irradiation conditions. At higher Tirr above ˜314 °C, it is interesting that the hardening effect decreases and the ductility seems to recover, probably due to a strong effect of high irradiation temperature.

  10. Microstructure evolution during helium irradiation and post-irradiation annealing in a nanostructured reduced activation steel

    NASA Astrophysics Data System (ADS)

    Liu, W. B.; Ji, Y. Z.; Tan, P. K.; Zhang, C.; He, C. H.; Yang, Z. G.

    2016-10-01

    Severe plastic deformation, intense single-beam He-ion irradiation and post-irradiation annealing were performed on a nanostructured reduced activation ferritic/martensitic (RAFM) steel to investigate the effect of grain boundaries (GBs) on its microstructure evolution during these processes. A surface layer with a depth-dependent nanocrystalline (NC) microstructure was prepared in the RAFM steel using surface mechanical attrition treatment (SMAT). Microstructure evolution after helium (He) irradiation (24.8 dpa) at room temperature and after post-irradiation annealing was investigated using Transmission Electron Microscopy (TEM). Experimental observation shows that GBs play an important role during both the irradiation and the post-irradiation annealing process. He bubbles are preferentially trapped at GBs/interfaces during irradiation and cavities with large sizes are also preferentially trapped at GBs/interfaces during post-irradiation annealing, but void denuded zones (VDZs) near GBs could not be unambiguously observed. Compared with cavities at GBs and within larger grains, cavities with smaller size and higher density are found in smaller grains. The average size of cavities increases rapidly with the increase of time during post-irradiation annealing at 823 K. Cavities with a large size are observed just after annealing for 5 min, although many of the cavities with small sizes also exist after annealing for 240 min. The potential mechanism of cavity growth behavior during post-irradiation annealing is also discussed.

  11. HRTEM Study of the Role of Nanoparticles in ODS Ferritic Steel

    SciTech Connect

    Hsiung, L; Tumey, S; Fluss, M; Serruys, Y; Willaime, F

    2011-08-30

    Structures of nanoparticles and their role in dual-ion irradiated Fe-16Cr-4.5Al-0.3Ti-2W-0.37Y{sub 2}O{sub 3} (K3) ODS ferritic steel produced by mechanical alloying (MA) were studied using high-resolution transmission electron microscopy (HRTEM) techniques. The observation of Y{sub 4}Al{sub 2}O{sub 9} complex-oxide nanoparticles in the ODS steel imply that decomposition of Y{sub 2}O{sub 3} in association with internal oxidation of Al occurred during mechanical alloying. HRTEM observations of crystalline and partially crystalline nanoparticles larger than {approx}2 nm and amorphous cluster-domains smaller than {approx}2 nm provide an insight into the formation mechanism of nanoparticles/clusters in MA/ODS steels, which we believe involves solid-state amorphization and re-crystallization. The role of nanoparticles/clusters in suppressing radiation-induced swelling is revealed through TEM examinations of cavity distributions in (Fe + He) dual-ion irradiated K3-ODS steel. HRTEM observations of helium-filled cavities (helium bubbles) preferably trapped at nanoparticle/clusters in dual-ion irradiated K3-ODS are presented.

  12. Optimization and testing results of Zr-bearing ferritic steels

    SciTech Connect

    Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata; Sridharan, K.

    2014-09-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional

  13. Development of oxide dispersion strengthened ferritic steels for fusion

    SciTech Connect

    Mukhopadhyay, D.K.; Froes, F.H.; Gelles, D.S.

    1998-03-01

    An oxide dispersion strengthened (ODS) ferritic steel with high temperature strength has been developed in line with low activation criteria for application in fusion power systems. The composition Fe-13.5Cr-2W-0.5Ti-0.25Y{sub 2}O{sup 3} was chosen to provide a minimum chromium content to insure fully delta-ferrite stability. High temperature strength has been demonstrated by measuring creep response of the ODS alloy in uniaxial tension at 650 and 900 C in an inert atmosphere chamber. Results of tests at 900 C demonstrate that this alloy has creep properties similar to other alloys of similar design and can be considered for use in high temperature fusion power system designs. The alloy selection process, materials production, microstructural evaluation and creep testing are described.

  14. Effect of internal nitriding on the fatigue strength of ferritic corrosion-resistant steel

    NASA Astrophysics Data System (ADS)

    Rogachev, S. O.; Nikulin, S. A.; Terent'ev, V. F.; Khatkevich, V. M.; Prosvirnin, D. V.; Savicheva, R. O.

    2015-04-01

    The effect of internal nitriding and subsequent annealing on the mechanical properties of ferritic corrosion-resistance 08Kh17T steel has been studied during static and cyclic loading. Nitriding was shown to increase the static and cyclic strength of ferritic steel substantially and to decrease its plasticity slightly. These changes are confirmed by results of fractographic studies.

  15. 78 FR 63517 - Control of Ferrite Content in Stainless Steel Weld Metal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... Information The NRC published DG-1279 in the Federal Register on October 3, 2012 (77 FR 60479), for a 60-day... COMMISSION Control of Ferrite Content in Stainless Steel Weld Metal AGENCY: Nuclear Regulatory Commission... revision to Regulatory Guide (RG) 1.31, ``Control of Ferrite Content in Stainless Steel Weld Metal.''...

  16. Cr-W-V bainitic/ferritic steel with improved strength and toughness and method of making

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.

    1994-01-01

    A high strength, high toughness Cr-W-V ferritic steel composition suitable for fast induced-radioactivity (FIRD) decay after irradiation in a fusion reactor comprises 2.5-3.5 wt % Cr, 2. This invention was made with Government support under contract DE-AC05-840R21400 awarded by the U.S. Department of Energy to Martin Marietta Energy Systems, Inc. and the Government has certain rights in this invention.

  17. Liquid metal embrittlement susceptibility of ferritic martensitic steel in liquid lead alloys

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Bosch, R. W.; Sapundjiev, D.; Almazouzi, A.

    2008-06-01

    The susceptibility of the ferritic-martensitic steels T91 and EUROFER97 to liquid metal embrittlement (LME) in lead alloys has been examined under various conditions. T91, which is currently the most promising candidate material for the high temperature components of the future accelerator driven system (ADS) was tested in liquid lead bismuth eutectic (LBE), whereas the reduced activation steel, EUROFER97 which is under consideration to be the structural steel for fusion reactors was tested in liquid lead lithium eutectic. These steels, similar in microstructure and mechanical properties in the unirradiated condition were tested for their susceptibility to LME as function of temperature (150-450 °C) and strain rate (1 × 10 -3-1 × 10 -6 s -1). Also, the influence of pre-exposure and surface stress concentrators was evaluated for both steels in, respectively, liquid PbBi and PbLi environment. To assess the LME effect, results of the tests in liquid metal environment are compared with tests in air or inert gas environment. Although both unirradiated and irradiated smooth ferritic-martensitic steels do not show any or little deterioration of mechanical properties in liquid lead alloy environment compared to their mechanical properties in gas as function of temperature and strain rate, pre-exposure or the presence of surface stress concentrators does lead to a significant decrease in total elongation for certain test conditions depending on the type of liquid metal environment. The results are discussed in terms of wetting enhanced by liquid metal corrosion or crack initiation processes.

  18. Corrosion Performance of Ferritic Steel for SOFC Interconnect Applications

    SciTech Connect

    Ziomek-Moroz, M.; Holcomb, G.R.; Covino, B.S., Jr.; Bullard, S.J.; Jablonski, P.D.; Alman, D.E.

    2006-11-01

    Ferritic stainless steels have been identified as potential candidates for interconnects in planar-type solid oxide fuel cells (SOFC) operating below 800ºC. Crofer 22 APU was selected for this study. It was studied under simulated SOFC-interconnect dual environment conditions with humidified air on one side of the sample and humidified hydrogen on the other side at 750ºC. The surfaces of the oxidized samples were studied by scanning electron microscopy (SEM) equipped with microanalytical capabilities. X-ray diffraction (XRD) analysis was also used in this study.

  19. Room temperature texturing of austenite/ferrite steel by electropulsing

    PubMed Central

    Rahnama, Alireza; Qin, Rongshan

    2017-01-01

    The work reports an experimental observation on crystal rotation in a duplex (austenite + ferrite) steel induced by the electropulsing treatment at ambient temperature, while the temperature rising due to ohmic heating in the treatment was negligible. The results demonstrate that electric current pulses are able to dissolve the initial material’s texture that has been formed in prior thermomechanical processing and to produce an alternative texture. The results were explained in terms of the instability of an interface under perturbation during pulsed electromigation. PMID:28195181

  20. Room temperature texturing of austenite/ferrite steel by electropulsing

    NASA Astrophysics Data System (ADS)

    Rahnama, Alireza; Qin, Rongshan

    2017-02-01

    The work reports an experimental observation on crystal rotation in a duplex (austenite + ferrite) steel induced by the electropulsing treatment at ambient temperature, while the temperature rising due to ohmic heating in the treatment was negligible. The results demonstrate that electric current pulses are able to dissolve the initial material’s texture that has been formed in prior thermomechanical processing and to produce an alternative texture. The results were explained in terms of the instability of an interface under perturbation during pulsed electromigation.

  1. Surface modification of ferritic steels using MEVVA and duoplasmatron ion sources

    SciTech Connect

    Kulevoy, Timur V. Orlov, Nikolay N.; Rogozhkin, Sergey V.; Bogachev, Alexey A.; Nikitin, Alexander A.; Iskandarov, Nasib A.; Golubev, Alexander A.; Chalyhk, Boris B.; Fedin, Petr A.; Sitnikov, Alexey L.; Kozlov, Alexander V.; Kuibeda, Rostislav P.; Andrianov, Stanislav L.; Kravchuk, Konstantin S.; Useinov, Alexey S.; Oks, Efim M.

    2016-02-15

    Metal Vapor Vacuum Arc (MEVVA) ion source (IS) is a unique tool for production of high intensity metal ion beam that can be used for material surface modification. From the other hand, the duoplasmatron ion source provides the high intensity gas ion beams. The MEVVA and duoplasmatron IS developed in Institute for Theoretical and Experimental Physics were used for the reactor steel surface modification experiments. Response of ferritic-martensitic steel specimens on titanium and nitrogen ions implantation and consequent vacuum annealing was investigated. Increase in microhardness of near surface region of irradiated specimens was observed. Local chemical analysis shows atom mixing and redistribution in the implanted layer followed with formation of ultrafine precipitates after annealing.

  2. Oxide dispersion strengthened ferritic steels: a basic research joint program in France

    NASA Astrophysics Data System (ADS)

    Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A. L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

    2014-12-01

    AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9-14% Cr ferritic-martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical-chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α‧ unmixing.

  3. Precipitation sequence in niobium-alloyed ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Fujita, Nobuhiro; Bhadeshia, H. K. D. H.; Kikuchi, Masao

    2004-03-01

    Niobium is an important alloying element in the design of heat-resistant ferritic stainless steels for automotive exhaust systems. When in solid solution, it improves both the high temperature strength and the resistance to thermal fatigue. However, it also forms several kinds of precipitates during service. These reactions have been modelled, taking into account the multicomponent nature of the diffusion process and allowing for capillarity effects. It has been possible to estimate not only the volume fractions but also the particle sizes for Fe2Nb (Laves phase) and Fe3Nb3C (M6C) carbide in a 19Cr-0.8Nb steel, with good agreement against experimental data.

  4. Effects of hydrogen isotopes in the irradiation damage of CLAM steel

    NASA Astrophysics Data System (ADS)

    Zhao, M. Z.; Liu, P. P.; Zhu, Y. M.; Wan, F. R.; He, Z. B.; Zhan, Q.

    2015-11-01

    The isotope effect of hydrogen in irradiation damage plays an important role in the development of reduced activation Ferritic/Martensitic steels in nuclear reactors. The evolutions of microstructures and mechanical properties of China low active martensitic (CLAM) steel subjected to hydrogen and deuterium ions irradiation are studied comparatively. Under the same irradiation conditions, larger size and smaller density of dislocation loops are generated by deuterium ion than by hydrogen ion. Irradiation hardening occurs under the ion irradiation and the hardening induced by hydrogen ion is higher than by deuterium ion. Moreover, the coarsening of M23C6 precipitates is observed, which can be explained by the solute drag mechanisms. It turns out that the coarsening induced by deuterium ion irradiation is more distinct than by hydrogen ion irradiation. No distinct variations for the compositions of M23C6 precipitates are found by a large number of statistical data after hydrogen isotopes irradiation.

  5. Toughness of 12%Cr ferritic/martensitic steel welds produced by non-arc welding processes

    SciTech Connect

    Ginn, B.J.; Gooch, T.G.

    1998-08-01

    Low carbon 12%Cr steels can offer reduced life cycle costs in many applications. The present work examined the behavior of commercial steels of varying composition when subject to low heat input welding by the electron beam (EB) process and to a forge cycle by linear friction welding (LFW). Charpy impact testing was carried out on the high temperature heat-affected zone (HAZ)/fusion boundary or weld interface, with metallographic examination. With EB welding, the ductile-brittle transition temperature (DBTT) was below 0 C (32 F) only for steel of low ferrite factor giving a fully martensitic weld area. Higher ferrite factor alloys showed predominantly ferritic transformed microstructures and a transition well above room temperature. Grain coarsening was found even with low EB process power, the peak grain size increasing with both heat input and steel ferrite factor. Use of LFW gave a fine weld area structure and DBTTs around 0 C even in high ferrite factor (FF) material.

  6. Surface modification to improve fireside corrosion resistance of Fe-Cr ferritic steels

    DOEpatents

    Park, Jong-Hee; Natesan, Krishnamurti; Rink, David L.

    2010-03-16

    An article of manufacture and a method for providing an Fe--Cr ferritic steel article of manufacture having a surface layer modification for corrosion resistance. Fe--Cr ferritic steels can be modified to enhance their corrosion resistance to liquid coal ash and other chemical environments, which have chlorides or sulfates containing active species. The steel is modified to form an aluminide/silicide passivating layer to reduce such corrosion.

  7. Diffusion bonding between ODS ferritic steel and F82H steel for fusion applications

    NASA Astrophysics Data System (ADS)

    Noh, Sanghoon; Kim, Byungjun; Kasada, Ryuta; Kimura, Akihiko

    2012-07-01

    Diffusion bonding techniques were employed to join high Cr oxide dispersion strengthened (ODS) ferritic steel (Fe-15Cr-2W-0.2Ti-0.35Y2O3) and F82H steel under uni-axial hydrostatic pressure using a high vacuum hot press, and the microstructure and mechanical properties of the joints were investigated. The dissimilar joints were bonded by solid-state diffusion bonding (SSDB) and liquid phase diffusion bonding (LPDB). After bonding process, heat treatments were conducted to utilize the phase transformation of F82H steel for recovering the martensitic structure. Tensile tests with miniaturized specimens were carried out to investigate and compare the bonding strengths of each joint. Microstructure was observed for the bonding interface, and fracture mode was investigated after the tensile tests. LPDB joint of interfacial F82H steel fully recovered to martensite phase by post-joining heat treatments, while SSDB joint had ferrite phases at the interface even after heat treatment, which is considered to be due to decarburization of F82H steel during the bonding process. Therefore it is considered that the insert material plays a role as diffusion barrier of carbon during LPDB process. Microstructure observations and tensile tests of the joints revealed that the LPDB joints possess suitable tensile properties which are comparable to that of F82H steel. This indicates that LPDB is more promising method to bond ODS-FS and F82H steel than SSDB.

  8. Notch-Fatigue Properties of Advanced TRIP-Aided Bainitic Ferrite Steels

    NASA Astrophysics Data System (ADS)

    Yoshikawa, Nobuo; Kobayashi, Junya; Sugimoto, Koh-ichi

    2012-11-01

    To develop a transformation-induced plasticity (TRIP)-aided bainitic ferrite steel (TBF steel) with high hardenability for a common rail of the next generation diesel engine, 0.2 pct C-1.5 pct Si-1.5 pct Mn-0.05 pct Nb TBF steels with different contents of Cr, Mo, and Ni were produced. The notch-fatigue strength of the TBF steels was investigated and was related to the microstructural and retained austenite characteristics. If Cr, Mo, and/or Ni were added to the base steel, then the steels achieved extremely higher notch-fatigue limits and lower notch sensitivity than base TBF steel and the conventional structural steels. This was mainly associated with (1) carbide-free and fine bainitic ferrite lath structure matrix without proeutectoid ferrite, (2) a large amount of fine metastable retained austenite, and (3) blocky martensite phase including retained austenite, which may suppress a fatigue crack initiation and propagation.

  9. New ferritic steels increase the thermal efficiency of steam turbines

    SciTech Connect

    Mayer, K.H.; Bakker, W.T.

    1996-12-31

    The further development of ferritic high-temperature-resistant 9--11%Cr steels has paved the way for fossil-fired power stations to be operated at turbine steam inlet temperatures of up to around 600 C and high supercritical steam pressures with a distinct improvement in thermal efficiency, a significant contribution towards reducing the environmental impact of SO{sub 2}, NO{sub x} and CO{sub 2} emissions and to a more economical utilization of fossil fuels. Advances in the development of these steels are primarily attributable to joint research projects undertaken by the manufacturers and operators of power stations in Japan (EPDC), in the USA (EPRI) and in Europe (COST 501). The report gives details on the results achieved under EPRI Research Project RP 140 3-15/23 on the creep behavior of modified 9%CrMo cast steel used in the manufacture of steam turbines for coal-fired power plants. The modified 9%CrMo cast steel also offers great benefits as regards improving the useful life and thermal efficiency of existing power plants.

  10. Effect of δ-ferrite evolution and high-temperature annealing on mechanical properties of 11Cr3W3Co ferritic/martensitic steel

    NASA Astrophysics Data System (ADS)

    Shang, Zhongxia; Shen, Yinzhong; Ji, Bo; Zhang, Lanting

    2016-03-01

    An 11Cr3W3Co ferritic/martensitic steel was annealed at 1100 °C for different time to gradually dissolve δ-ferrite, and then conducted tensile, hardness, and short-term creep tests in combination with microstructural characterization to study the effect of δ-ferrite on the mechanical properties of high-Cr ferritic/martensitic steels. The amount of δ-ferrite gradually decreased to a minimum value with increasing annealing time up to 10 h, and then tended to an ascending tendency when annealed for 15 and 20 h. Accordingly the tensile strength at 300 and 650 °C, and Vickers hardness of the steel had an increase and a decrease tendency when δ-ferrite amount decreased down to its minimum value and increased again, respectively. The short-term creep property at 210 MPa at 650 °C of the steel exhibited a serious degradation as annealing time gradually increased to 15 h. The morphology and orientation of δ-ferrite grains seriously affected the short-term creep property of the steel. δ-ferrite with a continuously bamboo-like shape parallel to loading direction effectively improved the short-term creep property of the steel at high temperature, while δ-ferrite with a granular or block shape seriously damaged the short-term creep property of the steel. These findings have also been discussed.

  11. In-beam fatigue of a ferritic-martensitic steel. First results

    NASA Astrophysics Data System (ADS)

    Marmy, P.

    1994-09-01

    Due to its pulsed operation mode, a fusion device will have to sustain thermal stresses and at the same time be exposed to a flux of 14 MeV neutrons. In order to simulate this irradiation condition, a new irradiation device has been developed, in which the specimen can be stress and strain-controlled during the irradiation. For the simulation of the fusion neutrons, a 590 MeV proton beam is used. This type of particle produces from spallation reactions the displacement damage and the helium typical of fusion neutrons in the material. The influence of the in situ deformation on the low cycle fatigue of a 12% Cr ferritic-martensitic steel (MANET II) has been investigated. The results are compared with results from non-irradiated specimens and from specimens tested after irradiation to the same end-of-life fluence. The effects of the different conditions are reported for a temperature TTest = TIrr = 573 K and a total imposed strain of 0.7%.

  12. Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiß, O. J.; Aktaa, J.; Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V.

    2011-10-01

    Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 °C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

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

    DOEpatents

    Leitnaker, J.M.

    Austenitic stainless steel alloys containing delta ferrite, such as are used as weld deposits, are protected against the transformation of delta ferrite to sigma phase during aging by the presence of carbon plus nitrogen in a weight percent 0.015 to 0.030 times the volume percent ferrite present in the alloy. The formation of chi phase upon aging is controlled by controlling the Mo content.

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

    DOEpatents

    Leitnaker, James M.

    1981-01-01

    Austenitic stainless steel alloys containing delta ferrite, such as are used as weld deposits, are protected against the transformation of delta ferrite to sigma phase during aging by the presence of carbon plus nitrogen in a weight percent 0.015-0.030 times the volume percent ferrite present in the alloy. The formation of chi phase upon aging is controlled by controlling the Mo content.

  15. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Saleh, T. A.; Anderoglu, O.; Romero, T. J.; Odette, G. R.; Yamamoto, T.; Li, S.; Cole, J. I.; Fielding, R.

    2016-01-01

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress-strain curves are analyzed to provide true stress-strain constitutive σ(ɛ) laws for all of these alloys. In the irradiated condition, the σ(ɛ) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where the latter can be understood in terms of the alloy's σ(ɛ) behavior. Increases in the average σ(ɛ) in the range of 0-10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are also analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  16. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    DOE PAGES

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; ...

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms ofmore » the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.« less

  17. Precipitates and boundaries interaction in ferritic ODS steels

    NASA Astrophysics Data System (ADS)

    Sallez, Nicolas; Hatzoglou, Constantinos; Delabrouille, Fredéric; Sornin, Denis; Chaffron, Laurent; Blat-Yrieix, Martine; Radiguet, Bertrand; Pareige, Philippe; Donnadieu, Patricia; Bréchet, Yves

    2016-04-01

    In the course of a recrystallization study of Oxide Dispersion Strengthened (ODS) ferritic steels during extrusion, particular interest was paid to the (GB) Grain Boundaries interaction with precipitates. Complementary and corresponding characterization experiments using Transmission Electron Microscopy (TEM), Energy Dispersive X-ray spectroscopy (EDX) and Atom Probe Tomography (APT) have been carried out on a voluntarily interrupted extrusion or extruded samples. Microscopic observations of Precipitate Free Zones (PFZ) and precipitates alignments suggest precipitate interaction with migrating GB involving dissolution and Oswald ripening of the precipitates. This is consistent with the local chemical information gathered by EDX and APT. This original mechanism for ODS steels is similar to what had been proposed in the late 80s for similar observation made on Ti alloys reinforced by nanosized yttrium oxides: An interaction mechanism between grain boundaries and precipitates involving a diffusion controlled process of precipitates dissolution at grain boundaries. It is believed that this mechanism can be of primary importance to explain the mechanical behaviour of such steels.

  18. Sensitization and stabilization of type 409 ferritic stainless steel

    SciTech Connect

    Fritz, J.D.; Franson, I.A.

    1997-08-01

    Type 409 (UNS S40900) ferritic stainless steel, used widely in automotive exhaust systems, can be subject to intergranular corrosion (IGC) of weld heat-affected zones (HAZ), even though ASTM stabilization requirements (Ti = 6 {times} C) are met. A boiling Cu/6% CuSO{sub 4}/0.5% H{sub 2}SO{sub 4} test is shown to be appropriate for detecting IGC of welds and HAZ. This test was used to establish stabilization requirements for type 409, whether dual-stabilized with Ti + Nb or singly stabilized with Ti alone. It was found that the stabilization requirement should be Ti + Nb {ge} 0.08 + 8 (C + N). Benefits of dual stabilization include improved surface quality and formability without sacrifice of mechanical properties or weldability.

  19. New grain formation during warm deformation of ferritic stainless steel

    SciTech Connect

    Belyakov, A.; Sakai, Taku; Kaibyshev, R.

    1998-01-01

    Microstructural evolution accompanied by localization of plastic flow was studied in compression of a ferritic stainless steel with high stacking fault energy (SFE) at 873 K ({approx} 0.5 Tm). The structure evolution is characterized by the formation of dense dislocation walls at low strains and subsequently of microbands and their clusters at moderate strains, followed by the evolution of fragmented structure inside the clusters of microbands at high strains. The misorientations of the fragmented boundaries and the fraction of high-angle grain boundaries increase substantially with increasing strain. Finally, further straining leads to the formation of new fine grains with high-angle boundaries, which become more equiaxed than the previous fragmented structure. The mechanisms operating during such structure changes are discussed in detail.

  20. A reassessment of the effects of helium on Charpy impact properties of ferritic/martensitic steels

    SciTech Connect

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

    1998-03-01

    To test the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of tests performed on specimens irradiated in reactors with very different neutron spectra, and isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400 C to 100 dpa and 1000 appm He will result in a ductile to brittle transition temperature shift of over 500 C. However, the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations also support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in HFIR. The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation.

  1. A review of some effects of helium on charpy impact properties of ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    To evaluate the effect of helium on Charpy impact properties of ferritic/martensitic steels, two approaches are reviewed: quantification of results of earlier tests performed by other researchers on specimens irradiated in reactors with very different neutron spectra, and evaluation of isotopic tailoring experiments. Data analysis can show that if the differences in reactor response are indeed due to helium effects, then irradiation in a fusion machine at 400°C to 100 dpa and 1000 appm He will result in a ductile-to-brittle transition temperature (DBTT) shift of over 500°C. However, it can be shown that the response as a function of dose and helium level is unlikely to be simply due to helium based on physical reasoning. Shear punch tests and microstructural examinations support this conclusion based on irradiated samples of a series of alloys made by adding various isotopes of nickel in order to vary the production of helium during irradiation in High Flux Isotope Reactor (HFIR). The addition of nickel at any isotopic balance to the Fe-12Cr base alloy significantly increased the shear yield and maximum strengths of the alloys. However, helium itself, up to 75 appm at over 7 dpa appears to have little effect on the mechanical properties of the alloys. This behavior is instead understood to result from complex precipitation response. The database for effects of helium on embrittlement based on nickel additions is therefore probably misleading and experiments should be redesigned to avoid nickel precipitation.

  2. Proceedings of the IEA Working Group meeting on ferritic/martensitic steels

    SciTech Connect

    Klueh, R.L.

    1996-12-31

    An IEA working group on ferritic/martensitic steels for fusion applications, consisting of researchers from Japan, European Union, USA, and Switzerland, met at the headquarters of the Joint European Torus, Culham, UK. At the meeting, preliminary data generated on the large heats of steels purchased for the IEA program and on other heats of steels were presented and discussed. Second purpose of the meeting was to continue planning and coordinating the collaborative test program in progress on reduced-activation ferritic/martensitic steels. The majority of this report consists of viewographs for the presentations.

  3. 46 CFR 54.25-20 - Low temperature operation-ferritic steels with properties enhanced by heat treatment (modifies...

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 false Low temperature operation-ferritic steels with... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and...

  4. 46 CFR 54.25-20 - Low temperature operation-ferritic steels with properties enhanced by heat treatment (modifies...

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 false Low temperature operation-ferritic steels with... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and...

  5. 46 CFR 54.25-10 - Low temperature operation-ferritic steels (replaces UCS-65 through UCS-67).

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 false Low temperature operation-ferritic steels (replaces UCS... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-10 Low temperature operation—ferritic steels (replaces UCS-65 through UCS-67). (a) Scope. (1)...

  6. 46 CFR 54.25-20 - Low temperature operation-ferritic steels with properties enhanced by heat treatment (modifies...

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Low temperature operation-ferritic steels with... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and...

  7. 46 CFR 54.25-10 - Low temperature operation-ferritic steels (replaces UCS-65 through UCS-67).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Low temperature operation-ferritic steels (replaces UCS... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-10 Low temperature operation—ferritic steels (replaces UCS-65 through UCS-67). (a) Scope. (1)...

  8. 46 CFR 54.25-20 - Low temperature operation-ferritic steels with properties enhanced by heat treatment (modifies...

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 false Low temperature operation-ferritic steels with... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and...

  9. 46 CFR 54.25-10 - Low temperature operation-ferritic steels (replaces UCS-65 through UCS-67).

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 false Low temperature operation-ferritic steels (replaces UCS... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-10 Low temperature operation—ferritic steels (replaces UCS-65 through UCS-67). (a) Scope. (1)...

  10. 46 CFR 54.25-10 - Low temperature operation-ferritic steels (replaces UCS-65 through UCS-67).

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 false Low temperature operation-ferritic steels (replaces UCS... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-10 Low temperature operation—ferritic steels (replaces UCS-65 through UCS-67). (a) Scope. (1)...

  11. 46 CFR 54.25-10 - Low temperature operation-ferritic steels (replaces UCS-65 through UCS-67).

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 false Low temperature operation-ferritic steels (replaces UCS... (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-10 Low temperature operation—ferritic steels (replaces UCS-65 through UCS-67). (a) Scope. (1)...

  12. 46 CFR 54.25-20 - Low temperature operation-ferritic steels with properties enhanced by heat treatment (modifies...

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 false Low temperature operation-ferritic steels with... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and...

  13. Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

    SciTech Connect

    Maloy, Stuart A.; Saleh, Tarik A.; Anderoglu, Osman; Romero, Tobias J.; Odette, G. Robert; Yamamoto, Takuya; Li, S.; Cole, James I.; Fielding, Randall

    2015-08-06

    Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσy) and reductions in uniform strain ductility (eu) are observed, where as the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσy, and vary more from alloy to alloy. The data are analyzed to establish relations between Δσy and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σyu). The latter shows that higher σyu correlates with lower Δσy. In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher eu than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

  14. Friction Stir Welding of HT9 Ferritic-Martensitic Steel: An Assessment of Microstructure and Properties

    DTIC Science & Technology

    2013-06-01

    development. While high speed steel or WC-Co tools can be used for aluminum and copper alloys, FSW of steel generally requires even more refractory... steel and the microstructure produced by FSW is much more critical than in aluminum alloys. The αγδ phase transformations can cause complex, multi...thesis explores the processing-microstructure-property relationships in friction stir welded ( FSW ) HT9A ferritic-martensitic steel . HT9 has previously

  15. Electrochemical and passivation behavior investigation of ferritic stainless steel in simulated concrete pore media.

    PubMed

    Luo, Hong; Su, Huaizhi; Dong, Chaofang; Xiao, Kui; Li, Xiaogang

    2015-12-01

    The applications of stainless steel are one of the most reliable solutions in concrete structures to reduce chloride-induced corrosion problems and increase the structures service life, however, due to high prices of nickel, especially in many civil engineering projects, the austenitic stainless steel is replaced by the ferritic stainless steels. Compared with austenite stainless steel, the ferritic stainless steel is known to be extremely resistant of stress corrosion cracking and other properties. The good corrosion resistance of the stainless steel is due to the formation of passive film. While, there is little literature about the electrochemical and passive behavior of ferritic stainless steel in the concrete environments. So, here, we present the several corrosion testing methods, such as the potentiodynamic measurements, EIS and Mott-Schottky approach, and the surface analysis methods like XPS and AES to display the passivation behavior of 430 ferritic stainless steel in alkaline solution with the presence of chloride ions. These research results illustrated a simple and facile approach for studying the electrochemical and passivation behavior of stainless steel in the concrete pore environments.

  16. Electrochemical and passivation behavior investigation of ferritic stainless steel in simulated concrete pore media

    PubMed Central

    Luo, Hong; Su, Huaizhi; Dong, Chaofang; Xiao, Kui; Li, Xiaogang

    2015-01-01

    The applications of stainless steel are one of the most reliable solutions in concrete structures to reduce chloride-induced corrosion problems and increase the structures service life, however, due to high prices of nickel, especially in many civil engineering projects, the austenitic stainless steel is replaced by the ferritic stainless steels. Compared with austenite stainless steel, the ferritic stainless steel is known to be extremely resistant of stress corrosion cracking and other properties. The good corrosion resistance of the stainless steel is due to the formation of passive film. While, there is little literature about the electrochemical and passive behavior of ferritic stainless steel in the concrete environments. So, here, we present the several corrosion testing methods, such as the potentiodynamic measurements, EIS and Mott–Schottky approach, and the surface analysis methods like XPS and AES to display the passivation behavior of 430 ferritic stainless steel in alkaline solution with the presence of chloride ions. These research results illustrated a simple and facile approach for studying the electrochemical and passivation behavior of stainless steel in the concrete pore environments. PMID:26501086

  17. Interatomic potential to study the formation of NiCr clusters in high Cr ferritic steels

    NASA Astrophysics Data System (ADS)

    Bonny, G.; Bakaev, A.; Olsson, P.; Domain, C.; Zhurkin, E. E.; Posselt, M.

    2017-02-01

    Under irradiation NiSiPCr clusters are formed in high-Cr ferritic martensitic steels as well as in FeCr model alloys. In the literature little is known about the origin and contribution to the hardening of these clusters. In this work we performed density functional theory (DFT) calculations to study the stability of small substitutional NiCr-vacancy clusters and interstitial configurations in bcc Fe. Based on DFT data and experimental considerations a ternary potential for the ferritic FeNiCr system was developed. The potential was applied to study the thermodynamic stability of NiCr clusters by means of Metropolis Monte Carlo (MMC) simulations. The results of our simulations show that Cr and Ni precipitate as separate fractions and suggest only a limited synergetic effect between Ni and Cr. Therefore our results suggest that the NiCrSiP clusters observed in experiments must be the result of other mechanisms than the synergy of Cr and Ni at thermal equilibrium.

  18. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA. Revision 1

    SciTech Connect

    Huang, F.H.

    1992-02-01

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

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

    SciTech Connect

    Huang, F.H.

    1992-02-01

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

  20. Effect of tin addition on the microstructure and properties of ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Li, Yang; Han, Ji-peng; Jiang, Zhou-hua; He, Pan

    2015-01-01

    This article reports the effects of Sn on the inclusions as well as the mechanical properties and hot workability of ferritic stainless steel. Precipitation phases and inclusions in Sn-bearing ferritic stainless steel were observed, and the relationship between the workability and the microstructure of the steel was established. Energy-dispersive X-ray spectroscopic analysis of the steel reveals that an almost pure Sn phase forms and MnS-Sn compound inclusions appear in the steel with a higher Sn content. Little Sn segregation was observed in grain boundaries and in the areas around sulfide inclusions; however, the presence of Sn does not adversely affect the workability of the steel containing 0.4wt% Sn. When the Sn content is 0.1wt%-0.4wt%, Sn improves the tensile strength and the plastic strain ratio and also improves the plasticity with increasing temperature. A mechanism of improving the workability of ferritic stainless steel induced by Sn addition was discussed: the presence of Sn lowers the defect concentration in the ultra-pure ferritic lattice and the good distribution of tin in the lattice overcomes the problem of hot brittleness that occurs in low-carbon steel as a result of Sn segregation.

  1. Changes in the fracture strength parameters of ferritic-bainitic and bainitic pipe steels during operation

    NASA Astrophysics Data System (ADS)

    Zikeev, V. N.; Filippov, G. A.; Shabalov, I. P.; Livanova, O. V.; Solov'ev, D. M.

    2016-10-01

    The fracture strength and the sensitivity to delayed fracture of the pipes in oil-trunk pipelines that are made of ferritic-bainitic and bainitic steels are studied. The results of modeling of the delayed brittle fracture of pipe steel during a simultaneous action of mechanical stresses and a corrosive medium are presented.

  2. Corrosion behavior of oxide dispersion strengthened ferritic steels in supercritical water

    NASA Astrophysics Data System (ADS)

    Gao, Wenhua; Guo, Xianglong; Shen, Zhao; Zhang, Lefu

    2017-04-01

    The corrosion resistance of three different Cr content oxide dispersion strengthened (ODS) ferritic steels in supercritical water (SCW) and their passive films formed on the surface have been investigated. The results show that the dissolved oxygen (DO) and chemical composition have significant influence on the corrosion behavior of the ODS ferritic steels. In 2000 ppb DO SCW at 650 °C, the 14Cr-4Al ODS steel forms a tri-layer oxide film and the surface morphologies have experienced four structures. For the tri-layer oxide film, the middle layer is mainly Fe-Cr spinel and the Al is gradually enriched in the inner layer.

  3. Lanthana-bearing nanostructured ferritic steels via spark plasma sintering

    NASA Astrophysics Data System (ADS)

    Pasebani, Somayeh; Charit, Indrajit; Wu, Yaqiao; Burns, Jatuporn; Allahar, Kerry N.; Butt, Darryl P.; Cole, James I.; Alsagabi, Sultan F.

    2016-03-01

    A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and1050 °C for 45 min. Significant densification occurred at temperatures greater than 950 °C with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 °C, the number density of Cr-Ti-La-O-enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 × 1024 m-3. The La + Ti:O ratio was close to 1 after SPS at 950 and 1050 °C; however, the number density of nanoclusters decreased at 1050 °C. With SPS above 950 °C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.

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

    DOE PAGES

    Chen, Wei-Ying; Li, Meimei; Kirk, Marquis A.; ...

    2015-08-21

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

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

    SciTech Connect

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

    2015-08-21

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

  6. Microstructure and Mechanical Properties of a Nitride-Strengthened Reduced Activation Ferritic/Martensitic Steel

    NASA Astrophysics Data System (ADS)

    Zhou, Qiangguo; Zhang, Wenfeng; Yan, Wei; Wang, Wei; Sha, Wei; Shan, Yiyin; Yang, Ke

    2012-12-01

    Nitride-strengthened reduced activation ferritic/martensitic (RAFM) steels are developed taking advantage of the high thermal stability of nitrides. In the current study, the microstructure and mechanical properties of a nitride-strengthened RAFM steel with improved composition were investigated. Fully martensitic microstructure with fine nitrides dispersion was achieved in the steel. In all, 1.4 pct Mn is sufficient to suppress delta ferrite and assure the steel of the full martensitic microstructure. Compared to Eurofer97, the steel showed similar strength at room temperature but higher strength at 873 K (600 °C). The steel exhibited very high impact toughness and a low ductile-to-brittle transition temperature (DBTT) of 243 K (-30 °C), which could be further reduced by purification.

  7. Effect of recrystallization on ion-irradiation hardening and microstructural changes in 15Cr-ODS steel

    NASA Astrophysics Data System (ADS)

    Ha, Yoosung; Kimura, Akihiko

    2015-12-01

    The effects of recrystallization on ion-irradiation hardening and microstructural changes were investigated for a 15Cr-ODS ferritic steel. Dual ion-irradiation experiments were performed at 470 °C using 6.4 MeV Fe3+ ions simultaneously with energy-degraded 1 MeV He+ ions. The displacement of damage at 600 nm depth from the specimen surface was 30 dpa. Nano-indentation test with Berkovich type indentation tip was measured by constant stiffness measurement (CSM) technique. Results from nano-indentation tests indicate irradiation hardening in ODS steels even at 470 °C, while it wasn't observed in reduced activation ferritic steel. Recrystallized ODS steel shows a larger irradiation hardening, which is considered to be due to the reduction of grain boundaries and interfaces of matrix/oxide particles. In 20% cold rolled ODS steel after recrystallization, both the hardening and bubble number density were lower than those of recrystallized ODS steel, suggesting that dislocations generated by cold rolling suppress bubble formation. Based on the estimation of irradiation hardening from TEM observation results, it is considered that the bubbles are not the main factor controlling ion-irradiation hardening.

  8. HRTEM Study of Oxide Nanoparticles in K3-ODS Ferritic Steel Developed for Radiation Tolerance

    SciTech Connect

    Hsiung, L; Fluss, M; Tumey, S; Kuntz, J; El-Dasher, B; Wall, M; Choi, W; Kimura, A; Willaime, F; Serruys, Y

    2009-11-02

    Crystal and interfacial structures of oxide nanoparticles and radiation damage in 16Cr-4.5Al-0.3Ti-2W-0.37 Y{sub 2}O{sub 3} ODS ferritic steel have been examined using high-resolution transmission electron microscopy (HRTEM) techniques. Oxide nanoparticles with a complex-oxide core and an amorphous shell were frequently observed. The crystal structure of complex-oxide core is identified to be mainly monoclinic Y{sub 4}Al{sub 2}O{sub 9} (YAM) oxide compound. Orientation relationships between the oxide and the matrix are found to be dependent on the particle size. Large particles (> 20 nm) tend to be incoherent and have a spherical shape, whereas small particles (< 10 nm) tend to be coherent or semi-coherent and have a faceted interface. The observations of partially amorphous nanoparticles and multiple crystalline domains formed within a nanoparticle lead us to propose a three-stage mechanism to rationalize the formation of oxide nanoparticles containing core/shell structures in as-fabricated ODS steels. Effects of nanoparticle size and density on cavity formation induced by (Fe{sup 8+} + He{sup +}) dual-beam irradiation are briefly addressed.

  9. Development of nano-structured duplex and ferritic stainless steels by pulverisette planetary milling followed by pressureless sintering

    SciTech Connect

    R, Shashanka Chaira, D.

    2015-01-15

    Nano-structured duplex and ferritic stainless steel powders are prepared by planetary milling of elemental Fe, Cr and Ni powder for 40 h and then consolidated by conventional pressureless sintering. The progress of milling and the continuous refinement of stainless steel powders have been confirmed by means of X-ray diffraction and scanning electron microscopy. Activation energy for the formation of duplex and ferritic stainless steels is calculated by Kissinger method using differential scanning calorimetry and is found to be 159.24 and 90.17 KJ/mol respectively. Both duplex and ferritic stainless steel powders are consolidated at 1000, 1200 and 1400 °C in argon atmosphere to study microstructure, density and hardness. Maximum sintered density of 90% and Vickers microhardness of 550 HV are achieved for duplex stainless steel sintered at 1400 °C for 1 h. Similarly, 92% sintered density and 263 HV microhardness are achieved for ferritic stainless steel sintered at 1400 °C. - Highlights: • Synthesized duplex and ferritic stainless steels by pulverisette planetary milling • Calculated activation energy for the formation of duplex and ferritic stainless steels • Studied the effect of sintering temperature on density, hardness and microstructure • Duplex stainless steel exhibits 90% sintered density and microhardness of 550 HV. • Ferritic stainless steel shows 92% sintered density and 263 HV microhardness.

  10. Positron annihilation study of proton-irradiated reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Liu, Xiangbing; Wang, Rongshan; Ren, Ai; Huang, Ping; Wu, Yichu; Jiang, Jing; Zhang, Chonghong; Wang, Xitao

    2012-10-01

    The microstructures, irradiation-induced defects and changes of mechanical property of Chinese domestic A508-3 steels after proton irradiation were investigated by TEM, positron lifetime, slow positron beam Doppler broadening spectroscopy and hardness measurements. The defects were induced by 240 keV proton irradiation with fluences of 1.25×1017 ions cm-2 (0.26 dpa), 2.5×1017 ions cm-2 (0.5 dpa), and 5.0×1017 ions cm-2 (1.0 dpa). The TEM observation revealed that the as-received steel had typical bainitic-ferritic microstructures. It was also observed that Doppler broadening S-parameter and average lifetime increased with dose level owing to the formation of defects and voids induced by proton irradiation. The correlation between positron parameters and hardness was found.

  11. Hydrogen-induced defects in austenite and ferrite of a duplex steel.

    PubMed

    Głowacka, A; Swiatnicki, W A; Jezierska, E

    2006-09-01

    The influence of hydrogen on the microstructure of two types of austeno-ferritic duplex stainless steel (Cr26-Ni6 model steel and Cr22-Ni5-Mo3 commercial steel), each of them after two thermo-mechanical treatments, was investigated. The aim of this study was to reveal microstructural changes appearing during the hydrogen charging and particularly to clarify the occurrence of phase transformations induced by hydrogen. The specific microstructural changes in the ferrite (alpha) and austenite (gamma) of both types of steel were observed. A strong increase of dislocation density was noticed in the alpha phase. In the case of model steel, longer hydrogen charging times led to significant ferrite grain refinement. In the commercial steel, the strips and twin plates appeared in the ferrite after hydrogenation. The appearance of stacking faults was revealed in the gamma phase. The martensite laths appeared in austenite after longer hydrogenation times. It seems that the microstructural changes gave rise to the formation of microcracks in the alpha and gamma phases as well as on the alpha/gamma interphase boundaries.

  12. Carbon concentration measurements by atom probe tomography in the ferritic phase of high-silicon steels

    DOE PAGES

    Rementeria, Rosalia; Poplawsky, Jonathan D.; Aranda, Maria M.; ...

    2016-12-19

    Current studies using atom probe tomography (APT) show that bainitic ferrite formed at low temperature contains more carbon than what is consistent with the paraequilibrium phase diagram. However, nanocrystalline bainitic ferrite exhibits a non-homogeneous distribution of carbon atoms in arrangements with specific compositions, i.e. Cottrell atmospheres, carbon clusters, and carbides, in most cases with a size of a few nanometers. The ferrite volume within a single platelet that is free of these carbon-enriched regions is extremely small. Proximity histograms can be compromised on the ferrite side, and a great deal of care should be taken to estimate the carbon contentmore » in regions of bainitic ferrite free from carbon agglomeration. For this purpose, APT measurements were first validated for the ferritic phase in a pearlitic sample and further performed for the bainitic ferrite matrix in high-silicon steels isothermally transformed between 200 °C and 350 °C. Additionally, results were compared with the carbon concentration values derived from X-ray diffraction (XRD) analyses considering a tetragonal lattice and previous APT studies. In conclusion, the present results reveal a strong disagreement between the carbon content values in the bainitic ferrite matrix as obtained by APT and those derived from XRD measurements. Those differences have been attributed to the development of carbon-clustered regions with an increased tetragonality in a carbon-depleted matrix.« less

  13. Carbon concentration measurements by atom probe tomography in the ferritic phase of high-silicon steels

    SciTech Connect

    Rementeria, Rosalia; Poplawsky, Jonathan D.; Aranda, Maria M.; Guo, Wei; Jimenez, Jose A.; Garcia-Mateo, Carlos; Caballero, Francisca G.

    2016-12-19

    Current studies using atom probe tomography (APT) show that bainitic ferrite formed at low temperature contains more carbon than what is consistent with the paraequilibrium phase diagram. However, nanocrystalline bainitic ferrite exhibits a non-homogeneous distribution of carbon atoms in arrangements with specific compositions, i.e. Cottrell atmospheres, carbon clusters, and carbides, in most cases with a size of a few nanometers. The ferrite volume within a single platelet that is free of these carbon-enriched regions is extremely small. Proximity histograms can be compromised on the ferrite side, and a great deal of care should be taken to estimate the carbon content in regions of bainitic ferrite free from carbon agglomeration. For this purpose, APT measurements were first validated for the ferritic phase in a pearlitic sample and further performed for the bainitic ferrite matrix in high-silicon steels isothermally transformed between 200 °C and 350 °C. Additionally, results were compared with the carbon concentration values derived from X-ray diffraction (XRD) analyses considering a tetragonal lattice and previous APT studies. In conclusion, the present results reveal a strong disagreement between the carbon content values in the bainitic ferrite matrix as obtained by APT and those derived from XRD measurements. Those differences have been attributed to the development of carbon-clustered regions with an increased tetragonality in a carbon-depleted matrix.

  14. Method for reducing formation of electrically resistive layer on ferritic stainless steels

    SciTech Connect

    Rakowski, James M.

    2013-09-10

    A method of reducing the formation of electrically resistive scale on a an article comprising a silicon-containing ferritic stainless subjected to oxidizing conditions in service includes, prior to placing the article in service, subjecting the article to conditions under which silica, which includes silicon derived from the steel, forms on a surface of the steel. Optionally, at least a portion of the silica is removed from the surface to placing the article in service. A ferritic stainless steel alloy having a reduced tendency to form silica on at least a surface thereof also is provided. The steel includes a near-surface region that has been depleted of silicon relative to a remainder of the steel.

  15. Method for reducing formation of electrically resistive layer on ferritic stainless steels

    DOEpatents

    Rakowski, James M.

    2017-02-28

    A method of reducing the formation of electrically resistive scale on a an article comprising a silicon-containing ferritic stainless subjected to oxidizing conditions in service includes, prior to placing the article in service, subjecting the article to conditions under which silica, which includes silicon derived from the steel, forms on a surface of the steel. Optionally, at least a portion of the silica is removed from the surface to placing the article in service. A ferritic stainless steel alloy having a reduced tendency to form silica on at least a surface thereof also is provided. The steel includes a near-surface region that has been depleted of silicon relative to a remainder of the steel.

  16. Elevated-Temperature Ferritic and Martensitic Steels and Their Application to Future Nuclear Reactors

    SciTech Connect

    Klueh, RL

    2005-01-31

    In the 1970s, high-chromium (9-12% Cr) ferritic/martensitic steels became candidates for elevated-temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe-12Cr-1Mo-0.5W-0.5Ni-0.25V-0.2C steel (composition in wt %), were considered in the United States, Europe, and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic, and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2-12% Cr for conventional power plants that are significant improvements over steels originally considered. This paper will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated-temperature mechanical properties will be emphasized. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

  17. HEAT INPUT AND POST WELD HEAT TREATMENT EFFECTS ON REDUCED-ACTIVATION FERRITIC/MARTENSITIC STEEL FRICTION STIR WELDS

    SciTech Connect

    Tang, Wei; Chen, Gaoqiang; Chen, Jian; Yu, Xinghua; Frederick, David Alan; Feng, Zhili

    2015-01-01

    Reduced-activation ferritic/martensitic (RAFM) steels are an important class of structural materials for fusion reactor internals developed in recent years because of their improved irradiation resistance. However, they can suffer from welding induced property degradations. In this paper, a solid phase joining technology friction stir welding (FSW) was adopted to join a RAFM steel Eurofer 97 and different FSW parameters/heat input were chosen to produce welds. FSW response parameters, joint microstructures and microhardness were investigated to reveal relationships among welding heat input, weld structure characterization and mechanical properties. In general, FSW heat input results in high hardness inside the stir zone mostly due to a martensitic transformation. It is possible to produce friction stir welds similar to but not with exactly the same base metal hardness when using low power input because of other hardening mechanisms. Further, post weld heat treatment (PWHT) is a very effective way to reduce FSW stir zone hardness values.

  18. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L.

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  19. Effect of constituent phase on mechanical properties of 9Cr-1WVTa reduced activation ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Lee, Chang-Hoon; Moon, Joonoh; Park, Min-Gu; Lee, Tae-Ho; Jang, Min-Ho; Kim, Hyoung Chan; Suh, Dong-Woo

    2014-12-01

    Influence of the formation of ferrite and accompanying carbides in martensite matrix on the tensile and Charpy impact properties was investigated for reduced activation ferritic-martensitic (RAFM) 9Cr-1WVTa steel. As the fractions of ferrite and carbide adjacent to the ferrite grain boundary increase, both tensile and Charpy impact properties deteriorated in as-normalized condition. In particular, the tensile strength and elongation decreased simultaneously, which is believed to be led by the localized deformation in ferrite which is softer than martensite, promoting the formation and growth of voids. In addition, the formation of ferrite was also detrimental to the Charpy impact properties regarding to the absorbed energy because the precipitation of carbides around ferrite were vulnerable to the nucleation and propagation of cleavage cracks. The degradation of tensile properties can be recovered by tempering, but the DBTT temperature still increases with presence of ferrite.

  20. Abnormal grain growth in Eurofer-97 steel in the ferrite phase field

    NASA Astrophysics Data System (ADS)

    Oliveira, V. B.; Sandim, H. R. Z.; Raabe, D.

    2017-03-01

    Reduced-activation ferritic-martensitic (RAFM) Eurofer-97 steel is a candidate material for structural applications in future fusion reactors. Depending on the amount of prior cold rolling strain and annealing temperature, important solid-state softening reactions such as recovery, recrystallization, and grain growth occur. Eurofer-97 steel was cold rolled up to 70, 80 and 90% reductions in thickness and annealed in the ferrite phase field (below ≈ 800 °C). Changes in microstructure, micro-, and mesotexture were followed by orientation mappings provided by electron backscatter diffraction (EBSD). Eurofer-97 steel undergoes abnormal grain growth above 650 °C and this solid-state reaction seems to be closely related to the high mobility of a few special grain boundaries that overcome pinning effects caused by fine particles. This solid-state reaction promotes important changes in the microstructure and microtexture of this steel. Abnormal grain growth kinetics for each condition was determined by means of quantitative metallography.

  1. Strength of "Light" Ferritic and Austenitic Steels Based on the Fe - Mn - Al - C System

    NASA Astrophysics Data System (ADS)

    Kaputkina, L. M.; Svyazhin, A. G.; Smarygina, I. V.; Kindop, V. E.

    2017-01-01

    The phase composition, the hardness, the mechanical properties at room temperature, and the resistance to hot (950 - 1000°C) and warm (550°C) deformation are studied for cast deformable "light" ferritic and austenitic steels of the Fe - (12 - 25)% Mn - (0 - 15)% Al - (0 - 2)% C system alloyed additionally with about 5% Ni. The high-aluminum high-manganese low-carbon and carbonless ferritic steels at a temperature of about 0.5 T melt have a specific strength close to that of the austenitic steels and may be used as weldable scale-resistant and wear-resistant materials. The high-carbon Fe - (20 - 24)% Mn - (5 - 9)% Al - 5% Ni - 1.5% C austenitic steels may be applied as light high-strength materials operating at cryogenic temperatures after a solution treatment and as scale- and heat-resistant materials in an aged condition.

  2. Summary of the IEA workshop/working group meeting on ferritic/martensitic steels for fusion

    SciTech Connect

    Klueh, R.L.

    1997-04-01

    An International Energy Agency (IEA) Working Group on Ferritic/Martensitic Steels for Fusion Applications, consisting of researchers from Japan, the European Union, the United States, and Switzerland, met at the headquarters of the Joint European Torus (JET), Culham, United Kingdom, 24-25 October 1996. At the meeting preliminary data generated on the large heats of steel purchased for the IEA program and on other heats of steels were presented and discussed. The second purpose of the meeting was to continue planning and coordinating the collaborative test program in progress on reduced-activation ferritic/martensitic steels. The next meeting will be held in conjunction with the International Conference on Fusion Reactor Materials (ICFRM-8) in Sendai, Japan, 23-31 October 1997.

  3. Fatigue strength of low-activation ferritic-martensitic high-chromium EK-181 steel

    NASA Astrophysics Data System (ADS)

    Kolmakov, A. G.; Terent'ev, V. F.; Prosvirnin, D. V.; Chernov, V. M.; Leont'eva-Smirnova, M. V.

    2016-04-01

    The static and cyclic mechanical properties of low-activation ferritic-martensitic EK-181 (Fe‒12Cr-2W-V-Ta-B-C) steel are studied in the temperature range 20-920°C (static tests) and at 20°C (cyclic tests). The fracture mechanisms of the steel under static tension and fatigue fracture conditions are analyzed by scanning electron microscopy.

  4. Positron annihilation Doppler broadening spectroscopy study on Fe-ion irradiated NHS steel

    NASA Astrophysics Data System (ADS)

    Zhu, Huiping; Wang, Zhiguang; Gao, Xing; Cui, Minghuan; Li, Bingsheng; Sun, Jianrong; Yao, Cunfeng; Wei, Kongfang; Shen, Tielong; Pang, Lilong; Zhu, Yabin; Li, Yuanfei; Wang, Ji; Song, Peng; Zhang, Peng; Cao, Xingzhong

    2015-02-01

    In order to study the evolution of irradiation-induced vacancy-type defects at different irradiation fluences and temperatures, a new type of ferritic/martensitic (F/M) steel named NHS (Novel High Silicon) was irradiated by 3.25 MeV Fe-ion at room temperature and 723 K to fluences of 4.3 × 1015 and 1.7 × 1016 ions/cm2. After irradiation, vacancy-type defects were investigated with variable-energy positron beam Doppler broadening spectra. Energetic Fe-ions produced a large number of vacancy-type defects in the NHS steel, but one single main type of vacancy-type defect was observed in both unirradiated and irradiated samples. The concentration of vacancy-type defects decreased with increasing temperature. With the increase of irradiation fluence, the concentration of vacancy-type defects increased in the sample irradiated at RT, whereas for the sample irradiated at 723 K, it decreased. The enhanced recombination between vacancies and excess interstitial Fe atoms from deeper layers, and high diffusion rate of self-interstitial atoms further improved by diffusion via grain boundary and dislocations at high temperature, are thought to be the main reasons for the reversed trend of vacancy-type defects between the samples irradiated at RT and 723 K.

  5. Hydrogen retention in ion irradiated steels

    SciTech Connect

    Hunn, J.D.; Lewis, M.B.; Lee, E.H.

    1998-11-01

    In the future 1--5 MW Spallation Neutron Source, target radiation damage will be accompanied by high levels of hydrogen and helium transmutation products. The authors have recently carried out investigations using simultaneous Fe/He,H multiple-ion implantations into 316 LN stainless steel between 50 and 350 C to simulate the type of radiation damage expected in spallation neutron sources. Hydrogen and helium were injected at appropriate energy and rate, while displacement damage was introduced by nuclear stopping of 3.5 MeV Fe{sup +}, 1 {micro}m below the surface. Nanoindentation measurements showed a cumulative increase in hardness as a result of hydrogen and helium injection over and above the hardness increase due to the displacement damage alone. TEM investigation indicated the presence of small bubbles of the injected gases in the irradiated area. In the current experiment, the retention of hydrogen in irradiated steel was studied in order to better understand its contribution to the observed hardening. To achieve this, the deuterium isotope ({sup 2}H) was injected in place of natural hydrogen ({sup 1}H) during the implantation. Trapped deuterium was then profiled, at room temperature, using the high cross-section nuclear resonance reaction with {sup 3}He. Results showed a surprisingly high concentration of deuterium to be retained in the irradiated steel at low temperature, especially in the presence of helium. There is indication that hydrogen retention at spallation neutron source relevant target temperatures may reach as high as 10%.

  6. Impact behavior of reduced-activation steels irradiated to 24 dpa

    SciTech Connect

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

    1996-04-01

    Charpy impact properties of eight reduced-activation Cr-W ferritic steels were determined after irradiation to {approx}21-24 dpa in the Fast Flux Test Facility (FFTF) at 365{degree}C. Chromium concentrations in the eight steels ranged from 2.25 to 12wt% Cr (steels contained {approx}0.1%C). the 2 1/4Cr steels contained variations of tungsten and vanadium, and the steels with 5, 9, and 12% Cr, contained a combination of 2% W and 0.25% V. A 9Cr in FFTF to {approx}6-8 and {approx}15-17 dpa. Irradiation caused an increase in the DBTT and decrease in the USE, but there was little further change in the DBTT from that observed after the 15-17 dpa irradiation, indicating that the shift had essentially saturated with fluence. The results are encouraging because they indicate that the effect of irradiation on toughness can be faorably affected by changing composition and microstructure.

  7. Influence of high temperature pre-deformation on the dissolution rate of delta ferrites in martensitic heat-resistant steels

    NASA Astrophysics Data System (ADS)

    Li, Junru; Liu, Jianjun; Jiang, Bo; Zhang, Chaolei; Liu, Yazheng

    2017-03-01

    The dissolution process of delta ferrites and the influence of high temperature pre-deformation on the dissolution rate of delta ferrites in martensitic heat-resistant steel 10Cr12Ni3Mo2VN were studied by isothermal heating and thermal simulation experiments. The precipitation temperature of delta ferrites in experimental steel is about 1195 °C. M23C6-type carbides incline to precipitate and coarsen at the boundaries of delta ferrites below 930 °C, and can be rapidly dissolved by heating at 1180 °C. The percentage of delta ferrites gradually decreases with heating time. And a Kolmogorov-Johnson-Mehl-Avrami equation was established to describe the dissolution process of delta ferrites at 1180 °C. High temperature pre-deformation can markedly increase the dissolution rate of delta ferrites. Pre-deformation can largely increase the interface area between delta ferrite and matrix and thus increase the unit-time diffusing quantities of alloying elements between delta ferrites and matrix. In addition, high temperature pre-deformation leads to dynamic recrystallization and increases the number of internal grain boundaries in the delta ferrites. This can also greatly increase the diffusing rate of alloying elements. In these cases, the dissolution of delta ferrites can be promoted.

  8. Influence of high temperature pre-deformation on the dissolution rate of delta ferrites in martensitic heat-resistant steels

    NASA Astrophysics Data System (ADS)

    Li, Junru; Liu, Jianjun; Jiang, Bo; Zhang, Chaolei; Liu, Yazheng

    2017-02-01

    The dissolution process of delta ferrites and the influence of high temperature pre-deformation on the dissolution rate of delta ferrites in martensitic heat-resistant steel 10Cr12Ni3Mo2VN were studied by isothermal heating and thermal simulation experiments. The precipitation temperature of delta ferrites in experimental steel is about 1195 °C. M23C6-type carbides incline to precipitate and coarsen at the boundaries of delta ferrites below 930 °C, and can be rapidly dissolved by heating at 1180 °C. The percentage of delta ferrites gradually decreases with heating time. And a Kolmogorov-Johnson-Mehl-Avrami equation was established to describe the dissolution process of delta ferrites at 1180 °C. High temperature pre-deformation can markedly increase the dissolution rate of delta ferrites. Pre-deformation can largely increase the interface area between delta ferrite and matrix and thus increase the unit-time diffusing quantities of alloying elements between delta ferrites and matrix. In addition, high temperature pre-deformation leads to dynamic recrystallization and increases the number of internal grain boundaries in the delta ferrites. This can also greatly increase the diffusing rate of alloying elements. In these cases, the dissolution of delta ferrites can be promoted.

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

    SciTech Connect

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

    1998-09-01

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

  10. Recent status and improvement of reduced-activation ferritic-martensitic steels for high-temperature service

    DOE PAGES

    Tan, L.; Katoh, Y.; Tavassoli, A. -A. F.; ...

    2016-07-26

    Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. Additionally, to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti,more » X = C/N) precipitates and reducing coarse M23C6 (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. We present and compare limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic with data for F82H and Eurofer97 irradiated up to ~70 displacements per atom at ~300–325 °C.« less

  11. Recent status and improvement of reduced-activation ferritic-martensitic steels for high-temperature service

    NASA Astrophysics Data System (ADS)

    Tan, L.; Katoh, Y.; Tavassoli, A.-A. F.; Henry, J.; Rieth, M.; Sakasegawa, H.; Tanigawa, H.; Huang, Q.

    2016-10-01

    Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. In addition to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M23C6 (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. Limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic are presented and compared with data for F82H and Eurofer97 irradiated up to ∼70 displacements per atom at ∼300-325 °C.

  12. Recent status and improvement of reduced-activation ferritic-martensitic steels for high-temperature service

    SciTech Connect

    Tan, L.; Katoh, Y.; Tavassoli, A. -A. F.; Henry, J.; Rieth, M.; Sakasegawa, H.; Tanigawa, H.; Huang, Q.

    2016-07-26

    Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. Additionally, to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M23C6 (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. We present and compare limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic with data for F82H and Eurofer97 irradiated up to ~70 displacements per atom at ~300–325 °C.

  13. Ultrahigh Charpy impact toughness (~450J) achieved in high strength ferrite/martensite laminated steels.

    PubMed

    Cao, Wenquan; Zhang, Mingda; Huang, Chongxiang; Xiao, Shuyang; Dong, Han; Weng, Yuqing

    2017-02-02

    Strength and toughness are a couple of paradox as similar as strength-ductility trade-off in homogenous materials, body-centered-cubic steels in particular. Here we report a simple way to get ultrahigh toughness without sacrificing strength. By simple alloying design and hot rolling the 5Mn3Al steels in ferrite/austenite dual phase temperature region, we obtain a series of ferrite/martensite laminated steels that show up-to 400-450J Charpy V-notch impact energy combined with a tensile strength as high as 1.0-1.2 GPa at room temperature, which is nearly 3-5 times higher than that of conventional low alloy steels at similar strength level. This remarkably enhanced toughness is mainly attributed to the delamination between ferrite and martensite lamellae. The current finding gives us a promising way to produce high strength steel with ultrahigh impact toughness by simple alloying design and hot rolling in industry.

  14. Impurity content of reduced-activation ferritic steels and a vanadium alloy

    SciTech Connect

    Klueh, R.L.; Grossbeck, M.L.; Bloom, E.E.

    1997-04-01

    Inductively coupled plasma mass spectrometry was used to analyze a reduced-activation ferritic/martensitic steel and a vanadium alloy for low-level impurities that would compromise the reduced-activation characteristics of these materials. The ferritic steel was from the 5-ton IEA heat of modified F82H, and the vanadium alloy was from a 500-kg heat of V-4Cr-4Ti. To compare techniques for analysis of low concentrations of impurities, the vanadium alloy was also examined by glow discharge mass spectrometry. Two other reduced-activation steels and two commercial ferritic steels were also analyzed to determine the difference in the level of the detrimental impurities in the IEA heat and steels for which no extra effort was made to restrict some of the tramp impurities. Silver, cobalt, molybdenum, and niobium proved to be the tramp impurities of most importance. The levels observed in these two materials produced with present technology exceeded the limits for low activation for either shallow land burial or recycling. The chemical analyses provide a benchmark for the improvement in production technology required to achieve reduced activation; they also provide a set of concentrations for calculating decay characteristics for reduced-activation materials. The results indicate the progress that has been made and give an indication of what must still be done before the reduced-activation criteria can be achieved.

  15. Ultrahigh Charpy impact toughness (~450J) achieved in high strength ferrite/martensite laminated steels

    PubMed Central

    Cao, Wenquan; Zhang, Mingda; Huang, Chongxiang; Xiao, Shuyang; Dong, Han; Weng, Yuqing

    2017-01-01

    Strength and toughness are a couple of paradox as similar as strength-ductility trade-off in homogenous materials, body-centered-cubic steels in particular. Here we report a simple way to get ultrahigh toughness without sacrificing strength. By simple alloying design and hot rolling the 5Mn3Al steels in ferrite/austenite dual phase temperature region, we obtain a series of ferrite/martensite laminated steels that show up-to 400–450J Charpy V-notch impact energy combined with a tensile strength as high as 1.0–1.2 GPa at room temperature, which is nearly 3–5 times higher than that of conventional low alloy steels at similar strength level. This remarkably enhanced toughness is mainly attributed to the delamination between ferrite and martensite lamellae. The current finding gives us a promising way to produce high strength steel with ultrahigh impact toughness by simple alloying design and hot rolling in industry. PMID:28150692

  16. Ultrahigh Charpy impact toughness (~450J) achieved in high strength ferrite/martensite laminated steels

    NASA Astrophysics Data System (ADS)

    Cao, Wenquan; Zhang, Mingda; Huang, Chongxiang; Xiao, Shuyang; Dong, Han; Weng, Yuqing

    2017-02-01

    Strength and toughness are a couple of paradox as similar as strength-ductility trade-off in homogenous materials, body-centered-cubic steels in particular. Here we report a simple way to get ultrahigh toughness without sacrificing strength. By simple alloying design and hot rolling the 5Mn3Al steels in ferrite/austenite dual phase temperature region, we obtain a series of ferrite/martensite laminated steels that show up-to 400–450J Charpy V-notch impact energy combined with a tensile strength as high as 1.0–1.2 GPa at room temperature, which is nearly 3–5 times higher than that of conventional low alloy steels at similar strength level. This remarkably enhanced toughness is mainly attributed to the delamination between ferrite and martensite lamellae. The current finding gives us a promising way to produce high strength steel with ultrahigh impact toughness by simple alloying design and hot rolling in industry.

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

    SciTech Connect

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

    1996-04-01

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

  18. Effect of irradiation temperature on void swelling of China Low Activation Martensitic steel (CLAM)

    SciTech Connect

    Zhao Fei; Qiao Jiansheng; Huang Yina; Wan Farong Ohnuki, Soumei

    2008-03-15

    CLAM is one composition of a Reduced Activation Ferritic/Martensitic steel (RAFM), which is being studied in a number of institutes and universities in China. The effect of electron-beam irradiation temperature on irradiation swelling of CLAM was investigated by using a 1250 kV High Voltage Electron Microscope (HVEM). In-situ microstructural observations indicated that voids formed at each experimental temperature - 723 K, 773 K and 823 K. The size and number density of voids increased with increasing irradiation dose at each temperature. The results show that CLAM has good swelling resistance. The maximum void swelling was produced at 723 K; the swelling was about 0.3% when the irradiation damage was 13.8 dpa.

  19. Effects of mechanical force on grain structures of friction stir welded oxide dispersion strengthened ferritic steel

    NASA Astrophysics Data System (ADS)

    Han, Wentuo; Kimura, Akihiko; Tsuda, Naoto; Serizawa, Hisashi; Chen, Dongsheng; Je, Hwanil; Fujii, Hidetoshi; Ha, Yoosung; Morisada, Yoshiaki; Noto, Hiroyuki

    2014-12-01

    The weldability of oxide dispersion strengthened (ODS) ferritic steels is a critical obstructive in the development and use of these steels. Friction stir welding has been considered to be a promising way to solve this problem. The main purpose of this work was to reveal the effects of mechanical force on grain structures of friction stir welded ODS ferritic steel. The grain appearances and the misorientation angles of grain boundaries in different welded zones were investigated by the electron backscatter diffraction (EBSD). Results showed that the mechanical force imposed by the stir tool can activate and promote the recrystallization characterized by the transformation of boundaries from LABs to HABs, and contribute to the grain refinement. The type of recrystallization in the stir zone can be classified as the continuous dynamic recrystallization (CDRX).

  20. Defect induced modification of structural, topographical and magnetic properties of zinc ferrite thin films by swift heavy ion irradiation

    NASA Astrophysics Data System (ADS)

    Raghavan, Lisha; Joy, P. A.; Vijaykumar, B. Varma; Ramanujan, R. V.; Anantharaman, M. R.

    2017-04-01

    Swift heavy ion irradiation provides unique ways to modify physical and chemical properties of materials. In ferrites, the magnetic properties can change significantly as a result of swift heavy ion irradiation. Zinc ferrite is an antiferromagnet with a Neel temperature of 10 K and exhibits anomalous magnetic properties in the nano regime. Ion irradiation can cause amorphisation of zinc ferrite thin films; thus the role of crystallinity on magnetic properties can be examined. The influence of surface topography in these thin films can also be studied. Zinc ferrite thin films, of thickness 320 nm, prepared by RF sputtering were irradiated with 100 MeV Ag ions. Structural characterization showed amorphisation and subsequent reduction in particle size. The change in magnetic properties due to irradiation was correlated with structural and topographical effects of ion irradiation. A rough estimation of ion track radius is done from the magnetic studies.

  1. Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

    NASA Astrophysics Data System (ADS)

    Lee, J. S.; Kim, I. S.; Kasada, R.; Kimura, A.

    2004-03-01

    The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10 18 and 1.02 × 10 19 n/cm 2 (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.

  2. Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.; Horváth, M.

    2012-11-01

    The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.

  3. Passivation and Corrosion Behavior of Modified Ferritic-Pearlitic Railway Axle Steels

    NASA Astrophysics Data System (ADS)

    Moon, A. P.; Sangal, S.; Srivastav, Simant; Gajbhiye, N. S.; Mondal, K.

    2015-01-01

    Electrochemical polarization behavior of two newly developed ferritic-pearlitic railway axle steels (MS3 and MS6) and the standard Indian conventional axle steel has been studied in sodium borate buffer solution of pH 8.4 with and without the presence of NaCl. The polarization behavior of both the new axle steels shows close resemblance, whereas, different polarization behavior has been observed for the conventional axle steel. Electrochemical impedance spectroscopy measurements have clearly reflected significantly improved passivation behavior for the newly developed steels compared to that of the conventional axle steel. NaCl salt fog exposure tests have also shown superior corrosion resistance of the newly developed axle steels as compared to the conventional axle steel. Higher surface roughness on the corroded conventional axle steel has also been observed compared to the smoother surface in case of the new axle steels. Higher corrosion resistance of the new axle steels has been attributed to their finer microstructure and strongly adherent protective rusts.

  4. Determining the shear fracture properties of HIP joints of reduced-activation ferritic/martensitic steel by a torsion test

    NASA Astrophysics Data System (ADS)

    Nozawa, Takashi; Noh, Sanghoon; Tanigawa, Hiroyasu

    2012-08-01

    Hot isostatic pressing (HIP) is a key technology used to fabricate a first wall with cooling channels for the fusion blanket system utilizing a reduced-activation ferritic/martensitic steel. To qualify the HIPped components, small specimen test techniques are beneficial not only to evaluate the thin-wall cooling channels containing the HIP joint but also to use in neutron irradiation studies. This study aims to develop the torsion test method with special emphasis on providing a reasonable and comprehensive method to determine interfacial shear properties of HIP joints during the torsional fracture process. Torsion test results identified that the torsion process shows yield of the base metal followed by non-elastic deformation due to work hardening of the base metal. By considering this work hardening issue, we propose a reasonable and realistic solution to determine the torsional yield shear stress and the ultimate torsional shear strength of the HIPped interface. Finally, a representative torsion fracture process was identified.

  5. Development and characterization of advanced 9Cr ferritic/martensitic steels for fission and fusion reactors

    NASA Astrophysics Data System (ADS)

    Saroja, S.; Dasgupta, A.; Divakar, R.; Raju, S.; Mohandas, E.; Vijayalakshmi, M.; Bhanu Sankara Rao, K.; Raj, Baldev

    2011-02-01

    This paper presents the results on the physical metallurgy studies in 9Cr Oxide Dispersion Strengthened (ODS) and Reduced Activation Ferritic/Martensitic (RAFM) steels. Yttria strengthened ODS alloy was synthesized through several stages, like mechanical milling of alloy powders and yttria, canning and consolidation by hot extrusion. During characterization of the ODS alloy, it was observed that yttria particles possessed an affinity for Ti, a small amount of which was also helpful in refining the dispersoid particles containing mixed Y and Ti oxides. The particle size and their distribution in the ferrite matrix, were studied using Analytical and High Resolution Electron Microscopy at various stages. The results showed a distribution of Y 2O 3 particles predominantly in the size range of 5-20 nm. A Reduced Activation Ferritic/Martensitic steel has also been developed with the replacement of Mo and Nb by W and Ta with strict control on the tramp and trace elements (Mo, Nb, B, Cu, Ni, Al, Co, Ti). The transformation temperatures ( Ac1, Ac3 and Ms) for this steel have been determined and the transformation behavior of the high temperature austenite phase has been studied. The complete phase domain diagram has been generated which is required for optimization of the processing and fabrication schedules for the steel.

  6. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

    SciTech Connect

    Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

    2013-01-01

    Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for the advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.

  7. Effect of Structural Heterogeneity on In Situ Deformation of Dissimilar Weld Between Ferritic and Austenitic Steel

    NASA Astrophysics Data System (ADS)

    Ghosh, M.; Santosh, R.; Das, S. K.; Das, G.; Mahato, B.; Korody, J.; Kumar, S.; Singh, P. K.

    2015-08-01

    Low-alloy steel and 304LN austenitic stainless steel were welded using two types of buttering material, namely 309L stainless steel and IN 182. Weld metals were 308L stainless steel and IN 182, respectively, for two different joints. Cross-sectional microstructure of welded assemblies was investigated. Microhardness profile was determined perpendicular to fusion boundary. In situ tensile test was performed in scanning electron microscope keeping low-alloy steel-buttering material interface at the center of gage length. Adjacent to fusion boundary, low-alloy steel exhibited carbon-depleted region and coarsening of matrix grains. Between coarse grain and base material structure, low-alloy steel contained fine grain ferrite-pearlite aggregate. Adjacent to fusion boundary, buttering material consisted of Type-I and Type-II boundaries. Within buttering material close to fusion boundary, thin cluster of martensite was formed. Fusion boundary between buttering material-weld metal and weld metal-304LN stainless steel revealed unmixed zone. All joints failed within buttering material during in situ tensile testing. The fracture location was different for various joints with respect to fusion boundary, depending on variation in local microstructure. Highest bond strength with adequate ductility was obtained for the joint produced with 309L stainless steel-buttering material. High strength of this weld might be attributed to better extent of solid solution strengthening by alloying elements, diffused from low-alloy steel to buttering material.

  8. Ferrite Measurement in Austenitic and Duplex Stainless Steel Castings - Final Report

    SciTech Connect

    Lundin, C.D.; Zhou, G.; Ruprecht, W.

    1999-08-01

    The ability to determine ferrite rapidly, accurately and directly on a finished casting, in the solution annealed condition, can enhance the acceptance, save on manufacturing costs and ultimately improve service performance of duplex stainless steel cast products. If the suitability of a non-destructive ferrite determination methodology can be demonstrated for standard industrial measurement instruments, the production of cast secondary standards for calibration of these instruments is a necessity. With these concepts in mind, a series of experiments were carried out to demonstrate, in a non-destructive manner, the proper methodology for determining ferrite content. The literature was reviewed, with regard to measurement techniques and vagaries, an industrial ferrite measurement round-robin was conducted, the effects of casting surface finish, preparation of the casting surface for accurate measurement and the evaluation of suitable means for the production of cast secondary standards for calibration were systematically investigated. The data obtained from this research program provide recommendations to ensure accurate, repeatable, and reproducible ferrite measurement and qualifies the Feritscope for field use on production castings.

  9. Ferrite Measurement in Austenitic and Duplex Stainless Steel Castings - Literature Review

    SciTech Connect

    Lundin, C.D.; Zhou, G.; Ruprecht, W.

    1999-08-01

    The ability to determine ferrite rapidly, accurately and directly on a finished casting, in the solution annealed condition, can enhance the acceptance, save on manufacturing costs and ultimately improve service performance of duplex stainless steel cast products. If the suitability of a non-destructive ferrite determination methodology can be demonstrated for standard industrial measurement instruments, the production of cast secondary standards for calibration of these instruments is a necessity. With these concepts in mind, a series of experiments were carried out to demonstrate, in a non-destructive manner, the proper methodology for determining ferrite content. The literature was reviewed, with regard to measurement techniques and vagaries, an industrial ferrite measurement round-robin was conducted, the effects of casting surface finish, preparation of the casting surface for accurate measurement and the evaluation of suitable means for the production of cast secondary standards for calibration were systematically investigated. The data obtained from this research program provides recommendations to insure accurate, repeatable and reproducible ferrite measurement and qualifies the Feritscope for field use on production castings.

  10. Adsorption of Pb(2+) from aqueous solution using spinel ferrite prepared from steel pickling sludge.

    PubMed

    Fang, Binbin; Yan, Yubo; Yang, Yang; Wang, Fenglian; Chu, Zhen; Sun, Xiuyun; Li, Jiansheng; Wang, Lianjun

    2016-01-01

    In this paper, spinel ferrite with high crystallinity and high saturation magnetization was successfully prepared from steel pickling sludge by adding iron source and precipitator in the hydrothermal condition. The obtained spinel ferrite was characterized by X-ray diffraction (XRD), field emission scanning electron microscopy (FE-SEM), vibrating sample magnetometer (VSM), and Zeta potential methods and investigated as an adsorbent for removal of Pb(2+) from aqueous solution. Batch experiments were performed by varying the pH values, contact time, temperature and initial metal concentration. The result of pH impact showed that the adsorption of Pb(2+) was a pH dependent process, and the pH 5.8 ± 0.2 was found to be the optimum condition. The achieved experimental data were analyzed with various kinetic and isotherm models. The kinetic studies revealed that Pb(2+) adsorption onto spinel ferrite followed a pseudo-second order model, and the Langmuir isotherm model provided the perfect fit to the equilibrium experimental data. At different temperatures, the maximum Pb(2+) adsorption capacities calculated from the Langmuir equation were in the range of 126.5-175.4 mg/g, which can be in competition with other adsorbents. The thermodynamic results showed that the spinel ferrite could spontaneously and endothermically adsorb Pb(2+) from aqueous solution. The regeneration studies showed that spinel ferrite could be used five times (removal efficiency (%) >90%) by desorption with HNO3 reagent.

  11. Blister formation on 13Cr2MoNbVB ferritic-martensitic steel exposed to hydrogen plasma

    NASA Astrophysics Data System (ADS)

    Nikitin, A. V.; Tolstolutskaya, G. D.; Ruzhytskyi, V. V.; Voyevodin, V. N.; Kopanets, I. E.; Karpov, S. A.; Vasilenko, R. L.; Garner, F. A.

    2016-09-01

    The influence of pre-irradiation specimen deformation level on surface blister formation and sub-surface cracking of dual-phase 13Cr2MoNbVB ferritic-martensitic steel was studied using glow discharge hydrogen plasma with ion energy of 1 keV to fluences of 2 × 1025 H/m2. Protium was used for most studies, but deuterium was used for measuring the depth dependence of hydrogen diffusion. Formation of blisters was observed in the temperature range 230-340 K. It was found that pre-irradiation deformation caused changes in the threshold fluences of blister formation and also in blister size distribution. Subsurface cracks located on grain boundaries far beyond the implantation zone were formed concurrently with blisters, arising from hydrogen diffusion and trapping at defects. It was observed that cracks as long as 1 mm in length were formed in 95% deformed steel at depths up to 500 μm from surface.

  12. Mechanical properties of 15%Mn steel with fine lamellar structure consisting of ferrite and austenite phases

    NASA Astrophysics Data System (ADS)

    Ueji, R.; Okitsu, Y.; Nakamura, T.; Takagi, Y.; Tanaka, Y.

    2010-07-01

    New steel with fine lamellar structure consisting of austenite and ferrite was developed. 15mass%Mn-3%Al-3%Si steel sheet was used in this study. First of all, the effect of the cooling rate on the microstructure was examined. The cooling at the slower speed of 100 deg/hour created the dual phase structure consisting of both austenite and ferrite. The additional rolling developed the fine lamellar duplex structure. Improvement of both the tensile strength and elongation was achieved by rolling. The strength increases furthermore by the rolling up to larger reduction. The 90% rolled sheet shows high tensile strength around 1000MPa with large elongation (15%-20%). These results indicate that the multi-phased structure with controlled lamellar morphology is beneficial for the management of both high strength and large ductility.

  13. Finite element residual stress analysis of induction heating bended ferritic steel piping

    NASA Astrophysics Data System (ADS)

    Kima, Jong Sung; Kim, Kyoung-Soo; Oh, Young-Jin; Chang, Hyung-Young; Park, Heung-Bae

    2014-10-01

    Recently, there is a trend to apply the piping bended by induction heating process to nuclear power plants. Residual stress can be generated due to thermo-mechanical mechanism during the induction heating bending process. It is well-known that the residual stress has important effect on crack initiation and growth. The previous studies have focused on the thickness variation. In part, some studies were performed for residual stress evaluation of the austenitic stainless steel piping bended by induction heating. It is difficult to find the residual stresses of the ferritic steel piping bended by the induction heating. The study assessed the residual stresses of induction heating bended ferriticsteel piping via finite element analysis. As a result, it was identified that high residual stresses are generated on local outersurface region of the induction heating bended ferritic piping.

  14. Finite element residual stress analysis of induction heating bended ferritic steel piping

    SciTech Connect

    Kima, Jong Sung; Kim, Kyoung-Soo; Oh, Young-Jin; Chang, Hyung-Young; Park, Heung-Bae

    2014-10-06

    Recently, there is a trend to apply the piping bended by induction heating process to nuclear power plants. Residual stress can be generated due to thermo-mechanical mechanism during the induction heating bending process. It is well-known that the residual stress has important effect on crack initiation and growth. The previous studies have focused on the thickness variation. In part, some studies were performed for residual stress evaluation of the austenitic stainless steel piping bended by induction heating. It is difficult to find the residual stresses of the ferritic steel piping bended by the induction heating. The study assessed the residual stresses of induction heating bended ferriticsteel piping via finite element analysis. As a result, it was identified that high residual stresses are generated on local outersurface region of the induction heating bended ferritic piping.

  15. Evaluation of examination techniques for ferritic stainless steel feedwater heater tubing

    SciTech Connect

    Nugent, M.J.; Catapano, M.C.

    1995-12-01

    Ferritic stainless steel has been finding increased application in utility plant feedwater heaters due to good strength and corrosion resistance and absence of potential copper contamination of feedwater system. Ferritic stainless steel is highly magnetic and is generally not inspectable using conventional eddy current testing techniques. A variety of techniques have been developed for inspection of this tubing material used in typical heat exchanger applications. Through a project funded by the Empire State Electric Energy Research Corporation (ESEERCO), the evaluation of data generated by four present state of the art NDE testing techniques were evaluated on a controlled mock-up of the heater tubing with service related defects. The primary objective was to determine the strengths and limitations of each method. The testing of two in service feedwater heaters at the Consolidated Edison Company of New York, Inc. (Con Edison`s) Arthur Kill Generating Station also allowed further evaluations based on actual field conditions.

  16. Microstructural evolution of delta ferrite in SAVE12 steel under heat treatment and short-term creep

    SciTech Connect

    Li, Shengzhi; Eliniyaz, Zumrat; Zhang, Lanting; Sun, Feng; Shen, Yinzhong; Shan, Aidang

    2012-11-15

    This research focused on the formation and microstructural evolution of delta ferrite phase in SAVE12 steel. The formation of delta ferrite was due to the high content of ferrite forming alloy elements such as Cr, W, and Ta. This was interpreted through either JMatPro-4.1 computer program or Cr{sub eq} calculations. Delta ferrite was found in bamboo-like shape and contained large amount of MX phase. It was surrounded by Laves phases before creep or aging treatment. Annealing treatments were performed under temperatures from 1050 Degree-Sign C to 1100 Degree-Sign C and various time periods to study its dissolution kinetics. The result showed that most of the delta ferrite can be dissolved by annealing in single phase austenitic region. Dissolution process of delta ferrite may largely depend on dissolution kinetic factors, rather than on thermodynamic factors. Precipitation behavior during short-term (1100 h) creep was investigated at temperature of 600 Degree-Sign C under a stress of 180 MPa. The results demonstrated that delta ferrite became preferential nucleation sites for Laves phase at the early stage of creep. Laves phase on the boundary around delta ferrite showed relatively slower growth and coarsening rate than that inside delta ferrite. - Highlights: Black-Right-Pointing-Pointer Delta ferrite is systematically studied under heat treatment and short-term creep. Black-Right-Pointing-Pointer Delta ferrite contains large number of MX phase and is surrounded by Laves phases before creep or aging treatment. Black-Right-Pointing-Pointer Formation of delta ferrite is interpreted by theoretical and empirical methods. Black-Right-Pointing-Pointer Most of the delta ferrite is dissolved by annealing in single phase austenitic region. Black-Right-Pointing-Pointer Delta ferrite becomes preferential nucleation sites for Laves phase at the early stage of creep.

  17. Response of ferritic steels to nonsteady loading at elevated temperatures

    SciTech Connect

    Swindeman, R.W.

    1984-04-01

    High-temperature operating experience is lacking in pressure vessel materials that have strength levels above 586 MPa. Because of their tendency toward strain softening, we have been concerned about their behavior under nonsteady loading. Testing was undertaken to explore the extent of softening produced by monotonic and cyclic strains. The specific materials included bainitic 2 1/4Cr-1Mo steel, a micro-alloyed version of 2 1/4Cr-1Mo steel, a micro-alloyed version of 2 1/4Cr-1Mo steel containing vanadium, titanium, and boron, and a martensitic 9Cr-1Mo-V-Nb steel. Tests included tensile, creep, variable stress creep, relaxation, strain cycling, stress cycling, and non-isothermal creep ratchetting experiments. We found that these steels had very low uniform elongation and exhibited small strains to the onset of tertiary creep compared to annealed 2 1/4Cr-1Mo steel. Repeated relaxation test data also indicated a limited capacity for strain hardening. Reversal strains produced softening. The degree of softening increased with increased initial strength level. We concluded that the high strength bainitic and martensitic steels should perform well when used under conditions where severe cyclic operation does not occur.

  18. Dynamic finite element modeling of the effects of size on the upper shelf energy of ferritic steels

    SciTech Connect

    Sidener, S.E.; Kumar, A.S.; Schubert, L.E.; Hamilton, M.L.; Rosinski, S.T.

    1996-04-01

    Both the fusion and light water reactor program require the use of the subsize specimens to obtain sufficient irradiation data on neutron-induced embrittlement of ferritic steels. While the development of fusion-relevant size effects correlations can proceed analytically, it is more cost-effective at this time to use data currently being obtained on embrittlement of pressure vessel steels to test and expand the correlations developed earlier using fusion relevant steels. Dynamic finite elements modeling of the fracture behavior of fatigue-precracked Charpy Specimens was performed to determine the effect of single variable changes in ligament size, width, span, and thickness on the upper shelf energy. A method based on tensile fracture strain was used for modeling crack initiation and propagation. It was found that the upper shelf energy of precracked specimens (USE{sub p}) is proportional to b{sup n}, where b is ligament size and n varies from about 1.6 for subsize to 1.9 for full size specimens. The USE{sub p} was found to be proportional to width according to W{sup 25}. The dependence on thickness was found to be linear for all cases studied. Some of the data from the FEM analysis were compared with experimental data and were found to be in reasonable agreement.

  19. Influence of Co content on the biocompatibility and bio-corrosion of super ferritic stainless steels

    NASA Astrophysics Data System (ADS)

    Yoo, Y. R.; Jang, S. G.; Nam, H. S.; Shim, G. T.; Cho, H. H.; Kim, J. G.; Kim, Y. S.

    2008-12-01

    Bio-metals require high corrosion resistance, because their biocompatibility is closely related to this parameter. Bio-metals release metal ions into the human body, leading to deleterious effects. Allergies, dermatitis, and asthma are the predominant systemic effects resulting in the human body. In particular, Ni is one of the most common causes of allergic contact dermatitis. In the present work, we designed new ferritic stainless steels wherein Ni is replaced with Co under consideration of allergic respondes and microstructural stability. This work focuses on the effect of Co content on the biocompatibility and corrosion resistance of high PRE super ferritic stainless steels in bio-solution and acidic chloride solution. In the case of the acidic chloride solution, with increasing Co content in the ferritic stainless steels, passive current density increased and critical pitting temperature (CPT) decreased. Also, in the passive state, AC impedance and repassivation rate were reduced. These results are attributed to the thermodynamic stability of cobalt ions, as indicated in the EpH diagram for a Co-H2O system. However, in the case of bio-solutions, with increasing Co content of the alloys, the passive current density decreased. AC impedance and repassivation rate meanwhile increased in the passive state. This is due to the increased ratios of Cr2O3/Cr(OH)3 and [Metal Oxide]/Metal + Metal Oxide] of the passive film formed in bio-solution.

  20. The influence of Cr content on the mechanical properties of ODS ferritic steels

    NASA Astrophysics Data System (ADS)

    Li, Shaofu; Zhou, Zhangjian; Jang, Jinsung; Wang, Man; Hu, Helong; Sun, Hongying; Zou, Lei; Zhang, Guangming; Zhang, Liwei

    2014-12-01

    The present investigation aimed at researching the mechanical properties of the oxide dispersion strengthened (ODS) ferritic steels with different Cr content, which were fabricated through a consolidation of mechanical alloyed (MA) powders of 0.35 wt.% nano Y2O3 dispersed Fe-12.0Cr-0.5Ti-1.0W (alloy A), Fe-16.0Cr-0.5Ti-1.0W (alloy B), and Fe-18.0Cr-0.5Ti-1.0W (alloy C) alloys (all in wt.%) by hot isostatic pressing (HIP) with 100 MPa pressure at 1150 °C for 3 h. The mechanical properties, including the tensile strength, hardness, and impact fracture toughness were tested by universal testers, while Young's modulus was determined by ultrasonic wave non-destructive tester. It was found that the relationship between Cr content and the strength of ODS ferritic steels was not a proportional relationship. However, too high a Cr content will cause the precipitation of Cr-enriched segregation phase, which is detrimental to the ductility of ODS ferritic steels.

  1. An equation-of-state for methane for modeling hydrogen attack in ferritic steels

    NASA Astrophysics Data System (ADS)

    Odette, G. R.; Vagarali, S. S.

    1982-02-01

    A statistical mechanical-based high temperature and high pressure equation-of-state for methane has been developed using the McQuarrie and Katz formulation based on Leonard-Jones (n, 6) intermolecular potential. Fugacity coefficients for methane have been estimated, and it is shown that for plain carbon steels during hydrogen attack the methane pressures are considerably lower than the fugacities and fall into a physically meaningful range (≤2500 MPa). Further, simple, but reasonably accurate, expressions for both the equation-of-state and fugacity coefficient have been developed for the purpose of modeling hydrogen attack kinetics in ferritic steels.

  2. Mixed-mode I/III fracture toughness of a ferritic/martensitic stainless steel

    SciTech Connect

    Li, Huaxin; Jones, R.H.; Gelles, D.S.; Hirth, J.P.

    1993-10-01

    The critical J-integrals of mode I (J{sub IC}), mixed-mode I/III (J{sub MC}), and mode III (J{sub IIIC}) were examined for a ferritic stainless steel (F-82H) at ambient temperature. A determination of J{sub MC} was made using modified compact-tension specimens. Different ratios of tension/shear stress were achieved by varying the principal axis of the crack plane between 0 and 55 degrees from the load line. Results showed that J{sub MC} and tearing modulus (T{sub M}) values varied with the crack angles and were lower than their mode I and mode III counterparts. Both the minimum J{sub MC} and T{sub M} values occurred at a crack angle between 40 and 50 degrees, where the load ratio of {sigma}{sub i}/{sigma}{sub iii} was 1.2 to 0.84. The J{sub min} was 240 Kj/M{sup 2}, and ratios of J{sub IC}/J{sub min} and J{sub IIIC}/J{sub min} were 2.1 and 1.9, respectively. The morphology of fracture surfaces was consistent with the change of J{sub MC} and T{sub M} values. While the upper shelf-fracture toughness of F-82H depends on loading mode, the J{sub min} remains very high. Other important considerations include the effect of mixed-mode loading on the DBT temperature, and effects of hydrogen and irradiation on J{sub min}.

  3. Tensile behavior of irradiated manganese-stabilized stainless steel

    SciTech Connect

    Klueh, R.L.

    1996-10-01

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

  4. Corrosion of Ferritic Steels in High Temperature Molten Salt Coolants for Nuclear Applications

    SciTech Connect

    Farmer, J; El-Dasher, B; de Caro, M S; Ferreira, J

    2008-11-25

    Corrosion of ferritic steels in high temperature molten fluoride salts may limit the life of advanced reactors, including some hybrid systems that are now under consideration. In some cases, the steel may be protected through galvanic coupling with other less noble materials with special neutronic properties such a beryllium. This paper reports the development of a model for predicting corrosion rates for various ferritic steels, with and without oxide dispersion strengthening, in FLiBe (Li{sub 2}BeF{sub 4}) and FLiNaK (Li-Na-K-F) coolants at temperatures up to 800 C. Mixed potential theory is used to account for the protection of steel by beryllium, Tafel kinetics are used to predict rates of dissolution as a function of temperature and potential, and the thinning of the mass-transfer boundary layer with increasing Reynolds number is accounted for with dimensionless correlations. The model also accounts for the deceleration of corrosion as the coolants become saturated with dissolved chromium and iron. This paper also reports electrochemical impedance spectroscopy of steels at their corrosion potentials in high-temperature molten salt environments, with the complex impedance spectra interpreted in terms of the interfacial charge transfer resistance and capacitance, as well as the electrolyte conductivity. Such in situ measurement techniques provide valuable insight into the degradation of materials under realistic conditions.

  5. Low cycle fatigue properties of a low activation ferritic steel (JLF-1) at room temperature

    NASA Astrophysics Data System (ADS)

    Nishimura, A.; Nagasaka, T.; Inoue, N.; Muroga, T.; Namba, C.

    2000-12-01

    To investigate fatigue properties of a low activation ferritic steel (9Cr-2W steel, JLF-1), low cycle fatigue tests were performed in air at room temperature under axial strain control for a complete push-pull condition. The strain rate was 0.4% s-1. Cyclic strain-hardening was observed within the initial 20 cycles, and then cyclic strain-softening occurred gradually until the final failure, though the plastic strain range did not change significantly. Tensile peak stresses in hysteresis curves measured at around half the number of cycles to failure depended on the total strain range. The drop in the peak stress by the cyclic strain-softening increased with decreasing total strain range. The regression curve of the total strain range against the fatigue life was formulated using the Manson-Coffin equation and the fatigue life of JLF-1 steel was compared with that of 8Cr-2W steel.

  6. TEM and HRTEM study of oxide particles in an Al-alloyed high-Cr oxide dispersion strengthened ferritic steel with Hf addition

    NASA Astrophysics Data System (ADS)

    Dou, Peng; Kimura, Akihiko; Kasada, Ryuta; Okuda, Takanari; Inoue, Masaki; Ukai, Shigeharu; Ohnuki, Somei; Fujisawa, Toshiharu; Abe, Fujio; Jiang, Shan; Yang, Zhigang

    2017-03-01

    The nanoparticles in an Al-alloyed high-Cr oxide dispersion strengthened (ODS) ferritic steel with Hf addition, i.e., SOC-16 (Fe-15Cr-2W-0.1Ti-4Al-0.62Hf-0.35Y2O3), have been examined by transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM). Relative to an Al-alloyed high-Cr ODS ferritic steel without Hf addition, i.e., SOC-9 (Fe-15.5Cr-2W-0.1Ti-4Al-0.35Y2O3), the dispersion morphology and coherency of the oxide nanoparticles in SOC-16 were significantly improved. Almost all the small nanoparticles (diameter <10 nm) in SOC-16 were found to be consistent with cubic Y2Hf2O7 oxides with the anion-deficient fluorite structure and coherent with the bcc steel matrix. The larger particles (diameter >10 nm) were also mainly identified as cubic Y2Hf2O7 oxides with the anion-deficient fluorite structure. The results presented here are compared with those of SOC-9 with a brief discussion of the underlying mechanisms of the unusual thermal and irradiation stabilities of the oxides as well as the superior strength, excellent irradiation tolerance and extraordinary corrosion resistance of SOC-16.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

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

  8. In-service inspection method for low-finned ferritic stainless steel tubes for new heat exchanger

    SciTech Connect

    Iwai, O.; Goto, M.

    1994-12-31

    Conventional inner eddy current test cannot obtain sufficient evaluation for low finned ferritic stainless steel tube inspection. The authors tried various methods and developed special partial saturation eddy current method. This paper summarizes typical experimental results of fundamental studies and trials, and introduces developed ECT data acquisition and evaluation system. Moisture Separator Heater (MSH) used in ABWR (Advanced Boiling Water Reactor) plant is a new type heat exchanger to increase plant thermal efficiency. There are four single tubesheet heaters in a MSH vessel. Each heater has hundreds of low finned tubes made of ferritic stainless steel. In nuclear power plants, non-magnetic materials (austenitic stainless steel, titanium, aluminum brass, etc.,) are mainly used as heat exchanger tubes such as the tubes of feedwater heater, condenser, evaporator and so on. Conventional ECT (Eddy Current Test) method are easily applied for the inspection of these heat exchanger tubes. In recent years, the authors started using ferritic stainless steel tube for new heat exchangers such as MSH because of its superior heat transfer efficiency. However, high permeability of ferritic stainless steel prevents the inspection of these tubes using conventional ECT method. To inspect MSH tubes periodically is important to confirm and maintain reliability of MSH. They tried applying various inspection methods and have developed special ECT method for low finned ferritic stainless steel tubes.

  9. Neodymium-rich precipitate phases in a high-chromium ferritic/martensitic steel

    NASA Astrophysics Data System (ADS)

    Shen, Yinzhong; Zhou, Xiaoling; Shang, Zhongxia

    2016-05-01

    Neodymium being considered as nitride forming element has been used in a design of advanced ferritic/martensitic (FM) steels for fossil fired power plants at service temperatures of 630 °C to 650 °C to effectively improve the creep strength of the steels. To fully understand the characteristics of neodymium precipitates in high-Cr FM steels, precipitate phases in an 11Cr FM steel with 0.03 wt% addition of Nd have been investigated by transmission electron microscopy. Three neodymium phases with a face-centered cubic crystal structure and different composition were observed in the steel. They consisted of neodymium carbonitride with an average lattice parameter of 1.0836 nm, Nd-rich carbonitride mainly containing Mn, and Nd-rich MN nitride mainly containing Mn and Co. Other three Nd-rich and Nd-containing phases, which appear to be Nd-Co-Cr/Nd-rich intermetallic compounds and Cr-Fe-rich nitride containing Nd, were also detected in the steel. Nd-relevant precipitates were found to be minor phases compared with M23C6 and Nb/V/Ta-rich MX phases in the steel. The content of Nd in other precipitate phases was very low. Most of added Nd is considered to be present as solid solution in the matrix of the steel.

  10. Stability of nanosized oxides in ferrite under extremely high dose self ion irradiations

    NASA Astrophysics Data System (ADS)

    Aydogan, E.; Almirall, N.; Odette, G. R.; Maloy, S. A.; Anderoglu, O.; Shao, L.; Gigax, J. G.; Price, L.; Chen, D.; Chen, T.; Garner, F. A.; Wu, Y.; Wells, P.; Lewandowski, J. J.; Hoelzer, D. T.

    2017-04-01

    A nanostructured ferritic alloy (NFA), 14YWT, was produced in the form of thin walled tubing. The stability of the nano-oxides (NOs) was determined under 3.5 MeV Fe+2 irradiations up to a dose of ∼585 dpa at 450 °C. Transmission electron microscopy (TEM) and atom probe tomography (APT) show that severe ion irradiation results in a ∼25% reduction in size between the unirradiated and irradiated case at 270 dpa while no further reduction within the experimental error was seen at higher doses. Conversely, number density increased by ∼30% after irradiation. This 'inverse coarsening' can be rationalized by the competition between radiation driven ballistic dissolution and diffusional NO reformation. No significant changes in the composition of the matrix or NOs were observed after irradiation. Modeling the experimental results also indicated a dissolution of the particles.

  11. Stability of nanosized oxides in ferrite under extremely high dose self ion irradiations

    DOE PAGES

    Aydogan, E.; Almirall, N.; Odette, G. R.; ...

    2017-01-10

    We produced a nanostructured ferritic alloy (NFA), 14YWT, in the form of thin walled tubing. The stability of the nano-oxides (NOs) was determined under 3.5 MeV Fe+2 irradiations up to a dose of ~585 dpa at 450 °C. Transmission electron microscopy (TEM) and atom probe tomography (APT) show that severe ion irradiation results in a ~25% reduction in size between the unirradiated and irradiated case at 270 dpa while no further reduction within the experimental error was seen at higher doses. Conversely, number density increased by ~30% after irradiation. Moreover, this ‘inverse coarsening’ can be rationalized by the competition betweenmore » radiation driven ballistic dissolution and diffusional NO reformation. There were no significant changes in the composition of the matrix or NOs observed after irradiation. Modeling the experimental results also indicated a dissolution of the particles.« less

  12. Response of neutron-irradiated RPV steels to thermal annealing

    SciTech Connect

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1997-03-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels.

  13. Development oxide dispersion strengthened ferritic steels for fusion

    SciTech Connect

    Mukhopadhyay, D.K.; Froes, F.H.; Gelles, D.S.

    1997-04-01

    Uniaxial tension creep response is reported for an oxide dispersion strengthened (ODS) steel, Fe-13.5Cr-2W-0.5Ti-0.25 Y{sub 2}O{sub 3} (in weight percent) manufactured using the mechanical alloying process. Acceptable creep response is obtained at 900{degrees}C.

  14. Characterization of ferritic G. M. A. weld deposits in 9% Ni steel for cryogenic applications

    SciTech Connect

    Mahin, K.W.

    1980-04-01

    Low temperature containment vessels of 9% Ni are normally fabricated using the shielded metal arc (S.M.A.W.) or the gas metal arc (G.M.A.W.) welding processes. Available filler metals compatible with these processes are highly alloyed austenitics, whose strength levels undermatch those of the base plate. A more efficient weld joint would be a low alloy ferritic deposit. Although acceptable matching ferritic gas tungsten arc weld (G.T.A.W.) wires have been developed, similar progress has not been made in the area of ferritic G.M.A. weld wires. Most of the prior work in this area has focused on correlating composition with mechanical properties, without a corresponding evaluation of resultant microstructure. The study presented focused on establishing correlations between chemistry, microstructure and mechanical properties for four different ferritic G.M.A. weld deposits in 9% Ni steel, with the purpose of developing a better understanding of the factors controlling the 77K (-196/sup 0/C) toughness behavior of these weld metals. Microstructural characterization was carried out using standard optical and scanning electron microscopes, as well as a variety of advanced analytical techniques, including transmission electron microscopy (T.E.M.), scanning T.E.M., Moessbauer spectroscopy and Auger electron spectroscopy.

  15. Thermal Growth and Performance of Manganese Cobaltite Spinel Protection Layers on Ferritic Stainless Steel SOFC Interconnects

    SciTech Connect

    Yang, Zhenguo; Xia, Guanguang; Simner, Steven P.; Stevenson, Jeffry W.

    2005-08-01

    To protect solid oxide fuel cells (SOFCs) from chromium poisoning and improve metallic interconnect stability, manganese cobaltite spinel protection layers with a nominal composition of Mn1.5Co1.5O4 were thermally grown on Crofer22 APU, a ferritic stainless steel. Thermal, electrical and electrochemical investigations indicated that the spinel protection layers not only significantly decreased the contact area specific resistance (ASR) between a LSF cathode and the stainless steel interconnect, but also inhibited the sub-scale growth on the stainless steel by acting as a barrier to the inward diffusion of oxygen. A long-term thermal cycling test demonstrated excellent structural and thermomechanical stability of these spinel protection layers, which also acted as a barrier to outward chromium cation diffusion to the interconnect surface. The reduction in the contact ASR and prevention of Cr migration achieved by application of the spinel protection layers on ferritic stainless steel resulted in improved stability and electrochemical performance of SOFCs.

  16. Report on thermal aging effects on tensile properties of ferritic-martensitic steels.

    SciTech Connect

    Li, M.; Soppet, W.K.; Rink, D.L.; Listwan, J.T.; Natesan, K.

    2012-05-10

    This report provides an update on the evaluation of thermal-aging induced degradation of tensile properties of advanced ferritic-martensitic steels. The report is the first deliverable (level 3) in FY11 (M3A11AN04030103), under the Work Package A-11AN040301, 'Advanced Alloy Testing' performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing tensile data on aged alloys and a mechanistic model, validated by experiments, with a predictive capability on long-term performance. The scope of work is to evaluate the effect of thermal aging on the tensile properties of advanced alloys such as ferritic-martensitic steels, mod.9Cr-1Mo, NF616, and advanced austenitic stainless steel, HT-UPS. The aging experiments have been conducted over a temperature of 550-750 C for various time periods to simulate the microstructural changes in the alloys as a function of time at temperature. In addition, a mechanistic model based on thermodynamics and kinetics has been used to address the changes in microstructure of the alloys as a function of time and temperature, which is developed in the companion work package at ANL. The focus of this project is advanced alloy testing and understanding the effects of long-term thermal aging on the tensile properties. Advanced materials examined in this project include ferritic-martensitic steels mod.9Cr-1Mo and NF616, and austenitic steel, HT-UPS. The report summarizes the tensile testing results of thermally-aged mod.9Cr-1Mo, NF616 H1 and NF616 H2 ferritic-martensitic steels. NF616 H1 and NF616 H2 experienced different thermal-mechanical treatments before thermal aging experiments. NF616 H1 was normalized and tempered, and NF616 H2 was normalized and tempered and cold-rolled. By examining these two heats, we evaluated the effects of thermal-mechanical treatments on material microstructures and

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

    SciTech Connect

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

    1996-04-01

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

  18. A correlative approach to segmenting phases and ferrite morphologies in transformation-induced plasticity steel using electron back-scattering diffraction and energy dispersive X-ray spectroscopy.

    PubMed

    Gazder, Azdiar A; Al-Harbi, Fayez; Spanke, Hendrik Th; Mitchell, David R G; Pereloma, Elena V

    2014-12-01

    Using a combination of electron back-scattering diffraction and energy dispersive X-ray spectroscopy data, a segmentation procedure was developed to comprehensively distinguish austenite, martensite, polygonal ferrite, ferrite in granular bainite and bainitic ferrite laths in a thermo-mechanically processed low-Si, high-Al transformation-induced plasticity steel. The efficacy of the ferrite morphologies segmentation procedure was verified by transmission electron microscopy. The variation in carbon content between the ferrite in granular bainite and bainitic ferrite laths was explained on the basis of carbon partitioning during their growth.

  19. Microstructure of a 14Cr-ODS ferritic steel before and after helium ion implantation

    NASA Astrophysics Data System (ADS)

    Lu, Chenyang; Lu, Zheng; Xie, Rui; Liu, Chunming; Wang, Lumin

    2014-12-01

    A 14Cr-ODS ferritic steel with the nominal compositions of Fe-14Cr-2 W-0.3Ti-0.3Y2O3 (wt.%) was produced by mechanical alloying (MA) and hot isostatic pressing (HIP). Helium ion was implanted into the 14Cr-ODS steel along with Eurofer 97 steel as reference at 400 °C to a fluence of 1 × 1017 He+/cm2. High resolution transmission electron microscopy (HRTEM), high angle annual dark field (HAADF) scanning TEM (STEM) and atom probe tomography (APT) were used to characterize the microstructure of 14Cr-ODS and Eurofer 97 steels before and after helium implantation. High-density Y-Ti-O-rich nanoclusters and Y2Ti2O7 precipitates as well as large Cr-Ti rich oxides were observed in the 14Cr-ODS steel. The average size of Y-Ti-O nanoclusters and Y2Ti2O7 precipitates is 9 nm. After helium implantation, the helium bubbles formed in the 14Cr-ODS steel exhibit the smaller size and the lower volume fraction than that in Eurofer 97 steel, indicating high-density nano-scale precipitates can effectively suppress the coarsening of helium bubbles.

  20. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  1. Neutron diffraction investigation of lattice microstrain in ferrite steel

    NASA Astrophysics Data System (ADS)

    Camanzi, A.; Moze, O.

    1992-06-01

    The degradation of carbon steels when exposed to H rich environments is well known to result in catastrophic failure. In order to characterize in a comprehensive manner the structural effects of hydrogenation, a series of high resolution neutron powder diffraction measurements were carried out on cross-sections of carbon steel segments used for gas pipelines. Peak positions were measured to an accuracy of 0.001%, whilst line broadening of individual peaks was measured to an accuracy of 0.1%. The ( h k l) dependent peak linewidths were fitted using a pseudo-Voigt peak shape function. Non-hydrogenated materials were found to display a different diffraction linewidth dependence on the crystal elastic anisotropy than hydrogenated materials.

  2. Wrought Cr--W--V bainitic/ferritic steel compositions

    DOEpatents

    Klueh, Ronald L.; Maziasz, Philip J.; Sikka, Vinod Kumar; Santella, Michael L.; Babu, Sudarsanam Suresh; Jawad, Maan H.

    2006-07-11

    A high-strength, high-toughness steel alloy includes, generally, about 2.5% to about 4% chromium, about 1.5% to about 3.5% tungsten, about 0.1% to about 0.5% vanadium, and about 0.05% to 0.25% carbon with the balance iron, wherein the percentages are by total weight of the composition, wherein the alloy is heated to an austenitizing temperature and then cooled to produce an austenite transformation product.

  3. Formable ferrite-degenerated pearlite steel (FDP-55) for automotive use

    SciTech Connect

    Nagao, N.; Hamamatsu, S.; Kunishige, K.

    1984-01-01

    In order to help the gauge reduction of wheels and chassis parts of automobiles, a formable and weldable hot rolled steel of 550 MPa grade, named FDP-55, has been developed. FDP-55 is an 0.14% C, 0.1% Si, 1.1% Mn and Nb free Alkilled steel obtained by controlled-cooling to a low coiling temperature on a runout table, and it is featured by ferrite-degenerated pearlite microstructure. Results of co-operative works with automotive makers showed that FDP-55 was successful in the application to wheels and chassis parts attaining the large weight reduction. This paper reports the metallurgical features and characteristics of the steel.

  4. High-Temperature Performance of Ferritic Steels in Fireside Corrosion Regimes: Temperature and Deposits

    NASA Astrophysics Data System (ADS)

    Dudziak, T.; Hussain, T.; Simms, N. J.

    2016-11-01

    The paper reports high temperature resistance of ferritic steels in fireside corrosion regime in terms of temperature and deposits aggressiveness. Four candidate power plant steels: 15Mo3, T22, T23 and T91 were exposed under simulated air-fired combustion environment for 1000 h. The tests were conducted at 600, 650 and 700 °C according to deposit-recoat test method. Post-exposed samples were examined via dimensional metrology (the main route to quantify metal loss), and mass change data were recorded to perform the study of kinetic behavior at elevated temperatures. Microstructural investigations using ESEM-EDX were performed in order to investigate corrosion degradation and thickness of the scales. The ranking of the steels from most to the least damage was 15Mo3 > T22 > T23 > T91 in all three temperatures. The highest rate of corrosion in all temperatures occurred under the screening deposit.

  5. Structure of Oxide Nanoparticles in Fe-16Cr MA/ODS Ferritic Steel

    SciTech Connect

    Hsiung, L; Fluss, M; Kimura, A

    2010-04-06

    Oxide nanoparticles in Fe-16Cr ODS ferritic steel fabricated by mechanical alloying (MA) method have been examined using high-resolution transmission electron microscopy (HRTEM) techniques. A partial crystallization of oxide nanoparticles was frequently observed in as-fabricated ODS steel. The crystal structure of crystalline oxide particles is identified to be mainly Y{sub 4}Al{sub 2}O{sub 9} (YAM) with a monoclinic structure. Large nanoparticles with a diameter larger than 20 nm tend to be incoherent and have a nearly spherical shape, whereas small nanoparticles with a diameter smaller than 10 nm tend to be coherent or semi-coherent and have faceted boundaries. The oxide nanoparticles become fully crystallized after prolonged annealing at 900 C. These results lead us to propose a three-stage formation mechanism of oxide nanoparticles in MA/ODS steels.

  6. High-Temperature Performance of Ferritic Steels in Fireside Corrosion Regimes: Temperature and Deposits

    NASA Astrophysics Data System (ADS)

    Dudziak, T.; Hussain, T.; Simms, N. J.

    2017-01-01

    The paper reports high temperature resistance of ferritic steels in fireside corrosion regime in terms of temperature and deposits aggressiveness. Four candidate power plant steels: 15Mo3, T22, T23 and T91 were exposed under simulated air-fired combustion environment for 1000 h. The tests were conducted at 600, 650 and 700 °C according to deposit-recoat test method. Post-exposed samples were examined via dimensional metrology (the main route to quantify metal loss), and mass change data were recorded to perform the study of kinetic behavior at elevated temperatures. Microstructural investigations using ESEM-EDX were performed in order to investigate corrosion degradation and thickness of the scales. The ranking of the steels from most to the least damage was 15Mo3 > T22 > T23 > T91 in all three temperatures. The highest rate of corrosion in all temperatures occurred under the screening deposit.

  7. Kinetic transition during the growth of proeutectoid ferrite in Fe-C-Mn-Si quaternary steel

    NASA Astrophysics Data System (ADS)

    Zhang, Guo-Hong; Heo, Yoon-Uk; Song, Eun-Ju; Suh, Dong-Woo

    2013-03-01

    The kinetics of ferrite growth in Fe-0.1C-1.5Mn-0.94Si (mass pct) quaternary steel is investigated through the characterization of isothermal growth behavior, the thermodynamic prediction of kinetic boundary and the diffusional growth simulations using DICTRA. The change in microstructural evolution from slow growth to fast one is consistent with the calculated change of interface condition from the partitioning local equilibrium (PLE) to the negligible partitioning local equilibrium (NPLE). Compared with the DICTRA simulation, the observed growth kinetics of ferrite are between the calculated ones assuming local equilibrium (LE) and paraequilibrium (PE) criterions. At temperatures below the PLE/NPLE kinetic boundary, the observed growth behavior can be reasonably described by kinetic transition from PE to NPLE condition as isothermal time elapses, taking into account the critical velocity of interface at which trans-interface diffusion of subsitutional element permits the transition from PE to NPLE growth.

  8. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    SciTech Connect

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L.; Matsui, H.

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  9. Alloying effect of Ni and Cr on irradiated microstructural evolution of type 304 stainless steels

    NASA Astrophysics Data System (ADS)

    Tan, L.; Busby, J. T.

    2013-11-01

    Life extension of the existing nuclear power plants imposes significant challenges to core structural materials that suffer increased fluences. This paper presents the microstructural evolution of a type 304 stainless steel and its variants alloyed with extra Ni and Cr under neutron irradiation at ˜320 °C for up to 10.2 dpa. Similar to the reported data of type 304 variants, a large amount of Frank loops, ultrafine G-phase/M23C6 particles, and limited amount of cavities were observed in the irradiated samples. The irradiation promoted the growth of pre-existing M23C6 at grain boundaries and resulted in some phase transformation to CrC in the alloy with both extra Ni and Cr. A new type of ultrafine precipitates, possibly (Ti,Cr)N, was observed in all the samples, and its amount was increased by the irradiation. Additionally, α-ferrite was observed in the type 304 steel but not in the Ni or Ni + Cr alloyed variants. The effect of Ni and Cr alloying on the microstructural evolution is discussed.

  10. Corrosion and toughness of experimental and commercial super ferritic stainless steels

    SciTech Connect

    Dowling, N.J.E.; Kim, H.; Kim, J.N.; Ahn, S.K.; Lee, Y.D.

    1999-08-01

    The effect of minor alloying in a super ferritic stainless steel 26% Cr-3% Mo matrix was investigated. The corrosion resistance of several experimental heats was examined in terms of their sensitization and intergranular corrosion (IGA) susceptibility following a low-temperature anneal (620 C). Constant potential etching, electrochemical (electropotentiokinetic reactivation [EPR]), and immersion (modified Strauss test) studies showed that the principal corrosion initiation site of heat-treated steels was intergranular even at low [C + N] (< 130 ppm) and relatively high [Nb + Tl] concentrations. Despite the significant contribution of Nb and Ti to the IGA resistance, these elements had a deleterious effect on the notch toughness. Charpy V testing demonstrated increases in the upper shelf energy and ductile-to-brittle transition temperature (DBTT), the latter shifting from {approx} {minus}50 C in the absence of stabilizing elements to > 25 C for the stabilizing ratio [(Nb + Ti)/(C + N)] > 9. Corrosion resistance of the experimental heats was compared with that of several commercial alloys with intermediate stabilization ratios. The interaction of toughness and corrosion resistance in ferritic stainless steel was discussed with respect to the lack of consistency between published evaluation methods and the ideal stabilization ratio at low [C + N] values.

  11. Development of new generation reduced activation ferritic-martensitic steels for advanced fusion reactors

    NASA Astrophysics Data System (ADS)

    Tan, L.; Snead, L. L.; Katoh, Y.

    2016-09-01

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ∼500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimental results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. The strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9-20Cr oxide dispersion-strengthened ferritic alloys.

  12. Surface Treatments for Improved Performance of Spinel-coated AISI 441 Ferritic Stainless Steel

    SciTech Connect

    Stevenson, Jeffry W.; Riel, Eric M.; Stephens, Elizabeth V.; Khaleel, Mohammad A.

    2013-01-01

    Ferritic stainless steels are promising candidates for IT-SOFC interconnect applications due to their low cost and resistance to oxidation at SOFC operating temperatures. However, steel candidates face several challenges; including long term oxidation under interconnect exposure conditions, which can lead to increased electrical resistance, surface instability, and poisoning of cathodes due to volatilization of Cr. To potentially extend interconnect lifetime and improve performance, a variety of surface treatments were performed on AISI 441 ferritic stainless steel coupons prior to application of a protective spinel coating. The coated coupons were then subjected to oxidation testing at 800 and 850°C in air, and electrical testing at 800°C in air. While all of the surface-treatments resulted in improved surface stability (i.e., increased spallation resistance) compared to untreated AISI 441, the greatest degree of improvement (through 20,000 hours of testing at 800°C and 14,000 hours of testing at 850°C) was achieved by surface blasting.

  13. Precipitation and mechanical properties of Nb-modified ferritic stainless steel during isothermal aging

    SciTech Connect

    Yan Haitao Bi Hongyun; Li Xin; Xu Zhou

    2009-03-15

    The influence of isothermal aging on precipitation behavior and mechanical properties of Nb-modified ferritic stainless steel was investigated using Thermo-calc software, scanning electron microscopy and transmission electron microscopy. It was observed that TiN, NbC and Fe{sub 2}Nb formed in the investigated steel and the experimental results agreed well with the results calculated by Thermo-calc software. During isothermal aging at 800 deg. C, the coarsening rate of Fe{sub 2}Nb is greater than that of NbC, and the calculated average sizes of NbC and Fe{sub 2}Nb of the aged specimen agreed with the experimental results. In addition, the tensile strength and micro-hardness of the ferritic stainless steel increased with increased aging time from 24 h to 48 h. But aging at 800 deg. C for 96 h caused the coarsening of the precipitation, which led to a decrease of tensile strength and micro-hardness.

  14. Development of new generation reduced activation ferritic-martenstic steels for advanced fusion reactors

    SciTech Connect

    Tan, Lizhen; Snead, Lance Lewis; Katoh, Yutai

    2016-05-26

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ~500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimental results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. Furthermore, the strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9–20Cr oxide dispersion-strengthened ferritic alloys.

  15. Development of new generation reduced activation ferritic-martenstic steels for advanced fusion reactors

    DOE PAGES

    Tan, Lizhen; Snead, Lance Lewis; Katoh, Yutai

    2016-05-26

    International development of reduced activation ferritic-martensitic (RAFM) steels has focused on 9 wt percentage Cr, which primarily contain M23C6 (M = Cr-rich) and small amounts of MX (M = Ta/V, X = C/N) precipitates, not adequate to maintain strength and creep resistance above ~500 °C. To enable applications at higher temperatures for better thermal efficiency of fusion reactors, computational alloy thermodynamics coupled with strength modeling have been employed to explore a new generation RAFM steels. The new alloys are designed to significantly increase the amount of MX nanoprecipitates, which are manufacturable through standard and scalable industrial steelmaking methods. Preliminary experimentalmore » results of the developed new alloys demonstrated noticeably increased amount of MX, favoring significantly improved strength, creep resistance, and Charpy impact toughness as compared to current RAFM steels. Furthermore, the strength and creep resistance were comparable or approaching to the lower bound of, but impact toughness was noticeably superior to 9–20Cr oxide dispersion-strengthened ferritic alloys.« less

  16. Effect of Hot Band Annealing on Forming Limit Diagrams of Ultra-Pure Ferritic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Shu, Jun; Bi, Hongyun; Li, Xin; Xu, Zhou

    2014-03-01

    In order to better understand the texture evolution, coincidence site lattice (CSL) and forming limit diagrams (FLDs) of ferritic stainless steels with and without hot band annealing, the texture evolution and CSL of ferritic stainless steels with 15% Cr content were studied by using x-ray diffraction and electron back-scattered diffraction technique. The strain hardening exponent n value, the strength coefficient K value, and Plastic strain ratio r value are the key parameters for the FLD. It was found out that the FLDo of plane strain condition and the stretchability were mainly influenced by their n value and K value, respectively. The higher n value and K value, better was the stretchability of investigated steels. The intensity of the γ-fiber dominated by {111} <112> was improved significantly in the cold rolled and annealed sheets because of a hot band annealing treatment and the sharp increase of Σ13b CSL boundaries. The increase of the formability is attributed to the significantly increase of the r value.

  17. Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging

    NASA Astrophysics Data System (ADS)

    Kumar, Deepak; Prakash, Ujjwal; Dabhade, Vikram V.; Laha, K.; Sakthivel, T.

    2017-02-01

    Oxide dispersion strengthened (ODS) ferritic steels are candidates for cladding tubes in fast breeder nuclear reactors. In this study, an 18%Cr ODS ferritic steel was prepared through powder forging route. Elemental powders with a nominal composition of Fe-18Cr-2 W-0.2Ti (composition in wt.%) with 0 and 0.35% yttria were prepared by mechanical alloying in a Simoloyer attritor under argon atmosphere. The alloyed powders were heated in a mild steel can to 1473 K under flowing hydrogen atmosphere. The can was then hot forged. Steps of sealing, degassing and evacuation are eliminated by using powder forging. Heating ODS powder in hydrogen atmosphere ensures good bonding between alloy powders. A dense ODS alloy with an attractive combination of strength and ductility was obtained after re-forging. On testing at 973 K, a loss in ductility was observed in yttria-containing alloy. The strength and ductility increased with increase in strain rate at 973 K. Reasons for this are discussed. The ODS alloy exhibited a recrystallized microstructure which is difficult to achieve by extrusion. No prior particle boundaries were observed after forging. The forged compacts exhibited isotropic mechanical properties. It is suggested that powder forging may offer several advantages over the traditional extrusion/HIP routes for fabrication of ODS alloys.

  18. Study of martensitic-ferritic dual phase steels produced by hot stamping

    NASA Astrophysics Data System (ADS)

    Erişir, E.; Bilir, O. G.

    2017-02-01

    The effects of heat treatment and initial microstructure on tensile properties of 22MnB5 and 30MnB5 high-strength hot stamping steels with martensite-ferrite matrix were investigated. Hot stamping steels possessed limited elongations of about 5% in a tensile strength ranging from 1300 to 1500 MPa when quenched at temperatures above A3 temperatures. The total elongations were tried to improve by partial austenization between Ac1 and Ac3 temperature and quenching. Ac1 and Ac3 temperatures were calculated via ThermoCalc. Microstructural characterization was made by using Light Microscope and Scanning Electron Microscope. Microstructure is composed of ferrite+martensite. It was seen that annealing temperature affects the volume fraction of phases. It was concluded that initial microstructure is an important parameter for the final microstructure. This method can be used for automobile parts which require higher TE with sufficient yield and tensile strength. Also this process may be a way of using Zn coated steel sheets in hot stamping process.

  19. Response of nanostructured ferritic alloys to high-dose heavy ion irradiation

    SciTech Connect

    Parish, Chad M.; White, Ryan M.; LeBeau, James M.; Miller, Michael K.

    2014-02-01

    A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at -100°C and 750°C, as well as pre-irradiation reference material. Irradiation at -100°C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti-Y-O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750°C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr-W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

  20. Characterizing microstructural changes in ferritic steels by positron annihilation spectroscopy: Studies on modified 9Cr-1Mo steel

    NASA Astrophysics Data System (ADS)

    Hari Babu, S.; Rajkumar, K. V.; Hussain, S.; Amarendra, G.; Sundar, C. S.; Jayakumar, T.

    2013-01-01

    Applicability of positron annihilation spectroscopy in probing the microstructural changes in ferritic steels has been investigated with thermal treatment studies on modified 9Cr-1Mo steel, during 300-1273 K. Positron lifetime results are compared with those of ultrasonic velocity and hardness techniques with two initial microstructural conditions i.e., normalized and tempered condition as well as only normalized condition. In first case, positron lifetime is found to be sensitive to small changes in metal carbide precipitation which could not be probed by other two techniques. In later case, positron lifetime is found to be sensitive to defect annealing until 673 K and in distinguishing the growth and coarsening of metal carbide precipitation stages during 773-1073 K. The present study suggests that by combining positron lifetime, ultrasonic velocity and hardness measurements, it is possible to distinguish distinct microstructures occurring at different stages.

  1. Morphology, structure, and chemistry of nanoclusters in a mechanically alloyed nanostructured ferritic steel

    SciTech Connect

    Brandes, M. C.; Kovarik, Libor; Miller, Michael K.; Mills, M. J.

    2012-01-14

    Nanostructured ferritic steels have excellent high temperature creep properties and radiation tolerance due to the presence of a high density of Ti-Y-O-enriched nanoclusters. The morphology of the nanoclusters is found to be consistent with a truncated rhombic dodecahedron defined by the {l_brace}100{r_brace} and {l_brace}110{r_brace} planes in the Fe matrix. The derived symmetry and the compositional information indicate that the nanoclusters are inconsistent with the cubic Y2Ti2O7 or the polymorphs of Y2TiO5 phase. Possible structural models are discussed.

  2. Tensile properties and deformation mechanisms of a 14Cr ODS ferritic steel

    NASA Astrophysics Data System (ADS)

    Steckmeyer, A.; Praud, M.; Fournier, B.; Malaplate, J.; Garnier, J.; Béchade, J. L.; Tournié, I.; Tancray, A.; Bougault, A.; Bonnaillie, P.

    2010-10-01

    The search for a new cladding material is part of the research studies carried out at CEA to develop a sodium-cooled fast reactor meeting the expectations of the Generation IV International Forum. In this study, the tensile properties of a ferritic oxide dispersion strengthened steel produced by hot extrusion at CEA have been evaluated. They prove the studied alloy to be as resistant as and more ductile than the other nano-reinforced alloys of literature. The effects of the strain rate and temperature on the total plastic strain of the material remind of diffusion phenomena. Intergranular damage and intergranular decohesion are clearly highlighted.

  3. Corrosion of ferritic-martensitic steels and nickel-based alloys in supercritical water

    NASA Astrophysics Data System (ADS)

    Ren, Xiaowei

    The corrosion behavior of ferritic/martensitic (F/M) steels and Ni-based alloys in supercritical water (SCW) has been studied due to their potential applications in future nuclear reactor systems, fossil fuel power plants and waste treatment processes. 9˜12% chromium ferritic/martensitic steels exhibit good radiation resistance and stress corrosion cracking resistance. Ni-based alloys with an austenitic face-centered cubic (FCC) structure are designed to retain good mechanical strength and corrosion/oxidation resistance at elevated temperatures. Corrosion tests were carried out at three temperatures, 360°C, 500°C and 600°C, with two dissolved oxygen contents, 25 ppb and 2 ppm for up to 3000 hours. Alloys modified by grain refinement and reactive element addition were also investigated to determine their ability to improve the corrosion resistance in SCW. A duplex oxide structure was observed in the F/M steels after exposure to 25 ppb oxygen SCW, including an outer oxide layer with columnar magnetite grains and an inner oxide layer constituted of a mixture of spinel and ferrite phases in an equiaxed grain structure. An additional outermost hematite layer formed in the SCW-exposed samples when the oxygen content was increased to 2 ppm. Weight gain in the F/M steels increased with exposure temperatures and times, and followed parabolic growth kinetics in most of the samples. In Ni-based alloys after exposure to SCW, general corrosion and pitting corrosion were observed, and intergranular corrosion was found when exposed at 600°C due to formation of a local healing layer. The general oxide structure on the Ni-based alloys was characterized as NiO/Spinel/(CrxFe 1-x)2O3/(Fe,Ni). No change in oxidation mechanism was observed in crossing the critical point despite the large change in water properties. Corrosion resistance of the F/M steels was significantly improved by plasma-based yttrium surface treatment because of restrained outward diffusion of iron by the

  4. Development of oxide dispersion strengthened ferritic steels for fusion reactors

    SciTech Connect

    Mukhopadhay, D.K.; Suryanarayana, C.; Froes, F.H.; Hebeisen, J.; Gelles, D.S.

    1996-12-31

    Seven ODS steels, Fe-(5.13.5)Cr-2W-0.5Ti-0.25 Y{sub 2}O{sub 3} (in weight percent) were manufactured using the mechanical alloying process. Only the composition Fe-13.5Cr-2W-0.5Ti-0.25Y{sub 2}O{sub 3} showed no austenite formation at any temperature using differential thermal analysis and hence was selected as an experimental alloy for the present investigation. Milled powders were consolidated by hot isostatic pressing and hot swaging. Electron microscopy studies indicated high material homogeneity. The hardness of the as-swaged specimen was 65 R{sub c}. Annealing of the as-swaged material at 800 C, 900 C, 1,000 C, 1,100 C and 1,200 C showed a minor decrease in the hardness.

  5. Development of oxide dispersion strengthened ferritic steels for fusion

    SciTech Connect

    Mukhopadhyay, D.K.; Suryanarayana, C.; Froes, F.H.; Gelles, D.S.

    1996-04-01

    Seven ODS steels, Fe(5-13.5)Cr-2W-0.5Ti-0.25 Y{sub 2}O{sub 3} (in weight percent) were manufactured using the mechanical alloying process. Only the composition Fe-13.5Cr3W-0.5Ti-0.25Y{sub 2}O{sub 3} showed no austenite formation at any temperature using differential thermal analysis and hence was selected as an experimental alloy for the present investigation. Milled powders were consolidated by hot isostatic pressing and hot swaging. Electron microscopy studies indicated high material homogeneity. The hardness of the as-swaged specimen was 65 R{sub c}. Annealing of the as-swaged material at 800, 900, 1000, 1100, and 1200{degrees}C showed a minor decrease in the hardness.

  6. Triple Ion-Beam Studies of Radiation Damage in 9Cr2WVTa Ferritic/Martensitic Steel for a High Power Spallation Neutron Source

    SciTech Connect

    Lee, EH

    2001-08-01

    To simulate radiation damage under a future Spallation Neutron Source (SNS) environment, irradiation experiments were conducted on a candidate 9Cr-2WVTa ferritic/martensitic steel using the Triple Ion Facility (TIF) at ORNL. Irradiation was conducted in single, dual, and triple ion beam modes using 3.5 MeV Fe{sup 2}, 360 keV He{sup +}, and 180 keV H{sup +} at 80, 200, and 350 C. These irradiations produced various defects comprising black dots, dislocation loops, line dislocations, and gas bubbles, which led to hardening. The largest increase in hardness, over 63%, was observed after 50 dpa for triple beam irradiation conditions, revealing that both He and H are augmenting the hardening. Hardness increased less than 30% after 30 dpa at 200 C by triple beams, compatible with neutron irradiation data from previous work which showed about a 30% increase in yield strength after 27.2 dpa at 365 C. However, the very large concentrations of gas bubbles in the matrix and on lath and grain boundaries after these simulated SNS irradiations make predictions of fracture behavior from fission reactor irradiations to spallation target conditions inadvisable.

  7. Synergy effects of Cu and Sn on pitting corrosion resistance of ultra-purified medium chromium ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Zhang, XiangJun; Liu, ZhenYu

    2017-03-01

    The influence of combination of Cu and Sn on pitting resistance of ultra-purified medium chromium ferritic stainless steel in 3.5 wt.% NaCl at 25°C was investigated by using electrochemical method. The results show that there is synergy effect between Cu and Sn, and the strong interaction between Cu and Sn in ferritic stainless steels clearly affects their pitting corrosion behaviour in 3.5% NaCl. A mechanism of the synergy of Cu and Sn was discussed.

  8. Development of (Mn,Co)3O4 Protection Layers for Ferritic Stainless Steel Interconnects

    SciTech Connect

    Yang, Zhenguo; Simner, Steven P.; Singh, Prabhakar; Xia, Guanguang; Stevenson, Jeffry W.

    2005-07-05

    A spinel-based surface protection layer has been developed for alloy SOFC current collectors and bi-polar gas separators. The (Mn,Co)3O4 spinel with a nominal composition of Mn1.5Co1.5O4 demonstrates an excellent electrical conductivity and thermal expansion match to ferritic stainless steel interconnects. A slurry-coating technique provides a viable approach for fabricating protective layers of the spinel onto the steel interconnects. Thermally grown protection layers of Mn1.5Co1.5O4 have been found not only to significantly decrease the contact resistance between a LSF cathode and stainless steel interconnect, but also inhibits the sub-scale growth on the stainless steel. The combination of the inhibited sub-scale growth, good thermal expansion matching between the spinel and the stainless steel, and the closed-pore structure contribute to the excellent structural and thermomechanical stability of these spinel protection layers, which was verified by a long-term thermal-cycling test. The spinel protection layers can also act effectively to prevent outward diffusion of chromium from the interconnect alloy, preventing subsequent chromium migration into the cathode and contact materials. PNNL is currently engaged in studies intended to optimize the composition, microstructure, and fabrication procedure for the spinel protection layers.

  9. Studies on A-TIG welding of Low Activation Ferritic/Martensitic (LAFM) steel

    NASA Astrophysics Data System (ADS)

    Vasantharaja, P.; Vasudevan, M.

    2012-02-01

    Low Activation Ferritic-Martensitic steels (LAFM) are chosen as the candidate material for structural components in fusion reactors. The structural components are generally fabricated by welding processes. Activated Tungsten Inert Gas (A-TIG) welding is an emerging process for welding of thicker components. In the present work, attempt was made to develop A-TIG welding technology for LAFM steel plates of 10 mm thick. Activated flux was developed for LAFM steel by carrying out various bead-on-plate TIG welds without flux and with flux. The optimum flux was identified as one which gave maximum depth of penetration at minimum heat input values. With the optimized flux composition, LAFM steel plate of 10 mm thickness was welded in square butt weld joint configuration using double side welding technique. Optical and Scanning Electron Microscopy was used for characterizing the microstructures. Microhardness measurements were made across the weld cross section for as welded and post weld heat treated samples. Tensile and impact toughness properties were determined. The mechanical properties values obtained in A-TIG weld joint were comparable to that obtained in weld joints of LAFM steel made by Electron beam welding process.

  10. Kinetic Model of Decarburization and Denitrogenation in Vacuum Oxygen Decarburization Process for Ferritic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Xu, Yingtie; Chen, Zhaoping; Zhang, Ge

    2009-06-01

    The characteristics and classification of decarburization and denitrogenation in the vacuum vessel for stainless steel production are analyzed. Based on the analysis of movements of the liquid steel and bubbles, the kinetics of decarburization and denitrogenation in the vacuum oxygen decarburization (VOD) process has been studied. A kinetic model of decarburization and denitrogenation has been developed to simulate the VOD process, considering each reaction zone as oxygen blowing crater, bottom blowing plume, steel/slag interface, and plume eye. As a result, it is possible to quantify the contribution of each reaction zone in decarburization and denitrogenation rate at a different stage in the VOD process. Specific trials at a vacuum induction furnace were performed to refine stainless steel in vacuum carbon deoxidation (VCD) and VOD style, respectively. The trial results are in good agreement with the model calculation. Combining the trials and the model calculation and the influence of temperature control, critical carbon content selection on the terminal total [C] + [N] content can be discussed further to provide a reasonable proposal for high-quality ferritic stainless steel production.

  11. Irradiation behavior of Ti-stabilized 316L type steel

    NASA Astrophysics Data System (ADS)

    Rodchenkov, B. S.; Kalinin, G. M.; Strebkov, Yu. S.; Shamardin, V. K.; Prokhorov, V. I.; Bulanova, T. M.

    2009-04-01

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose ˜4 and 10 dpa at 265 ± 15 °C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  12. Improvement of High Temperature Mechanical Property by Precipitation Hardening of Reduced Activation Ferritic/Martensitic Steels

    SciTech Connect

    Sakasegawa, H.; Kohyama, A.; Katoh, Y.; Tamura, M.; Khono, Y.; Kimura, A.

    2003-07-15

    Reduced Activation Ferritic/Martensitic steels (RAFs) are leading candidates for blanket and first wall structures of the D-T fusion reactors. Recently, in order to achieve better efficiency of energy conversion by using RAFs in advanced blanket systems, improvement of high temperature mechanical property of RAFs is desired. In this work, experimental alloys, FETA-series (Fe-Ta-C or N) steels, were prepared to observe precipitation hardening mechanism by MX-type particles at elevated temperatures in detail. According to the results, innovative improvement of creep property can be achieved by applying of precipitation hardening by very fine TaX (X=C, N) particles. With increasing tantalum content, finer dispersion of MX-type particles, dislocation structures and sub-grain structures were observed by TEM (Transmission Electron Microscopy). These fine structures contributed to the improvement of creep property.

  13. Variation of carbon concentration in proeutectoid ferrite during austenitization in hypoeutectoid steel

    SciTech Connect

    Jung, Minsu; Cho, Wontae; Park, Jihye; Jung, Jae-Gil; Lee, Young-Kook

    2014-08-15

    The variation of the C concentration in proeutectoid ferrite (α{sub PF}) during austenitization in hypoeutectoid steels was quantitatively investigated using the massive transformation start temperature (T{sub m}) of α{sub PF} to austenite (γ) measured by high-temperature confocal laser scanning microscopy and hardness of α{sub PF}. The C concentration in α{sub PF} at T{sub m} in hypoeutectoid steels increased with increasing total C concentration up to approximately 0.2 wt.% during heating. The hardness of α{sub PF} with isothermal holding time at 775 °C in S20C steel revealed C enrichment in α{sub PF} at the early stage of isothermal holding and its reduction with further holding. These results explain the redistribution of the C in α{sub PF} during austenitization as follows: free C atoms released from cementite during pearlite decomposition diffuse excessively into neighboring α{sub PF} as well as pearlitic ferrite. The supersaturated C concentration in α{sub PF} is reduced during the long-range diffusive transformation of α{sub PF} to γ. However, some of the excess C atoms still remain in α{sub PF} until α{sub PF} starts to massively transform to γ. - Highlights: • Massive transformation of αPF to γ in hypoeutectoid steels was observed using CLSM. • C content in αPF during austenitization was analyzed by measured Tm and hardness. • Tm decreases and C content in αPF at Tm increases with increasing total C. • C atoms released from θ during formation of P to γ diffuse excessively into αPF. • Supersaturated C content in αPF is reduced during transformation of αPF to γ.

  14. Diffusive transport parameters of deuterium through China reduced activation ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Wang, Bo; Liu, Lingbo; Xiang, Xin; Rao, Yongchu; Ye, Xiaoqiu; Chen, Chang An

    2016-03-01

    Reduced Activation Ferritic/Martensitic (RAFM) steels have been considered as the most promising candidate structure materials for a fusion reactor. In the recent decades, two new types of RAFM steels, called China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel, have been developed. The gas evolution permeation technique has been used to investigate diffusive transport parameters of deuterium through CLAM and CLF-1 over the temperature range 623 ∼ 873 K at deuterium pressure of 105 Pa. The resultant transport parameters are: Φ (mol. m-1 s-1 Pa-1/2) = 5.40 × 10-8 exp (-46.8 (kJ. mol-1)/RT), D(m2 s-1) = 3.81 × 10-7 exp(-24.0(kJ. mol-1)/RT) and S (mol. m-3 Pa-1/2) = 1.42 × 10-1 exp(-22.8(kJ. mol-1)/RT) for CLAM; while Φ(mol m-1 s-1 Pa-1/2) = 1.76 × 10-8 exp(-43.9(kJ. mol-1)/RT), D(m2. s-1) = 1.02 × 10-7 exp(-16.9(kJ. mol-1)/RT) and S(mol. m-1 Pa-1/2) = 1.73 × 10-1 exp(-27.0(kJ. mol-1) /RT) for CLF-1. The results show that CLAM is more permeable than CLF-1, thus it is easier for hydrogen isotopes to transport and be removed.

  15. Deposition and Evaluation of Protective PVD Coatings on Ferritic Stainless Steel SOFC Interconnects

    SciTech Connect

    Gorokhovsky, Vladimir I.; Gannon, Paul; Deibert, Max; Smith, Richard J.; Kayani, Asghar N.; Kopczyk, M.; Van Vorous, D.; Yang, Z Gary; Stevenson, Jeffry W.; Visco, s.; jacobson, c.; Kurokawa, H.; Sofie, Stephen W.

    2006-09-21

    Reduced operating temperatures (600-800°C) of Solid Oxide Fuel Cells (SOFCs) may enable the use of inexpensive ferritic steels as interconnects. Due to the demanding SOFC interconnect operating environment, protective coatings are gaining attention to increase longterm stability. In this study, large area filtered arc deposition (LAFAD) and hybrid filtered arc assisted electron beam physical vapor deposition (FA-EBPVD) technologies were used for deposition of two-segment coatings with Cr-Co-Al-O-N based sublayer and Mn-Co-O top layer. Coatings were deposited on ferritic steel and subsequently annealed in air for various time intervals. Surface oxidation was investigated using RBS, SEM and EDS analyses. Cr volatilization was evaluated using a transpiration apparatus and ICP-MS analysis of the resultant condensate. Electrical conductivity (Area Specific Resistance) was studied as a function of time using the four-point technique with Ag electrodes. The oxidation behavior, Cr volatilization rate, and electrical conductivity of the coated and uncoated samples are reported. Transport mechanisms for various oxidizing species and coating diffusion barrier properties are discussed.

  16. Assessment of Tungsten Content on Tertiary Creep Deformation Behavior of Reduced Activation Ferritic-Martensitic Steel

    NASA Astrophysics Data System (ADS)

    Vanaja, J.; Laha, Kinkar

    2015-10-01

    Tertiary creep deformation behavior of reduced activation ferritic-martensitic (RAFM) steels having different tungsten contents has been assessed. Creep tests were carried out at 823 K (550 °C) over a stress range of 180 to 260 MPa on three heats of the RAFM steel (9Cr-W-0.06Ta-0.22V) with tungsten content of 1, 1.4, and 2.0 wt pct. With creep exposure, the steels exhibited minimum in creep rate followed by progressive increase in creep rate until fracture. The minimum creep rate decreased, rupture life increased, and the onset of tertiary stage of creep deformation delayed with the increase in tungsten content. The tertiary creep behavior has been assessed based on the relationship, , considering minimum creep rate () instead of steady-state creep rate. The increase in tungsten content was found to decrease the rate of acceleration of tertiary parameter ` p.' The relationships between (1) tertiary parameter `p' with minimum creep rate and time spent in tertiary creep deformation and (2) the final creep rate with minimum creep rate revealed that the same first-order reaction rate theory prevailed in the minimum creep rate as well as throughout the tertiary creep deformation behavior of the steel. A master tertiary creep curve of the steels has been developed. Scanning electron microscopic investigation revealed enhanced coarsening resistance of carbides in the steel on creep exposure with increase in tungsten content. The decrease in tertiary parameter ` p' with tungsten content with the consequent decrease in minimum creep rate and increase in rupture life has been attributed to the enhanced microstructural stability of the steel.

  17. Characterization of Ferrite in Tempered Martensite of Modified 9Cr-1Mo Steel Using the Electron Backscattered Diffraction Technique

    NASA Astrophysics Data System (ADS)

    Das, C. R.; Albert, S. K.; Bhaduri, A. K.; Murty, B. S.

    2011-12-01

    Ferrite was identified and characterized in tempered martensitic modified 9Cr-1Mo steel using the electron backscattered diffraction (EBSD) technique. Microstructural examination of the as-received modified 9Cr-1Mo steel revealed the presence of polycrystalline grains without lath morphology having low hardness within a predominantly tempered lath martensitic matrix. These grains were identified as the ferrite phase, and subsequent EBSD data analysis confirmed that the image quality (IQ) index of these grains is higher and boundary line length per unit area is lower than those of martensitic matrix. Therefore, it is proposed that characterization of ferrite phase in martensitic matrix can be carried out using microstructural parameters such as IQ index and boundary line length per unit area obtained from EBSD data analysis.

  18. TEM microscopical examination of the magnetic domain boundaries in a super duplex austenitic-ferritic stainless steel

    SciTech Connect

    Fourlaris, G.; Gladman, T.; Maylin, M.

    1996-12-31

    It has been demonstrated in an earlier publication that significant improvements in the coercivity, maximum induction and remanence values can be achieved, by using a 2205 type Duplex austenitic-ferritic stainless steel (DSS) instead of the low alloy medium carbon steels currently being used. These improvements are achieved in the as received 2205 material, and after small amounts of cold rolling have been applied, to increase the strength. In addition, the modification of the duplex austenitic-ferritic microstructure, via a heat treatment route, results in a finer austenite `island` dispersion in a ferritic matrix and provides an attractive option for further modification of the magnetic characteristics of the material. However, the 2205 type DSS exhibits {open_quotes}marginal{close_quotes} corrosion protection in a marine environment, so that a study has been undertaken to examine whether the beneficial effects exhibited by the 2205 DSS, are also present in a 2507 type super-DSS.

  19. Mechanical properties and characteristics of nanometer-sized precipitates in hot-rolled low-carbon ferritic steel

    NASA Astrophysics Data System (ADS)

    Wang, Xiao-pei; Zhao, Ai-min; Zhao, Zheng-zhi; Huang, Yao; Li, Liang; He, Qing

    2014-03-01

    The microstructures and properties of hot-rolled low-carbon ferritic steel have been investigated by optical microscopy, field-emission scanning electron microscopy, transmission electron microscopy, and tensile tests after isothermal transformation from 600°C to 700°C for 60 min. It is found that the strength of the steel decreases with the increment of isothermal temperature, whereas the hole expansion ratio and the fraction of high-angle grain boundaries increase. A large amount of nanometer-sized carbides were homogeneously distributed throughout the material, and fine (Ti, Mo)C precipitates have a significant precipitation strengthening effect on the ferrite phase because of their high density. The nanometer-sized carbides have a lattice parameter of 0.411-0.431 nm. After isothermal transformation at 650°C for 60 min, the ferrite phase can be strengthened above 300 MPa by precipitation strengthening according to the Ashby-Orowan mechanism.

  20. Experimental and Numerical Study on the Effect of ZDDP Films on Sticking During Hot Rolling of Ferritic Stainless Steel Strip

    NASA Astrophysics Data System (ADS)

    Hao, Liang; Jiang, Zhengyi; Wei, Dongbin; Gong, Dianyao; Cheng, Xiawei; Zhao, Jingwei; Luo, Suzhen; Jiang, Laizhu

    2016-10-01

    The aim of this study is to investigate the effect of zinc dialkyl dithio phosphate (ZDDP) films on sticking during hot rolling of a ferritic stainless steel strip. The surface characterization and crack propagation of the oxide scale are very important for understanding the mechanism of the sticking. The high-temperature oxidation of one typical ferritic stainless was conducted at 1373 K (1100 °C) for understanding its microstructure and surface morphology. Hot-rolling tests of a ferritic stainless steel strip show that no obvious cracks among the oxide scale were observed with the application of ZDDP. A finite element method model was constructed with taking into consideration different crack size ratios among the oxide scale, surface profile, and ZDDP films. The simulation results show that the width of the crack tends to be reduced with the introduction of ZDDP films, which is beneficial for improving sticking.

  1. Effect of gamma irradiation on the structural and magnetic properties of Co–Zn spinel ferrite nanoparticles

    SciTech Connect

    Raut, Anil V.; Kurmude, D.V.; Shengule, D.R.; Jadhav, K.M.

    2015-03-15

    Highlights: • Co–Zn ferrite nanoparticles were examined before and after γ-irradiation. • Single phase cubic spinel structure of Co–Zn was confirmed by XRD data. • The grain size was reported in the range of 52–62 nm after γ-irradiation. • Ms, Hc, n{sub B} were reported to be increased after gamma irradiation. - Abstract: In this work, the structural and magnetic properties of Co{sub 1−x}Zn{sub x}Fe{sub 2}O{sub 4} (0.0 ≤ x ≤ 1.0) ferrite nanoparticles were studied before and after gamma irradiation. The as-synthesized samples of Co–Zn ferrite nanoparticles prepared by sol–gel auto-combustion technique were analysed by XRD which suggested the single phase; cubic spinel structure of the material. Crystal defects produced in the spinel lattice were studied before and after Co{sup 60} γ-irradiation in a gamma cell with a dose rate of 0.1 Mrad/h in order to report the changes in structural and magnetic properties of the Co–Zn ferrite nanoparticles. The average crystallite size (t), lattice parameter (α) and other structural parameters of gamma-irradiated and un-irradiated Co{sub 1−x}Zn{sub x}Fe{sub 2}O{sub 4} spinel ferrite system was calculated from XRD data. The morphological characterizations were performed using scanning electron microscopy (SEM). The magnetic properties were measured using pulse field hysteresis loop tracer by applying magnetic field of 1000 Oe, and the analysis of data obtained revealed that the magnetic property such as saturation magnetization (Ms), coecivity (Hc), magneton number (n{sub B}) etc. magnetic parameters were increased after irradiation.

  2. Mechanisms of Sticking Phenomenon Occurring during Hot Rolling of Two Ferritic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Son, Chang-Young; Kim, Chang Kyu; Ha, Dae Jin; Lee, Sunghak; Lee, Jong Seog; Kim, Kwang Tae; Lee, Yong Deuk

    2007-11-01

    Mechanisms of sticking phenomenon occurring during hot rolling of two ferritic stainless steels, STS 430J1L and STS 436L, were investigated in the present study. A hot-rolling simulation test was carried out using a high-temperature wear tester capable of controlling rolling speed, load, and temperature. The test results at 900 °C and 1000 °C revealed that the sticking process proceeded with three stages, i.e., nucleation, growth, and saturation, for the both stainless steels, and that STS 430J1L had a smaller number of sticking nucleation sites and slower growth rate than the STS 436L because of higher high-temperature hardness, thereby leading to less serious sticking. When the test was conducted at 1070 °C, the sticking hardly occurred in both stainless steels as Fe-Cr oxide layers were formed on the surface of the rolled materials. Thus, in order to prevent or minimize the sticking, it was suggested to improve high-temperature properties of stainless steels in the case of hot rolling at 900 °C to 1000 °C, and to establish appropriate rolling conditions and alloy compositions for ready formation of oxide layers in the case of hot rolling at higher temperatures than 1000 °C.

  3. Microstructural characterizations of 14Cr ODS ferritic steels subjected to hot torsion

    NASA Astrophysics Data System (ADS)

    Karch, A.; Sornin, D.; Barcelo, F.; Bosonnet, S.; de Carlan, Y.; Logé, R.

    2015-04-01

    Oxide dispersion strengthened (ODS) steels are very promising materials for nuclear applications. In this paper, the hot working behavior of ODS ferritic steels, consolidated by hot extrusion, is studied through torsion tests. Three ODS steels are produced acting on both the quantity of Ti and Y2O3 added to the matrix (wt% Fe-14Cr-1W), and the density and size of the nanoparticles. A temperature range of 1000-1200 °C and strain rates from 5 ṡ 10-2 to 5 s-1 are considered. The microstructures of deformed samples are examined by Electron Back-Scatter Diffraction and X-ray diffraction techniques. It is observed that hot plastic strain leads to an early damage with nucleation and growth of cavities along grain boundaries. Except for the damage, very few microstructural and textural evolutions are noticed. The three tested ODS steels exhibit almost the same behavior under hot torsion straining, regardless of the precipitation state. Overall, the experimental results are interpreted through a mechanism of strain accommodation at grain boundaries, with low dislocation activity in the bulk of the grains.

  4. Precipitate phases in normalized and tempered ferritic/martensitic steel P92

    NASA Astrophysics Data System (ADS)

    Shen, Yinzhong; Liu, Huan; Shang, Zhongxia; Xu, Zhiqiang

    2015-10-01

    Ferritic/martensitic steel P92 is a promising candidate for cladding and duct applications in Sodium-Cooled Fast Reactor. The precipitate phases of the P92 steel normalized at 1323 K (1050 °C) for 30 min and tempered at 1038 K (765 °C) for 1 h have been investigated using transmission electron microscopes. Four types of phases consisting of M23C6, MX, M2X and sigma-FeCr were identified in the steel. MX phases consist of Nb-rich M(C,N) carbonitride, Nb-rich MC carbide, V-rich M(C,N) carbonitride, V-rich MC carbide, V-rich MN nitride, and complex MC carbides with Nb-rich MC core and V-rich MC wings. M2X phases consist of Cr-rich M2(C,N) carbonitride, Cr-rich M2C carbide and M2N nitride. Sigma-FeCr has a simple tetragonal lattice and a typical chemical formula of Fe0.45Cr0.45W0.1. M23C6 and MX are the dominant phases, while the sigma-FeCr has the lowest content. The formation of sigma-FeCr and M2X phases in the steel is also discussed.

  5. Effect of alloy composition on high-temperature bending fatigue strength of ferritic stainless steels

    NASA Astrophysics Data System (ADS)

    Ahn, Yong-Sik; Song, Jeon-Young

    2011-12-01

    Exhaust manifolds are subjected to an environment in which heating and cooling cycles occur due to the running pattern of automotive engines. This temperature profile results in the repeated bending stress of exhaust pipes. Therefore, among high-temperature characteristics, the bending fatigue strength is an important factor that affects the lifespan of exhaust manifolds. Here, we report on the effect of the alloy composition, namely the weight fraction of the elements Cr, Mo, Nb, and Ti, on the high-temperature bending fatigue strength of the ferritic stainless steel used in exhaust manifolds. Little difference in the tensile strength and bending fatigue strength of the different composition steels was observed below 600 °C, with the exception of the low-Cr steel. However, steels with high Cr, Mo, or Nb fractions showed considerably larger bending fatigue strength at temperatures of 800 °C. After heating, the precipitates from the specimens were extracted electrolytically and analyzed using scanning electron microscopy energy dispersive spectrometry and transmission electron microscopy. Alloying with Cr and Mo was found to increase the bending fatigue strength due to the substitutional solid solution effect, while alloying with Nb enhanced the strength by forming fine intermetallic compounds, including NbC and Fe2Nb.

  6. Effect of ultrasonic impact peening on the corrosion of ferritic-martensitic steels in supercritical water

    NASA Astrophysics Data System (ADS)

    Dong, Ziqiang; Liu, Zhe; Li, Ming; Luo, Jing-Li; Chen, Weixing; Zheng, Wenyue; Guzonas, Dave

    2015-02-01

    Ferritic-Martensitic (F/M) steels are important candidate alloys to be used in the next generation (Generation-IV) SCWRs. In this work, two F/M steels with the same Cr content of around 12 wt.% and varied Si content from 0.6 wt.% to 2.2 wt.% were evaluated in supercritical water (SCW) at 500 °C and 25 MPa for up to 1000 h. The effect of ultrasonic shot peening on the oxidation behavior of these F/M steels have been investigated. The results showed that the oxidation was affected by the Si content as well as the surface modification. The F/M steel with low Si concentration exhibited higher corrosion resistance than that of the alloy with high Si content. Shot peening, which could modify the microstructure at the surface, showed significantly beneficial effect to improving the oxidation resistance. A thin, uniform oxide layer formed on the peened sample could be attributed to the enhanced diffusion of Cr induced by the surface modification.

  7. Diffusion Bonding Beryllium to Reduced Activation Ferritic Martensitic Steel: Development of Processes and Techniques

    NASA Astrophysics Data System (ADS)

    Hunt, Ryan Matthew

    Only a few materials are suitable to act as armor layers against the thermal and particle loads produced by magnetically confined fusion. These candidates include beryllium, tungsten, and carbon fiber composites. The armor layers must be joined to the plasma facing components with high strength bonds that can withstand the thermal stresses resulting from differential thermal expansion. While specific joints have been developed for use in ITER (an experimental reactor in France), including beryllium to CuCrZr as well as tungsten to stainless steel interfaces, joints specific to commercially relevant fusion reactors are not as well established. Commercial first wall components will likely be constructed front Reduced Activation Ferritic Martensitic (RAFM) steel, which will need to be coating with one of the three candidate materials. Of the candidates, beryllium is particularly difficult to bond, because it reacts during bonding with most elements to form brittle intermetallic compounds. This brittleness is unacceptable, as it can lead to interface crack propagation and delamination of the armor layer. I have attempted to overcome the brittle behavior of beryllium bonds by developing a diffusion bonding process of beryllium to RAFM steel that achieves a higher degree of ductility. This process utilized two bonding aids to achieve a robust bond: a. copper interlayer to add ductility to the joint, and a titanium interlayer to prevent beryllium from forming unwanted Be-Cu intermetallics. In addition, I conducted a series of numerical simulations to predict the effect of these bonding aids on the residual stress in the interface. Lastly, I fabricated and characterized beryllium to ferritic steel diffusion bonds using various bonding parameters and bonding aids. Through the above research, I developed a process to diffusion bond beryllium to ferritic steel with a 150 M Pa tensile strength and 168 M Pa shear strength. This strength was achieved using a Hot Isostatic

  8. Electron work functions of ferrite and austenite phases in a duplex stainless steel and their adhesive forces with AFM silicon probe.

    PubMed

    Guo, Liqiu; Hua, Guomin; Yang, Binjie; Lu, Hao; Qiao, Lijie; Yan, Xianguo; Li, Dongyang

    2016-02-12

    Local electron work function, adhesive force, modulus and deformation of ferrite and austenite phases in a duplex stainless steel were analyzed by scanning force microscopy. It is demonstrated that the austenite has a higher electron work function than the ferrite, corresponding to higher modulus, smaller deformation and larger adhesive force. Relevant first-principles calculations were conducted to elucidate the mechanism behind. It is demonstrated that the difference in the properties between austenite and ferrite is intrinsically related to their electron work functions.

  9. Proceedings of the IEA working group meeting on ferritic/martensitic steels

    SciTech Connect

    Klueh, R.L.

    1995-02-01

    An International Energy Agency (IEA) working group consist- ng of researchers from Japan, the European Union (EU), and the United States, met at the Oak Ridge National Laboratory (ORNL) 16 February 1995 to continue planning a collaborative test program on reduced-activation ferritic/martensitic steels for fusion applications. Plates from a 5-ton, a 1-ton, and three 150 kg heats of reduced-activation martensitic steels have been melted and processed to 7.5- and 15-mm plates in Japan. Plates were delivered in 1994 to the three parties that will participate in the test program. A second 5-ton IEA heat of modified F82H steel was produced in Japan in late 1994, and it was processed into 15- and 25-mm plates, which are to be shipped in February, 1995. Weldments will be produced on plates from this heat, and they will be shipped in April, 1995. At the ORNL meeting, a detailed test program and schedule was presented by the EU representatives, and less detailed programs were presented by the Japanese and US representatives. Detailed program schedules are required from the latter two programs to complete the program planning stage. A meeting is planned for 19--20 September 1995 in Switzerland to continue the planning and coordination of the test program and to present the preliminary results obtained in the collaboration.

  10. Test research on sticking mechanism during hot rolling of SUS 430 ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Liu, Jun-Xian; Zhang, Yong-Jun; Han, Jing-Tao

    2010-10-01

    The sticking phenomenon during hot rolling of SUS 430 ferritic stainless steel was investigated by means of a two-disc type high-temperature wear tester. The test results indicate that sticking particles on the surfaces of high chromium steel (HiCr) and high-speed steel (HSS) rolls undergo nucleation, growth, and saturation stages. Grooves on the roll surface generated by grinding provide nucleation sites for sticking particles. The number of sticking particles on the HiCr roll surface is greater than that on the HSS roll surface. The average surface roughnesses ( R a) of HiCr and HSS rolls change from 0.502 and 0.493 μm at the initial stage to 0.837 and 0.530 μm at the saturation stage, respectively. The test further proves that the sticking behavior is strongly dependent on roll materials, and the HSS roll is more beneficial to prevent particles sticking compared with the HiCr roll under the same hot-rolling conditions.

  11. Insight into the microstructural characterization of ferritic steels using micromagnetic parameters

    SciTech Connect

    Moorthy, V.; Vaidyanathan, S.; Raj, B.; Jayakumar, T.; Kashyap, B.P.

    2000-04-01

    The influence of tempering-induced microstructural changes on the micromagnetic parameters such as magnetic Barkhausen emission (MBE), coercive force (H{sub c}), residual induction (B{sub r}), and maximum induction (B{sub max}) has been studied in 0.2 pct carbon steel, 2.25Cr-1Mo steel, and 9Cr-1Mo steel. It is observed that, after short tempering, the micromagnetic parameters show more or less linear correlation with hardness, which is attributed to the reduction in dislocation density, but long-term tempering produces nonlinear behavior. The variation in each of these parameters with tempering time has been explained based on the changes in the size and distribution of ferrite laths/grains and precipitates. It has been shown that the individual variation in the microstructural features such as size and distribution of laths/grains and precipitates during tempering can be clearly identified by the MBE parameters, which is not possible from the hysteresis loop parameters (H{sub c} and B{sub r}). It is also shown that the MBE parameters cannot only be used to identify different stages of tempering but also to quantify the average size of laths/grains and second-phase precipitates.

  12. Effect of silicon on the microstructure and mechanical properties of reduced activation ferritic/martensitic steel

    NASA Astrophysics Data System (ADS)

    Chen, Shenghu; Rong, Lijian

    2015-04-01

    The effect of Si in the range of 0.05-0.77 wt.% on the microstructure, tensile properties and impact toughness of reduced activation ferritic/martensitic (RAFM) steels has been investigated. An increase in Si content affected the prior austenite grain size resulting in an increase in the tensile strength at room temperature. The tensile strength of steels tested above 773 K did not change significantly with the addition of Si, which was due to the diminished carbide hardening effect and boundary strengthening effect. Detailed fractographic analysis revealed that tear fractures occurred in the samples tensile tested at room temperature, while cup and cone fractures were found in samples tensile tested at temperatures above 773 K, which were induced by the easing of dislocation pile-ups. The ductile-to-brittle transition temperature (DBTT) decreased when the Si content increased to 0.22 wt.%. However, the DBTT increased when the Si content reached 0.77 wt.% and this was due to the precipitation of Laves phase. The RAFM steel with approximately 0.22 wt.% Si content was found to possess an optimized combination of microstructure, tensile properties and impact toughness.

  13. Microstructure control for high strength 9Cr ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Tan, L.; Hoelzer, D. T.; Busby, J. T.; Sokolov, M. A.; Klueh, R. L.

    2012-03-01

    Ferritic-martensitic (F-M) steels with 9 wt.%Cr are important structural materials for use in advanced nuclear reactors. Alloying composition adjustment, guided by computational thermodynamics, and thermomechanical treatment (TMT) were employed to develop high strength 9Cr F-M steels. Samples of four heats with controlled compositions were subjected to normalization and tempering (N&T) and TMT, respectively. Their mechanical properties were assessed by Vickers hardness and tensile testing. Ta-alloying showed significant strengthening effect. The TMT samples showed strength superior to the N&T samples with similar ductility. All the samples showed greater strength than NF616, which was either comparable to or greater than the literature data of the PM2000 oxide-dispersion-strengthened (ODS) steel at temperatures up to 650 °C without noticeable reduction in ductility. A variety of microstructural analyses together with computational thermodynamics provided rational interpretations on the strength enhancement. Creep tests are being initiated because the increased yield strength of the TMT samples is not able to deduce their long-term creep behavior.

  14. Modification in the Microstructure of Mod. 9Cr-1Mo Ferritic Martensitic Steel Exposed to Sodium

    NASA Astrophysics Data System (ADS)

    Prasanthi, T. N.; Sudha, Cheruvathur; Paul, V. Thomas; Bharasi, N. Sivai; Saroja, S.; Vijayalakshmi, M.

    2014-09-01

    Mod. 9Cr-1Mo is used as the structural material in the steam generator circuit of liquid metal-cooled fast breeder reactors. Microstructural modifications on the surface of this steel are investigated after exposing to flowing sodium at a temperature of 798 K (525 °C) for 16000 hours. Sodium exposure results in the carburization of the ferritic steel up to a depth of ~218 µm from the surface. Electron microprobe analysis revealed the existence of two separate zones with appreciable difference in microchemistry within the carburized layer. Differences in the type, morphology, volume fraction, and microchemistry of the carbides present in the two zones are investigated using analytical transmission electron microscopy. Formation of separate zones within the carburized layer is understood as a combined effect of leaching, diffusion of the alloying elements, and thermal aging. Chromium concentration on the surface in the α-phase suggested possible degradation in the corrosion resistance of the steel. Further, concentration-dependent diffusivities for carbon are determined in the base material and carburized zones using Hall's and den Broeder's methods, respectively. These are given as inputs for simulating the concentration profiles for carbon using numerical computation technique based on finite difference method. Predicted thickness of the carburized zone agrees reasonably well with that of experiment.

  15. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    SciTech Connect

    Wiffen, F.W.; Santoro, R. T.

    1983-01-01

    Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m/sup 2/ service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V.

  16. Stability of nanoclusters in 14YWT oxide dispersion strengthened steel under heavy ion-irradiation by atom probe tomography

    SciTech Connect

    Jianchao He; Farong Wan; Kumar Sridharan; Todd R. Allen; A. Certain; V. Shutthanandan; Y.Q. Wu

    2014-12-01

    14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 C, 450 C, and 600 C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 1023 to 3.6 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.

  17. Characterization of twin-like structure in a ferrite-based lightweight steel

    NASA Astrophysics Data System (ADS)

    Nezhadfar, Pooriya Dastranjy; Zarei-Hanzaki, Abbas; Sohn, Seok Su; Abedi, Hamid Reza

    2016-09-01

    The present study examined cold to warm compressive deformation behavior of a ferrite- based lightweight steel through characterization of the banded structures. Compression tests were carried out at 25 to 500 °C at a strain rate of 0.01 s-1 up to true strain of 0.6. Analysis of the microstructural evolution using electron back scatter diffraction indicated that the twin-like bands in the large ferrite grains occurred with the {112}[111] system at a 60° misorientation. Density of the twin-like bands is increased by raising the deformation temperature. EBSD results showed that the primary and secondary twins occurred in the [-11-1] and [1-1-1] directions. In addition, the strain at 500 °C distorted the twin-like bands and resulted in wavy boundaries. The strain hardening behavior was also examined using the Crussard-Jaoul (C-J) model and the n-values were calculated for each stage of imposing strain. The results showed high dislocation density in the adjacent of twin-like boundaries intersections which resulted in the n-value increment.

  18. Microstructural origin of the skeletal ferrite morphology of austenitic stainless steel welds

    SciTech Connect

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

    1982-04-01

    Scanning transmission electron microscopy was conducted on welds exhibiting a variety of skeletal, or vermicular ferrite morphologies in addition to one lathy ferrite morphology. These ferrite morphologies result from primary ferrite solidification followed by a solid state transformation upon cooling. During cooling, a large fraction of the ferrite transforms to austenite leaving a variety of ferrite morphologies. Comparison of composition profiles and alloy partitioning showed both the skeletal and lathy ferrite structures result from a diffusion controlled solid state transformation. However, the overall measured composition profiles of the weld structure are a result of partitioning during both solidification and the subsequent solid state transformation.

  19. Low-Cycle Fatigue Properties of P92 Ferritic-Martensitic Steel at Elevated Temperature

    NASA Astrophysics Data System (ADS)

    Zhang, Zhen; Hu, ZhengFei; Schmauder, Siegfried; Mlikota, Marijo; Fan, KangLe

    2016-04-01

    The low-cycle fatigue behavior of P92 ferritic-martensitic steel and the corresponding microstructure evolution at 873 K has been extensively studied. The test results of fatigue lifetime are consistent with the Coffin-Manson relationship over a range of controlled total strain amplitudes from 0.15 to 0.6%. The influence of strain amplitude on the fatigue crack initiation and growth has been observed using optical microscopy and scanning electron microscopy. The formation mechanism of secondary cracks is established according to the observation of fracture after fatigue process and there is an intrinsic relationship between striation spacing, current crack length, and strain amplitude. Transmission electron microscopy has been employed to investigate the microstructure evolution after fatigue process. It indicates the interaction between carbides and dislocations together with the formation of cell structure inhibits the cyclic softening. The low-angle sub-boundary elimination in the martensite is mainly caused by the cyclic stress.

  20. Fracture toughness of low activation ferritic steel (JLF-1) weld joint at room temperature

    NASA Astrophysics Data System (ADS)

    Nishimura, A.; Inoue, N.; Muroga, T.

    1998-10-01

    A low activation ferritic steel has been developed for a candidate of a structural material of nuclear fusion reactors. Since welding must be performed when the support structures are constructed, fracture toughness of the weld joint has to be characterized as well as the base metal in an engineering sense. In this report, 25 mm thick plates of JLF-1, which contains 9% Cr and 2% W, are butt-welded by a tungsten inert gas (TIG) procedure, and the fracture toughness of the base plate and the weld metal is investigated at room temperature using 1T and 0.5T CT specimens. The base metal reveals high fracture toughness of about 430 kJ/m 2. However, the weld metal showed unstable big pop-ins. One sample fractured in a nearly elastic condition and another sample showed a toughness of over 400 kJ/m 2.

  1. Morphology, Structure, and Chemistry of Nanoclusters in a Mechanically-Alloyed Nanostructured Ferritic Steel

    SciTech Connect

    Brandes, Matthew C; Kovarik, L.; Miller, Michael K; Mills, Michael J.

    2012-01-01

    Nanostructured ferritic steels have excellent elevated temperature strengths, creep resistances, and radiation tolerances due to the presence of a high density of Ti-Y-O-enriched nanoclusters. The compositions, morphologies, and structures of the smallest of these nanoclusters with maximum dimensions of {approx}2-4 nm were investigated in alloy 14YWT by high-resolution scanning transmission electron microscopy and atom probe tomography. Nanoclusters are found to be coherent with truncated rhombic dodecahedron morphologies defined by the {l_brace}100{r_brace} and {l_brace}110{r_brace} planes in the Fe matrix. Particles have compositions rich in Ti, O, Y, and Cr that are inconsistent with known oxide structures. The smallest nanoclusters appear to lack an identifiable crystal structure. Both nano-diffraction and focal series imaging through the sample thickness suggest that they are amorphous.

  2. From embryos to precipitates: a study of nucleation and growth in a multicomponent ferritic steel

    SciTech Connect

    Zhang, Zhongwu; Liu, C T; Wang, Xun-Li; Littrell, Ken; Miller, Michael K; An, Ke; Chin, Bryan

    2011-01-01

    The nucleation and growth of nanoscale precipitates in a new class of high-strength, multicomponent, ferritic steels has been studied with complementary state-of-the-art microstructural characterization techniques of atom probe tomography for individual embryos and precipitates and small-angle neutron scattering for their statistical averages. Both techniques revealed a bimodal size distribution, with subnanometer embryos, and nanoscale precipitates. The embryos, which have a radius of {approx}0.4 nm, are enriched in Cu and served as preferential sites for nucleation. The critical radius for nucleation was determined to be {approx}0.7 nm. Subsequent growth of the precipitates is dictated by volumetric diffusion, as predicted by the Lifshitz-Slyozov-Wagner theory.

  3. Effect of deformation on ferrite nucleation and growth in a plain carbon and two microalloyed steels

    NASA Astrophysics Data System (ADS)

    Essadiqi, E.; Jonas, J. J.

    1989-06-01

    Isothermal compression tests were carried out on plain C, Mo, and Mo-Nb-V microalloyed steels in order to study the effect of austenite deformation on the ferrite nucleation and growth rates. The nucleation rate increases with deformation and the degree of supersaturation, Ae3- T; it appears to be reduced by the substitutional elements Mo, Nb, and V through reduction of the austenite grain boundary energy. The growth rate increases with the degree of supersaturation and is also reduced by these elements, apparently through the solute drag-like effect. Under static conditions, increasing the prestraining strain rate increases the nucleation rate, but this increase is small compared to the effect of concurrent deformation. The growth rate under static conditions decreases as the deformation or the strain rate is increased.

  4. Innovative Powder Processing of Oxide Dispersion Strengthened ODS Ferritic Stainless Steels

    SciTech Connect

    Rieken, Joel; Anderson, Iver; Kramer, Matthew

    2011-04-01

    An innovative gas atomization reaction synthesis technique was employed as a viable method to dramatically lower the processing cost for precursor oxide dispersion forming ferritic stainless steel powders (i.e., Fe-Cr-(Hf,Ti)-Y). During this rapid solidification process the atomized powders were enveloped by a nano-metric Cr-enriched metastable oxide film. Elevated temperature heat treatment was used to dissociate this metastable oxide phase through oxygen exchange reactions with Y-(Hf,Ti) enriched intermetallic compound precipitates. These solid state reactions resulted in the formation of highly stable nano-metric mixed oxide dispersoids (i.e., Y-Ti-O or Y-Hf-O) throughout the alloy microstructure. Subsequent high temperature (1200 C) heat treatments were used to elucidate the thermal stability of each nano-metric oxide dispersoid phase. Transmission electron microscopy coupled with X-ray diffraction was used to evaluate phase evolution within the alloy microstructure.

  5. Investigation of AISI 441 Ferritic Stainless Steel and Development of Spinel Coatings for SOFC Interconnect Applications

    SciTech Connect

    Yang, Zhenguo; Xia, Guanguang; Wang, Chong M.; Nie, Zimin; Templeton, Joshua D.; Singh, Prabhakar; Stevenson, Jeffry W.

    2008-05-30

    As part of an effort to develop cost-effective ferritic stainless steel-based interconnects for solid oxide fuel cell (SOFC) stacks, both bare and spinel coated AISI 441 were studied in terms of metallurgical characteristics, oxidation behavior, and electrical performance. The conventional melt metallurgy used for the bulk alloy fabrication leads to significant processing cost reduction and the alloy chemistry with the presence of minor alloying additions of Nb and Ti facilitate the strengthening by precipitation and formation of Laves phase both inside grains and along grain boundaries during exposure in the intermediate SOFC operating temperature range. The Laves phase formed along the grain boundaries also ties up Si and prevents the formation of an insulating silica layer at the scale/metal interface during prolonged exposure. The substantial increase in ASR during long term oxidation due to oxide scale growth suggested the need for a conductive protection layer, which could also minimize Cr evaporation. In particular, Mn1.5Co1.5O4 based surface coatings on planar coupons drastically improved the electrical performance of the 441, yielding stable ASR values at 800ºC for over 5,000 hours. Ce-modified spinel coatings retained the advantages of the unmodified spinel coatings, and also appeared to alter the scale growth behavior beneath the coating, leading to a more adherent scale. The spinel protection layers appeared also to improve the surface stability of 441 against the anomalous oxidation that has been observed for ferritic stainless steels exposed to dual atmosphere conditions similar to SOFC interconnect environments. Hence, it is anticipated that, compared to unmodified spinel coatings, the Ce-modified coatings may lead to superior structural stability and electrical performance.

  6. Long term high temperature oxidation characteristics of La and Cu alloyed ferritic stainless steels for solid oxide fuel cell interconnects

    NASA Astrophysics Data System (ADS)

    Swaminathan, Srinivasan; Lee, Young-Su; Kim, Dong-Ik

    2016-09-01

    To ensure the best performance of solid oxide fuel cell metallic interconnects, the Fe-22 wt.% Cr ferritic stainless steels with various La contents (0.006-0.6 wt.%) and Cu addition (1.57 wt.%), are developed. Long-term isothermal oxidation behavior of these steels is investigated in air at 800 °C, for 2700 h. Chemistry, morphology, and microstructure of the thermally grown oxide scale are examined using XPS, SEM-EDX, and XRD techniques. Broadly, all the steels show a double layer consisting of an inner Cr2O3 and outer (Mn, Cr)3O4. Distinctly, in the La-added steels, binary oxides of Cr, Mn and Ti are found at the oxide scale surface together with (Mn, Cr)3O4. Furthermore, all La-varied steels possess the metallic Fe protrusions along with discontinuous (Mn, Cr)3O4 spinel zones at the oxide scale/metal interface and isolated precipitates of Ti-oxides in the underlying matrix. Increase of La content to 0.6 wt.% is detrimental to the oxidation resistance. For the Cu-added steel, Cu is found to segregate strongly at the oxide scale/metal interface which inhibits the ingress of oxygen thereby suppressing the subscale formation of (Mn, Cr)3O4. Thus, Cu addition to the Fe-22Cr ferritic stainless steels benefits the oxidation resistance.

  7. Recent progress in US Japan collaborative research on ferritic steels R&D

    NASA Astrophysics Data System (ADS)

    Kimura, Akihiko; Kasada, Ryuta; Kohyama, Akira; Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki; Jitsukawa, Shiro; Ohtsuka, Satoshi; Ukai, Shigeharu; Sokolov, M. A.; Klueh, R. L.; Yamamoto, Takuya; Odette, G. R.

    2007-08-01

    The mechanisms of irradiation embrittlement of two Japanese RAFSs were different from each other. The larger DBTT shift observed in F82H is interpreted by means of both hardening effects and a reduction of cleavage fracture stress by M 23C 6 carbides precipitation along lath block and packet boundaries, while that of JLF-1 is due to only the hardening effect. Dimensional change measurement during in-pile creep tests revealed the creep strain of F82H was limited at 300 °C. Performance of the weld bond under neutron irradiation will be critical to determine the life time of blanket structural components. Application of the ODS steels, which are resistant to corrosion in supercritical pressurized water, to the water-cooled blanket is essential to increase thermal efficiency of the blanket systems beyond DEMO. The coupling of RAFS and ODS steel could be effective to realize a highly efficient fusion blanket.

  8. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    DOE PAGES

    Parish, Chad M.; Unocic, Kinga A.; Tan, Lizhen; ...

    2016-10-24

    Here we irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~1021 m-3 (CNA), and of ~3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces inmore » all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Finally, among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.« less

  9. Structural and chemical evolution in neutron irradiated and helium-injected ferritic ODS PM2000 alloy

    NASA Astrophysics Data System (ADS)

    Jung, Hee Joon; Edwards, Dan J.; Kurtz, Richard J.; Yamamoto, Takuya; Wu, Yuan; Odette, G. Robert

    2017-02-01

    An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 58Ni(nth,γ) 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

  10. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    NASA Astrophysics Data System (ADS)

    Parish, C. M.; Unocic, K. A.; Tan, L.; Zinkle, S. J.; Kondo, S.; Snead, L. L.; Hoelzer, D. T.; Katoh, Y.

    2017-01-01

    We irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ∼50 dpa, ∼15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ∼8 nm, ∼1021 m-3 (CNA), and of ∼3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ∼50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.

  11. Helium sequestration at nanoparticle-matrix interfaces in helium + heavy ion irradiated nanostructured ferritic alloys

    SciTech Connect

    Parish, Chad M.; Unocic, Kinga A.; Tan, Lizhen; Zinkle, S. J.; Kondo, Sosuke; Snead, Lance Lewis; Hoelzer, David T.; Katoh, Yutai

    2016-10-24

    Here we irradiated four ferritic alloys with energetic Fe and He ions: one castable nanostructured alloy (CNA) containing Ti-W-Ta-carbides, and three nanostructured ferritic alloys (NFAs). The NFAs were: 9Cr containing Y-Ti-O nanoclusters, and two Fe-12Cr-5Al NFAs containing Y-Zr-O or Y-Hf-O clusters. All four were subjected to simultaneous dual-beam Fe + He ion implantation (650 °C, ~50 dpa, ~15 appm He/dpa), simulating fusion-reactor conditions. Examination using scanning/transmission electron microscopy (STEM) revealed high-number-density helium bubbles of ~8 nm, ~1021 m-3 (CNA), and of ~3 nm, 1023 m-3 (NFAs). STEM combined with multivariate statistical analysis data mining suggests that the precipitate-matrix interfaces in all alloys survived ~50 dpa at 650 °C and serve as effective helium trapping sites. All alloys appear viable structural material candidates for fusion or advanced fission energy systems. Finally, among these developmental alloys the NFAs appear to sequester the helium into smaller bubbles and away from the grain boundaries more effectively than the early-generation CNA.

  12. Investigation of Iron-Chromium-Niobium-Titanium Ferritic Stainless Steel for Solid Oxide Fuel Cell Interconnect Applications

    SciTech Connect

    Yang, Zhenguo; Xia, Guanguang; Wang, Chong M.; Nie, Zimin; Templeton, Joshua D.; Stevenson, Jeffry W.; Singh, Prabhakar

    2008-09-01

    As part of an effort to develop cost-effective ferritic stainless steel-based interconnects for solid oxide fuel cell (SOFC) stacks, AL 441 HPTM was studied in terms of its metallurgical characteristics, oxidation behavior, and electrical performance. Minor alloying elements (Nb and Ti) captured interstitials such as C by forming carbides, stabilizing the ferritic structure and mitigating the risks of sensitization and inter-granular corrosion. Laves phases rich in Nb and Si precipitated along grain boundaries during high temperature exposure, improving the steel’s high temperature mechanical strength. The capture of Si in the Laves phase minimized the Si activity in the steel substrate and prevented formation of an insulating silica layer at the scale/metal interface. However, the relatively high oxidation rate, and thus increasing ASR over time, necessitates the application of a conductive protection layer on the steel. In particular, Mn1.5Co1.5O4 spinel protection layers drastically improved the electrical performance of the ferritic stainless steel 441, acting as barriers to chromium outward and oxygen inward diffusion.

  13. The Influence of the Induced Ferrite and Precipitates of Ti-bearing Steel on the Ductility of Continuous Casting Slab

    NASA Astrophysics Data System (ADS)

    Qian, Guoyu; Cheng, Guoguang; Hou, Zibing

    2015-11-01

    In order to investigate the loss of the ductility of Ti-bearing ship plate steel under 1000 °C, where the ductility begins to reduce rapidly, so the hot ductility of Ti-bearing ship plate steel has been obtained using the Gleeble 1500 thermal-mechanical simulator and also the studies about the effect of grain boundary ferrite films and precipitates containing Ti on the ductility has been carried out. The result showed that the TiN particles precipitating at 950 °C with a larger size and smaller volume fraction cannot effectively suppress the occurrence of recrystallization and the ductility still retains at a high level, although R.A. value presents a certain degree of decline compared with 1000 °C. A large number of smaller Ti(C,N) particles precipitate at 900 °C and can induce the formation of a very small amount of fine grain boundary ferrite, which deteriorates the adhesion strength of the grain boundary, so the R.A. value rapidly reduces to less than 50%. When the temperature falls to close Ae3 (827 °C), the amount of the grain boundary ferrite films increase due to the ferrite phase transformation, but the ferrite film thickness becomes more uneven at the same time, which results in the increase of strain concentration and plays a leading role in causing the decrease of ductility, so the R.A. value has been kept less than 40% as the temperature cooling to 800 °C from 850 °C. When the temperature further decreases, the ductility starts to recover due to the increase of average ferrite film thickness to a greater degree which greatly reduces the strain concentration of the grain boundary.

  14. Sensitivity of ultrasonic nonlinearity to irradiated, annealed, and re-irradiated microstructure changes in RPV steels

    SciTech Connect

    Matlack, Katie; Kim, J-Y.; Wall, J.J.; Jacobs, L.J.; Sokolov, Mikhail A

    2014-05-01

    The planned life extension of nuclear reactors throughout the US and abroad will cause reactor vessel and internals materials to be exposed to more neutron irradiation than was originally intended. A nondestructive evaluation (NDE) method to monitor radiation damage would enable safe and cost-effective continued operation of nuclear reactors. Radiation damage in reactor pressure vessel (RPV) steels causes microstructural changes that leave the material in an embrittled state. Nonlinear ultrasound is an NDE technique quantified by the measurable acoustic nonlinearity parameter, which is sensitive to microstructural changes in metallic materials such as dislocations, precipitates and their combinations. Recent research has demonstrated the sensitivity of the acoustic nonlinearity parameter to increasing neutron fluence in representative RPV steels. The current work considers nonlinear ultrasonic experiments conducted on similar RPV steel samples that had a combination of irradiation, annealing, re-irradiation, and/or re-annealing to a total neutron fluence of 0.5 5 1019 n/cm2 (E > 1 MeV) at an irradiation temperature of 290 C. The acoustic nonlinearity parameter generally increased with increasing neutron fluence, and consistently decreased from the irradiated to the annealed state over different levels of neutron fluence. Results of the measured acoustic nonlinearity parameter are compared with those from previous measurements on other RPV steel samples. This comprehensive set of results illustrates the dependence of the measured acoustic nonlinearity parameter on neutron fluence, material composition, irradiation temperature and annealing.

  15. Dependence of Precipitation Behavior and Creep Strength on Cr Content in High Cr Ferritic Heat Resistant Steels

    NASA Astrophysics Data System (ADS)

    Murata, Yoshinori; Yamashita, Koji; Morinaga, Masahiko; Hara, Toru; Miki, Kazuhiro; Azuma, Tsukasa; Ishiguro, Toru; Hashizume, Ryokichi

    It is known that high temperature tensile strength increases with increasing Cr content in Cr containing heat resistant steels. Recently, however, it was found that long-term creep strength decreased with increasing Cr content in the heat resistant steels containing 8.5-12%Cr. In this study, precipitation behavior of M23C6 carbide and the Z phase after creep tests was investigated using two kinds of high Cr ferritic steels (9Cr and 10.5Cr). As a result, 10.5Cr steel exhibited larger average particle size of M23C6 than 9Cr steel irrespective of creep stress levels, but the amount of M23C6 carbide was almost the same in both steels. On the other hand, the amount of the Z phase became large in 10.5Cr steel compared with 9Cr steel. These experimental results indicate that high level of Cr content accelerates precipitation and coalescence rate of both M23C6 carbide and the Z phase, resulting in degradation of long term creep strength in 10.5 Cr steel compared to 9Cr steel.

  16. Development of rapidly quenched brazing foils to join tungsten alloys with ferritic steel

    NASA Astrophysics Data System (ADS)

    Kalin, B. A.; Fedotov, V. T.; Sevrjukov, O. N.; Moeslang, A.; Rohde, M.

    2004-08-01

    Results on rapidly solidified filler metals for tungsten brazing are presented. A rapidly quenched foil-type filler metal based on Ni bal-15Cr-4Mo-4Fe-(0.5-1.0)V-7.5Si-1.5B was developed to braze tungsten to ferritic/martensitic Crl3Mo2NbVB steel (FS) for helium gas cooled divertors and plasma facing components. Polycrystalline W-2CeO 2 and monocrystalline pure tungsten were brazed to the steel under vacuum at 1150 °C, using a 0.5 mm thick foil spacer made of a 50Fe-50Ni alloy. As a result of thermocycling tests (100 cycles between 700 °C/20 min and air-water cooling/3-5 min) on brazed joints, tungsten powder metallurgically processed W-2CeO 2 failed due to residual stresses, whereas the brazed joint with zone-melted monocrystalline tungsten withstood the thermocycling tests.

  17. Microstructural changes induced near crack tip during corrosion fatigue tests in austenitic-ferritic steel.

    PubMed

    Gołebiowski, B; Swiatnicki, W A; Gaspérini, M

    2010-03-01

    Microstructural changes occurring during fatigue tests of austenitic-ferritic duplex stainless steel (DSS) in air and in hydrogen-generating environment have been investigated. Hydrogen charging of steel samples during fatigue crack growth (FCG) tests was performed by cathodic polarization of specimens in 0.1M H(2)SO(4) aqueous solution. Microstructural investigations of specimens after FCG tests were carried out using transmission electron microscopy to reveal the density and arrangement of dislocations formed near crack tip. To determine the way of crack propagation in the microstructure, electron backscatter diffraction investigations were performed on fatigue-tested samples in both kinds of environment. To reveal hydrogen-induced phase transformations the atomic force microscopy was used. The above investigations allowed us to define the character of fatigue crack propagation and microstructural changes near the crack tip. It was found that crack propagation after fatigue tests in air is accompanied with plastic deformation; a high density of dislocations is observed at large distance from the crack. After fatigue tests performed during hydrogen charging the deformed zone containing high density of dislocations is narrow compared to that after fatigue tests in air. It means that hydrogenation leads to brittle character of fatigue crack propagation. In air, fatigue cracks propagate mostly transgranularly, whereas in hydrogen-generating environment the cracks have mixed transgranular/interfacial character.

  18. Fretting corrosion resistance and fretting corrosion product cytocompatibility of ferritic stainless steel.

    PubMed

    Xulin, S; Ito, A; Tateishi, T; Hoshino, A

    1997-01-01

    To avoid nickel ion release from SUS317L as an implant material, a new type of nickel, commercially free, of high purity, and high chromium ferritic stainless steel, was developed. The new stainless steel (FJ) was studied for aspects of fretting corrosion and cytocompatibility compared with SUS317L. A pin-on-plate fretting corrosion test in an artificial physiologic solution, and cell culture in media with the addition of the artificial physiologic solution used for fretting was conducted. Resistance to the fretting induced crevice corrosion of FJ was higher than that of SUS317L because of the favorable electrochemical stability of the FJ alloy. The amount of iron ion or colloidal fine particles released from FJ was about a quarter of that from SUS317L, although the weight loss of a pin of FJ was almost 5/3 that of SUS317L. The artificial physiologic solution used for SUS317L fretting was more harmful to the growth of L929 and MC3T3-E1 cells than that used for FJ fretting. FJ was therefore superior to SUS317L as a biomaterial, judging from the resistance to fretting-induced crevice corrosion, electrochemical stability, and the cytocompatibility of fretting corrosion products.

  19. Effect of Proeutectoid Ferrite Morphology on the Microstructure and Mechanical Properties of Hot Rolled 60Si2MnA Spring Steel

    NASA Astrophysics Data System (ADS)

    Yang, Hu; Wei-qing, Chen; Huai-bin, Han; Rui-juan, Bai

    2017-02-01

    The hot rolled 60Si2MnA spring steel was transformed to obtain different proeutectoid ferrite morphologies by different cooling rates after finish rolling through dynamic thermal simulation test. The coexistence relationship between proeutectoid ferrite and pearlite, and the effect of proeutectoid ferrite morphology on mechanical properties were systematically investigated. Results showed that the reticular proeutectoid ferrite could be formed by the cooling rates of 0.5-2 °C/s; the small, dispersed and blocky proeutectoid ferrite could be formed by the increased cooling rates of 3-5 °C/s; and the bulk content of proeutectoid ferrite decreased. The pearlitic colony and interlamellar spacing also decreased, the reciprocal of them both followed a linear relationship with the reciprocal of proeutectoid ferrite bulk content. Besides, the tensile strength, percentage of area reduction, impact energy and microhardness increased, which all follow a Hall-Petch-type relationship with the inverse of square root of proeutectoid ferrite bulk content. The fracture morphologies of tensile and impact tests transformed from intergranular fracture to cleavage and dimple fracture, and the strength and plasticity of spring steel were both improved. The results have been explained on the basis of proeutectoid ferrite morphologies-microstructures-mechanical properties relationship effectively.

  20. Effect of Austenitic and Austeno-Ferritic Electrodes on 2205 Duplex and 316L Austenitic Stainless Steel Dissimilar Welds

    NASA Astrophysics Data System (ADS)

    Verma, Jagesvar; Taiwade, Ravindra V.

    2016-11-01

    This study addresses the effect of different types of austenitic and austeno-ferritic electrodes (E309L, E309LMo and E2209) on the relationship between weldability, microstructure, mechanical properties and corrosion resistance of shielded metal arc welded duplex/austenitic (2205/316L) stainless steel dissimilar joints using the combined techniques of optical, scanning electron microscope, energy-dispersive spectrometer and electrochemical. The results indicated that the change in electrode composition led to microstructural variations in the welds with the development of different complex phases such as vermicular ferrite, lathy ferrite, widmanstatten and intragranular austenite. Mechanical properties of welded joints were diverged based on compositions and solidification modes; it was observed that ferritic mode solidified weld dominated property wise. However, the pitting corrosion resistance of all welds showed different behavior in chloride solution; moreover, weld with E2209 was superior, whereas E309L exhibited lower resistance. Higher degree of sensitization was observed in E2209 weld, while lesser in E309L weld. Optimum ferrite content was achieved in all welds.

  1. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  2. Comparison between magnetic force microscopy and electron back-scatter diffraction for ferrite quantification in type 321 stainless steel.

    PubMed

    Warren, A D; Harniman, R L; Collins, A M; Davis, S A; Younes, C M; Flewitt, P E J; Scott, T B

    2015-01-01

    Several analytical techniques that are currently available can be used to determine the spatial distribution and amount of austenite, ferrite and precipitate phases in steels. The application of magnetic force microscopy, in particular, to study the local microstructure of stainless steels is beneficial due to the selectivity of this technique for detection of ferromagnetic phases. In the comparison of Magnetic Force Microscopy and Electron Back-Scatter Diffraction for the morphological mapping and quantification of ferrite, the degree of sub-surface measurement has been found to be critical. Through the use of surface shielding, it has been possible to show that Magnetic Force Microscopy has a measurement depth of 105-140 nm. A comparison of the two techniques together with the depth of measurement capabilities are discussed.

  3. Effect of microstructural evolution by isothermal aging on the mechanical properties of 9Cr-1WVTa reduced activation ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Park, Min-Gu; Lee, Chang-Hoon; Moon, Joonoh; Park, Jun Young; Lee, Tae-Ho; Kang, Namhyun; Chan Kim, Hyoung

    2017-03-01

    The influence of microstructural changes caused by aging condition on tensile and Charpy impact properties was investigated for reduced activation ferritic-martensitic (RAFM) 9Cr-1WVTa steels having single martensite and a mixed microstructure of martensite and ferrite. For the mixed microstructure of martensite and ferrite, the Charpy impact properties deteriorated in both as-normalized and tempered conditions due to the ferrite and the accompanying M23C6 carbides at the ferrite grain boundaries which act as path and initiation sites for cleavage cracks, respectively. However, aging at 550 °C for 20-100 h recovered gradually the Charpy impact toughness without any distinct drop in strength, as a result of the spheroidization of the coarse M23C6 carbides at the ferrite grain boundaries, which makes crack initiation more difficult.

  4. Effect of Austenization Temperature on the Microstructure and Strength of 9% and 12% Cr Ferritic-Martensitic Steels

    SciTech Connect

    Terry C. Totemeier

    2004-10-01

    The effect of reduced-temperature austenization on the microstructure and strength of two ferritic-martensitic steels was studied. Prototypic 9% and 12% Cr steels, grade 91 (9Cr-1MoVNb) and type 422 stainless (12Cr-1MoVW), respectively, were austenized at 925°C and 1050°C and tempered at 760°C. The reduced austenization temperature was intended to simulate potential inadequate austenization during field construction of large structures and also the thermal cycle experienced in the Type IV region of weld heat affected zones (HAZ). The microstructure, tensile behavior, and creep strength were characterized for both steels treated at each condition. The reduced austenization temperature resulted in general coarsening of carbides in both steels and polygonization of the tempered martensite structure in type 422. For this steel, a marked reduction in microhardness was observed, while there was little change in microhardness for grade 91. Slight reductions in tensile strength were observed for both steels at room temperature and elevated temperatures of 450 and 550°C. The strength reduction was greater for type 422 than for grade 91. At 650°C the tensile strength reduction was minimal for both steels. Marked reductions in creep rupture lives were observed for both steels at 650°C; the reductions were less at 600°C and minimal at 550°C. Overall, the higher Cr content steel was observed to be more sensitive to variations in heat treatment conditions.

  5. Microstructural Variations Across a Dissimilar 316L Austenitic: 9Cr Reduced Activation Ferritic Martensitic Steel Weld Joint

    NASA Astrophysics Data System (ADS)

    Thomas Paul, V.; Karthikeyan, T.; Dasgupta, Arup; Sudha, C.; Hajra, R. N.; Albert, S. K.; Saroja, S.; Jayakumar, T.

    2016-03-01

    This paper discuss the microstructural variations across a dissimilar weld joint between SS316 and 9Cr-RAFM steel and its modifications on post weld heat treatments (PWHT). Detailed characterization showed a mixed microstructure of austenite and martensite in the weld which is in agreement with the phases predicted using Schaeffler diagram based on composition measurements. The presence of very low volume fraction of δ-ferrite in SS316L has been identified employing state of the art electron back-scattered diffraction technique. PWHT of the ferritic steel did not reduce the hardness in the weld metal. Thermal exposure at 973 K (700 °C) showed a progressive reduction in hardness of weld joint with duration of treatment except in austenitic base metal. However, diffusion annealing at 1073 K (800 °C) for 100 hours resulted in an unexpected increase in hardness of weld metal, which is a manifestation of the dilution effects and enrichment of Ni on the transformation characteristics of the weld zone. Migration of carbon from ferritic steel aided the precipitation of fine carbides in the austenitic base metal on annealing at 973 K (700 °C); but enhanced diffusion at 1073 K (880 °C) resulted in coarsening of carbides and thereby reduction of hardness.

  6. The role of nitrogen in the preferential chromium segregation on the ferritic stainless steel (1 1 1) surface

    NASA Astrophysics Data System (ADS)

    Yuhara, J.; Matsui, T.

    2010-03-01

    The temperature dependence on the segregation behavior of the ferritic stainless steel single crystal (1 1 1) surface morphology has been examined by scanning tunneling microscopy (STM), Auger electron spectroscopy (AES), and low energy electron diffraction (LEED). AES clearly showed the surface segregations of chromium and nitrogen upon annealing. Nanoscale triangular chromium nitride clusters were formed around 650 °C and were regularly aligned in a hexagonal configuration. In contrast, for the ferritic stainless steel (1 1 1) surface with low-nitrogen content, chromium and carbon were found to segregate on the surface upon annealing and Auger spectra of carbon displayed the characteristic carbide peak. For the low-nitrogen surface, LEED identified a facetted surface with (2 × 2) superstructure at 650 °C. High-resolution STM identified a chromium carbide film with segregated carbon atoms randomly located on the surface. The facetted (2 × 2) superstructure changed into a (3 × 3) superstructure with no faceting upon annealing at 750 °C. Also, segregated sulfur seems to contribute to the reconstruction or interfacial relaxation between the ferritic stainless steel (1 1 1) substrate and chromium carbide film.

  7. Yttria-Stabilized Zirconia Ceramic Deposition on SS430 Ferritic Steel Grown by PLD - Pulsed Laser Deposition Method

    NASA Astrophysics Data System (ADS)

    Khalid Rivai, Abu; Mardiyanto; Agusutrisno; Suharyadi, Edi

    2017-01-01

    Development of high temperature materials are one of the key issues for the deployment of advanced nuclear reactors due to higher temperature operation. One of the candidate materials for that purpose is ceramic-coated ferritic steel that one of the functions is to be a thermal barrier coating (TBC). Thin films of YSZ (Ytrria-Stabilized Zirconia) ceramic have been deposited on a SS430 ferritic steel using Pulsed Laser Deposition (PLD) at Center For Science and Technology of Advanced Materials laboratory – National Nuclear Energy Agency of Indonesia (BATAN). The thin film was deposited with the chamber pressure range of 200-225 mTorr, the substrate temperature of 800oC, and the number of laser shots of 3×104, 6×104 and 9×104. Afterward, the samples were analyzed using Scanning Electron Microscope – Energy Dispersive X-ray Spectroscope (SEM-EDS), X-Ray Diffractometer (XRD), Atomic Force Microscope (AFM) and Vickers hardness tester. The results showed that the YSZ could homogeneously and sticky deposited on the surface of the ferritic steel. The surfaces were very smoothly formed with the surface roughness was in the range of 70 nm. Furthermore, thickness, composition of Zr4+ dan Y3+, the crystallinity, and hardness property was increased with the increasing the number of the shots.

  8. Microscale deformation of a tempered martensite ferritic steel: Modelling and experimental study of grain and sub-grain interactions

    NASA Astrophysics Data System (ADS)

    Golden, Brian J.; Li, Dong-Feng; Guo, Yina; Tiernan, Peter; Leen, Sean B.; O'Dowd, Noel P.

    2016-01-01

    In this paper, a finite-element modelling framework is presented with explicit representation of polycrystalline microstructure for a tempered martensite ferritic steel. A miniature notched specimen was manufactured from P91 steel with a 20,000 h service history and tested at room temperature under three point bending. Deformation at the microscale is quantified by electron back scattered diffraction (EBSD) before and after mechanical loading. A representative volume element was developed, based on the initial EBSD scan, and a crystal plasticity model used to account for slip-based inelastic deformation in the material. The model showed excellent correlation with the experimental data when the relevant comparisons were made.

  9. Effect of loading mode on the fracture toughness of a reduced activation ferritic/martensitic stainless steel

    SciTech Connect

    Li, H.; Hirth, J.P.; Jones, R.H.; Gelles, D.S.

    1993-09-01

    The critical J integrals of mode I (J{sub IC}), mixed-mode I/III (J{sub MC}), and mode III (J{sub IIIC}) were examined for a ferritic stainless steel (F-82H) at ambient temperature. A determination of J{sub MC} was made using modified compact-tension specimens. Different ratios of tension/shear stress were achieved by varying the principal axis of the crack plane between 0 and 55 degrees from the load line. The results showed that J{sub MC}s and tearing moduli (T{sub M}) varied with the crack angles and were lower than their mode I and mode III counterparts. Both the minimum J{sub MC} and T{sub M} occurred at a crack angle between 40 and 50 degrees, where {sigma}{sub i}/{sigma}{sub iii} was 1.2 to 0.84. The J{sub min} was 240 kJ/m{sup 2}, and ratios of J{sub IC}/J{sub min} and J{sub IIIC}/J{sub min} were about 2.1 and 1.9, respectively. Morphology of fracture surfaces was consistent with the change of J{sub MC} and T{sub M} values. While the upper shelf-fracture toughness of F-82H depends on loading mode, the J{sub min} remains very high. Other important considerations include the effect of mixed-mode loading on the DBT temperature, and effects of hydrogen and irradiation on J{sub min}.

  10. Osteoblast and monocyte responses to 444 ferritic stainless steel intended for a magneto-mechanically actuated fibrous scaffold.

    PubMed

    Malheiro, Vera N; Spear, Rose L; Brooks, Roger A; Markaki, Athina E

    2011-10-01

    The rationale behind this work is to design an implant device, based on a ferromagnetic material, with the potential to deform in vivo promoting osseointegration through the growth of a healthy periprosthetic bone structure. One of the primary requirements for such a device is that the material should be non-inflammatory and non-cytotoxic. In the study described here, we assessed the short-term cellular response to 444 ferritic stainless steel; a steel, with a very low interstitial content and a small amount of strong carbide-forming elements to enhance intergranular corrosion resistance. Two different human cell types were used: (i) foetal osteoblasts and (ii) monocytes. Austenitic stainless steel 316L, currently utilised in many commercially available implant designs, and tissue culture plastic were used as the control surfaces. Cell viability, proliferation and alkaline phosphatase activity were measured. In addition, cells were stained with alizarin red and fluorescently-labelled phalloidin and examined using light, fluorescence and scanning electron microscopy. Results showed that the osteoblast cells exhibited a very similar degree of attachment, growth and osteogenic differentiation on all surfaces. Measurement of lactate dehydrogenase activity and tumour necrosis factor alpha protein released from human monocytes indicated that 444 stainless steel did not cause cytotoxic effects or any significant inflammatory response. Collectively, the results suggest that 444 ferritic stainless steel has the potential to be used in advanced bone implant designs.

  11. Effects of Annealing Treatment Prior to Cold Rolling on Delayed Fracture Properties in Ferrite-Austenite Duplex Lightweight Steels

    NASA Astrophysics Data System (ADS)

    Sohn, Seok Su; Song, Hyejin; Kim, Jung Gi; Kwak, Jai-Hyun; Kim, Hyoung Seop; Lee, Sunghak

    2016-02-01

    Tensile properties of recently developed automotive high-strength steels containing about 10 wt pct of Mn and Al are superior to other conventional steels, but the active commercialization has been postponed because they are often subjected to cracking during formation or to the delayed fracture after formation. Here, the delayed fracture behavior of a ferrite-austenite duplex lightweight steel whose microstructure was modified by a batch annealing treatment at 1023 K (750 °C) prior to cold rolling was examined by HCl immersion tests of cup specimens, and was compared with that of an unmodified steel. After the batch annealing, band structures were almost decomposed as strong textures of {100}<011> α-fibers and {111}<112> γ-fibers were considerably dissolved, while ferrite grains were refined. The steel cup specimen having this modified microstructure was not cracked when immersed in an HCl solution for 18 days, whereas the specimen having unmodified microstructure underwent the delayed fracture within 1 day. This time delayed fracture was more critically affected by difference in deformation characteristics such as martensitic transformation and deformation inhomogeneity induced from concentration of residual stress or plastic strain, rather than the difference in initial microstructures. The present work gives a promise for automotive applications requiring excellent mechanical and delayed fracture properties as well as reduced specific weight.

  12. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  13. Annealing effects on the microstructure and coercive field of two ferritic-martensitic Eurofer steels: A comparative study

    NASA Astrophysics Data System (ADS)

    Oliveira, V. B.; Sandim, M. J. R.; Stamopoulos, D.; Renzetti, R. A.; Santos, A. D.; Sandim, H. R. Z.

    2013-04-01

    Reduced-activation ferritic-martensitic steels are promising candidates for structural applications in future nuclear fusion power plants. Oxide dispersion strengthened ODS-Eurofer and Eurofer 97 steels were cold rolled to 80% reduction in thickness and annealed in vacuum for 1 h from 200 to 1350 °C to evaluate both their thermal stability and magnetic behavior. The microstructural changes were followed by magnetic measurements, in particular the corresponding variation of the coercive field (Hc), as a function of both annealing and tempering treatments. Results show that Y2O3 nanoparticles strongly affect the mechanical properties of ODS-Eurofer steel but leave their magnetic properties fairly unchanged when compared with Eurofer-97 steel.

  14. Fracture mechanisms in dual phase steels based on the acicular ferrite + martensite/austenite microstructure

    NASA Astrophysics Data System (ADS)

    Poruks, Peter

    The fracture mechanisms of low carbon microalloyed plate steels based on the acicular ferrite + marten site/austenite microstructure (AF + M/A) are investigated. The final microstructure consists of a dispersed phase of submicron equi-axed martensite particles with a bainitic ferrite matrix. A series of plates with M/A volume fractions of 0.076--0.179 are studied. Brittle fracture is investigated by Instrumented Charpy impact testing of samples at -196°C and subsequent metallography. The M/A particles are identified as the crack nucleation sites and the cleavage fracture stress calculated to be 2400 MPa in a complete AF microstrucuture. This value is significantly larger than in steels that contain significant proportions of conventional bainite. Standard Charpy and Instrumented Charpy impact testing is conducted through a temperature range from -80 to + 22°C to study ductile fracture behaviour. The total absorbed energy is separated into energies of crack nucleation and of crack propagation. It is found that the energy of crack nucleation is weakly dependent on the volume fraction of M/A and completely independent of temperature over the range studied. The crack propagation energy varies significantly with both variables, decreasing with increased volume fraction of M/A and with decreasing temperature. The peak load in the instrumented Charpy data is used to calculate the dynamic fracture toughness, KId, which is found to be 105--120 MPa-m1/2. The void nucleation and void growth stages of ductile fracture are studied by metallographic examination of tensile bars. The sites of void nucleation are identified as inclusions and M/A particles. Voids nucleate at the M/A particles by decohesion of the particle-matrix interface. A constant void nucleation strain of epsilon = 0.90 +/- 0.05 is measured for all of the samples independent of the volume fraction of M/A. A stress-based criterion is used to predict void nucleation and the interface strength is determined to be

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

    SciTech Connect

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

    1997-04-01

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

  16. The effect of cooling speed on the structure and properties of the heat affected zone in welded compounds of ferrite-austenitic steels

    NASA Astrophysics Data System (ADS)

    Gonik, I. L.; Gurulev, D. N.; Bondareva, O. P.

    2017-02-01

    Such parameters as the maximum heating temperature, duration of stay at high temperatures, the rate of cooling influence greatly the structure and properties of the heat-affected zone of welded joints of steels and alloys. In the present work, the effect of different cooling speed upon the impact of the thermal cycle of welding on the structure, the fine structure and toughness of ferrite-austenitic steels is investigated. It is established that the cooling speed after welding has a great influence on the shock impact toughness, the phase composition and the structure of the zone of ferrite-austenitic steels.

  17. High temperature corrosion of welded ferritic stainless steel in flowing CO2 gas

    NASA Astrophysics Data System (ADS)

    Shariff, Nurul Atikah; Othman, Norinsan Kamil; Jalar, Azman; Hamid, Muhammad Azmi Abdul; Rahman, Irman Abdul

    2013-05-01

    The high temperature corrosion of welded structure of Ferritic Stainless Steel (FSS) in flowing Ar-75%CO2 gas at 700°C has been investigated. The welded structure of FSS joint using ER 308L filler metal by GTAW. The soundness of welded joint has been clarified by X-Ray CT Scan. Prior the high temperature exposure, the welded FSS compulsory passed the standard of ASME. The welded structure of FSS was heated in flowing CO2 gas for 50 h at 1 atm. The morphology and microstructure of oxide formation on welded FSS alloy was characterized by using SEM. The result shows that the different oxide morphologies were observed on parent and fusion metal. The formation of different oxide and element properties at the interface were revealed by X-Ray Diffraction. The differences of the physical condition and morphology microstructure of welded and parent metal were observed to respond to different exposure times. This phenomenon perhaps explained due to the differences of the minor alloying elements on both parent and filler metals. The high temperature corrosion behaviour was discussed in details in this paper regarding on the physical properties, morphology and the microstructure.

  18. Using nonlinear ultrasound measurements to track thermal aging in modified 9%Cr ferritic martensitic steel

    NASA Astrophysics Data System (ADS)

    Marino, Daniel; Kim, Jin-Yeon; Jacobs, Laurence J.; Ruiz, Alberto; Joo, Young-Sang

    2015-03-01

    This study investigates early thermal aging in 9%Cr ferritic martensitic (FM) steel, which is caused by the formation of second phases during high temperature exposure. This study employs a recently developed nonlinear ultrasonic technique to explore the sensitivity of the nonlinearity parameter. Experimental results show that the nonlinearity parameter is sensitive to certain changes in material's properties such as thermal embrittlement and hardness changes; therefore, it can be used as an indicator of the thermal damage. The specimens investigated are heat treated for different holding times ranging from 200h to 3000h at 650°C. Nonlinear ultrasonic experiments are conducted for each specimen using a wedge transducer to generate and an air-coupled transducer to detect Raleigh surface waves. The amplitudes of the first and second order harmonics are measured at different propagation distances and these amplitudes are used to obtain the relative nonlinearity parameter for each specimen with a different holding time. The nonlinear ultrasonic results are compared with independent mechanical measurements and metallographic images. This research proposes the nonlinear ultrasonic technique as a nondestructive evaluation tool not only to detect thermal damage in early stages, and also to qualitatively assess the stage of thermal damage.

  19. Substructural changes during hot deformation of an Fe-26Cr ferritic stainless steel

    SciTech Connect

    Gao, F.; Xu, Y.; Song, B.; Xia, K.

    2000-01-01

    Dynamic softening and substructural changes during hot deformation of a ferritic Fe-26Cr stainless steel were studied. The flow stress increased to reach a steady state in all the cases and the steady-state stress decreased with decreasing Z, the Zener-Hollomon parameter. A constant subgrain size was observed to correspond to the steady-state flow and the steady-state subgrain size increased with decreasing Z. Substructure examinations revealed that elongated, pancake-shaped subgrains formed in the early stage of deformation. Straight sub-boundaries and equiaxed subgrains developed progressively with strain, leading eventually to a stable substructure at strains greater than 0.7. During deformation at 1,100 C, dynamic recrystallization occurred by the migration and coalescence of sub-boundaries. Dynamic recovery dominated during deformation at 900 C, resulting in the formation of fine equiaxed subgrains. Based on microstructural observations, the process of substructural changes during hot deformation was described by a schematic diagram.

  20. The development of ferritic-martensitic steels with reduced long-term activation

    NASA Astrophysics Data System (ADS)

    Ehrlich, K.; Kelzenberg, S.; Röhrig, H.-D.; Schäfer, L.; Schirra, M.

    1994-09-01

    Ferritic-martensitic 9-12% CrMoVNb steels of MANET type possess a number of advantageous properties for fusion reactor application. Their optimization has led to improved creep and fracture-toughness properties. New 9-10% CrWVTa alloys have been developed by KfK/IMF in collaboration with the SAARSTAHL GmbH which have a reduced long-term activation and show in addition superior fracture toughness properties. The calculation of dose rate and other radiological parameters with the presently available FISPACT/EAF codes, extended by KfK files for sequential reactions has shown that the long-term dose-rate in these alloys is governed by the remaining 'impurity level' of Nb and the alloying elements W and Ta. Sequential reactions — though relevant for single alloying elements like Cr, Mn, V and N — provide only a second order effect in Fe-based alloys. A challenge for the future materials development is the production of alloys with the desired narrow specification of elements and impurities, which necessitates new ways of steelmaking.

  1. ECCI observations of dislocation structures around fatigue cracks in ferritic stainless steel single crystals

    NASA Astrophysics Data System (ADS)

    Taniguchi, T.; Kaneko, Y.; Hashimoto, S.

    2009-07-01

    Dislocation structures around the crack tips of ferritic stainless steel single crystals were observed with electron channelling contrast imaging (ECCI) method. The ECCI method enables us to observe dislocations lying near surface using a scanning electron microscope. Fatigue crack growth tests were conducted on compact tension (CT) specimens having loading axes of [221] and [110] directions. In the specimen having the [110] loading axis at which the fatigue crack having Mode I and II component propagated, a thin band-like structure consisting of dislocation wall array was observed ahead of the crack tip. On the other hand, the dislocation structures around the crack having Mode I and III components could be divided into three regions in the specimen with the [221] loading axis: the cell structure, the dislocation wall structure and the vein structure were observed in order of ascending distance from crack tip. Difference between the dislocation structures near the fatigue cracks could be understood from the crack mode by which edge and screw dislocation emissions from the crack tips are strongly affected.

  2. Flow behaviour and constitutive modelling of a ferritic stainless steel at elevated temperatures

    NASA Astrophysics Data System (ADS)

    Zhao, Jingwei; Jiang, Zhengyi; Zu, Guoqing; Du, Wei; Zhang, Xin; Jiang, Laizhu

    2016-05-01

    The flow behaviour of a ferritic stainless steel (FSS) was investigated by a Gleeble 3500 thermal-mechanical test simulator over the temperature range of 900-1100 °C and strain rate range of 1-50 s-1. Empirical and phenomenological constitutive models were established, and a comparative study was made on the predictability of them. The results indicate that the flow stress decreases with increasing the temperature and decreasing the strain rate. High strain rate may cause a drop in flow stress after a peak value due to the adiabatic heating. The Zener-Hollomon parameter depends linearly on the flow stress, and decreases with raising the temperature and reducing the strain rate. Significant deviations occur in the prediction of flow stress by the Johnson-Cook (JC) model, indicating that the JC model cannot accurately track the flow behaviour of the FSS during hot deformation. Both the multiple-linear and the Arrhenius-type models can track the flow behaviour very well under the whole hot working conditions, and have much higher accuracy in predicting the flow behaviour than that of the JC model. The multiple-linear model is recommended in the current work due to its simpler structure and less time needed for solving the equations relative to the Arrhenius-type model.

  3. Characterisation of Laves phase precipitation and its correlation to creep rupture strength of ferritic steels

    SciTech Connect

    Zhu, S.; Yang, M.; Song, X.L.; Tang, S.; Xiang, Z.D.

    2014-12-15

    The Laves phase precipitation process was characterised by means of field emission scanning electron microscopy to demonstrate its effect on creep rupture strength of steels with a fully ferritic matrix. To eliminate the effects of carbide and carbonitride precipitations so that the creep rupture data can be analysed exclusively in relation to the Laves phase precipitation process, an alloy Fe–9Cr–3Co–3W (wt.%) without C and N additions was used for the study. Creep rupture strengths were measured and volume fraction and particle size of Laves phase precipitates in the ruptured specimens were analysed. It was found that the creep rupture strength started to collapse (or decrease more rapidly) long before the Laves phase precipitation reached equilibrium fraction. This was related to the onset of the coarsening of Laves phase particles, which precipitated only on grain boundaries and hence contributed little to precipitation strengthening. Creep deformation had no effect either on the precipitation kinetics or on the growth kinetics of Laves phase particles. - Highlights: • Laves phase precipitation at 650 °C was characterised for Fe–9Cr–3W–3Co alloy. • Laves phase precipitated predominantly on grain boundaries. • Creep deformation had no effect on Laves phase precipitation and growth kinetics. • Creep strength started to collapse long before Laves phase precipitation is ended. • Collapse of creep strength was attributed to the coarsening of Laves phase particles.

  4. Hydrogen Embrittlement Behavior of 430 and 445NF Ferritic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Kim, Sun Mi; Chun, Young Soo; Won, Sung Yeun; Kim, Young Hwan; Lee, Chong Soo

    2013-03-01

    Hydrogen embrittlement behavior of two kinds of commercial ferritic stainless steels (STSs), 430 (UNS S43000) and 445NF (UNS S44536), was investigated by means of a series of cathodical hydrogen charging, slow strain rate tests, bending tests, and thermal desorption spectrometry analyses. The hydrogen concentration in 445NF STS was lower than that of 430 STS under identical hydrogen charging conditions because of the formation of a more passive layer. In addition, 445NF STS exhibited a larger passive range in the potentiodynamic polarization curve. However, resistance to hydrogen embrittlement of 445NF STS was inferior to that of 430 STS because of precipitation of the Laves phase at grain boundaries of the former at annealing temperatures of 873 K to 1123 K (600 °C to 850 °C). Crack propagation was found to occur along the interface between the Laves phase and the matrix. For 445NF STS, dissolution of the Laves phase by solution heat treatment at 1273 K (1000 °C) followed by quenching was effective in terms of suppressing degradation of its mechanical properties and formability, which were related to hydrogen embrittlement.

  5. Formation and Oxidation Performance of Low-Temperature Pack Aluminide Coatings on Ferritic-Martensitic Steels

    SciTech Connect

    Bates, Brian; Wang, Y. Q.; Zhang, Ying; Pint, Bruce A

    2009-01-01

    A pack cementation process was developed to coat commercial 9% Cr ferritic-martensitic steel T91 at temperatures below its normal tempering temperature to avoid any potential detrimental effect on the mechanical properties of the coated alloy. In order to prevent the formation of Fe{sub 2}Al{sub 5} coatings, the Al activity in the pack cementation process was reduced by substituting the pure Al masteralloy with binary Cr-Al masteralloys containing either 15 or 25 wt.% Al. When the Cr-25Al masteralloy was used, a duplex coating was formed at 700 C, consisting of a thin Fe{sub 2}Al{sub 5} outer layer and an inner layer of FeAl. With the Cr-15Al masteralloy, an FeAl coating of {approx} 12 {micro}m thick was achieved at 700 C. The pack aluminide coatings fabricated at 700 C are being evaluated in air + 10 vol.% H{sub 2}O at 650 C and 700 C to determine their long-term oxidation performance.

  6. Elevated-temperature tensile and creep properties of several ferritic stainless steels

    NASA Technical Reports Server (NTRS)

    Whittenberger, J. D.

    1977-01-01

    The elevated-temperature mechanical properties of several ferritic stainless steels were determined. The alloys evaluated included Armco 18SR, GE 1541, and NASA-18T-A. Tensile and creep strength properties at 1073 and 1273 K and residual room temperature tensile properties after creep testing were measured. In addition, 1273 K tensile and creep tests and residual property testing were conducted with Armco 18SR and GE 1541 which were exposed for 200 hours to a severe oxidizing environment in automotive thermal reactors. Aside from the residual tensile properties for Armco 18SR, prior exposure did not affect the mechanical properties of either alloy. The 1273 K creep strength parallel to the sheet-rolling direction was similar for all three alloys. At 1073 K, NASA-18T-A had better creep strength than either Armco 18SR or GE 1541. NASA-18T-A possesses better residual properties after creep testing than either Armco 18SR or Ge 1541.

  7. Corrosion behavior of ferritic stainless steel with 15wt% chromium for the automobile exhaust system

    NASA Astrophysics Data System (ADS)

    Li, Hua-bing; Jiang, Zhou-hua; Feng, Hao; Zhu, Hong-chun; Sun, Bin-han; Li, Zhen

    2013-09-01

    The effect of chloride ion concentration, pH value, and grain size on the pitting corrosion resistance of a new ferritic stainless steel with 15wt% Cr was investigated using the anodic polarization method. The semiconducting properties of passive films with different chloride ion concentrations were performed using capacitance measurement and Mott-Schottky analysis methods. The aging precipitation and intergranular corrosion behavior were evaluated at 400-900°C. It is found that the pitting potential decreases when the grain size increases. With the increase in chloride ion concentration, the doping density and the flat-bland potential increase but the thickness of the space charge layer decreases. The pitting corrosion resistance increases rapidly with the decrease in pH value. Precipitants is identified as Nb(C,N) and NbC, rather than Cr-carbide. The intergranular corrosion is attributed to the synergistic effects of Nb(C,N) and NbC precipitates and Cr segregation adjacent to the precipitates.

  8. Effect of Process Parameters on Microstructure and Hardness of Oxide Dispersion Strengthened 18Cr Ferritic Steel

    NASA Astrophysics Data System (ADS)

    Nagini, M.; Vijay, R.; Rajulapati, Koteswararao V.; Rao, K. Bhanu Sankara; Ramakrishna, M.; Reddy, A. V.; Sundararajan, G.

    2016-08-01

    Pre-alloyed ferritic 18Cr steel (Fe-18Cr-2.3W-0.3Ti) powder was milled with and without nano-yttria in high-energy ball mill for varying times until steady-state is reached. The milled powders were consolidated by upset forging followed by hot extrusion. Microstructural changes were examined at all stages of processing (milling, upset forging, and extrusion). In milled powders, crystallite size decreases and hardness increases with increasing milling time reaching a steady-state beyond 5 hours. The size of Y2O3 particles in powders decreases with milling time and under steady-state milling conditions; the particles either dissolve in matrix or form atomic clusters. Upset forged sample consists of unrecrystallized grain structure with few pockets of fine recrystallized grains and dispersoids of 2 to 4 nm. In extruded and annealed rods, the particles are of cuboidal Y2Ti2O7 at all sizes and their size decreased from 15 nm to 5 nm along with significant increase in number density. The oxide particles in ODS6 are of cuboidal Y2Ti2O7 with diamond cubic crystal structure ( Fd bar{3} m) having a lattice parameter of 10.1 Å and are semicoherent with the matrix. The hardness values of extruded and annealed samples predicted by linear summation model compare well with measured values.

  9. Formation of nano-size oxide particles and δ-ferrite at elevated temperature in 9Cr-ODS steel

    NASA Astrophysics Data System (ADS)

    Kim, Sawoong; Ohtsuka, Satoshi; Kaito, Takeji; Yamashita, Shinichiro; Inoue, Masaki; Asayama, Tai; Shobu, Takahisa

    2011-10-01

    Excellent high-temperature strength and resistance to radiation damage of 9Cr Oxide Dispersion Strengthened (9Cr-ODS) martensitic steel have been realized by nano-size Y-Ti-O complex oxide particles dispersed in the matrix and a dual phase structure consisting of α'-martensite and δ-ferrite. These are produced by mechanically alloying Fe-Cr-Ti powders with Y 2O 3 followed by a hot-consolidation process. Therefore, the hot-consolidation process is the issue to be clarified for the formation of nano-size oxide particle and δ-ferrite. The temperature dependence of the formation and development of nano-size oxide particles and δ-ferrite using mechanically alloyed 9Cr-ODS raw powder were investigated applying X-ray Diffraction and Small Angle X-ray Scattering measurement at SPring-8 and by Electron Probe Micro Analysis. In situ heating measurement techniques with XRD and SAXS enabled real-time observation of phase transformations and allowed correlation between formation of nano-size oxide particle and δ-ferrite.

  10. PERFORMANCE IMPROVEMENT OF CREEP-RESISTANT FERRITIC STEEL WELDMENTS THROUGH THERMO-MECHANICAL TREATMENT AND ALLOY DESIGN

    SciTech Connect

    Yamamoto, Yukinori; Babu, Prof. Sudarsanam Suresh; Shassere, Benjamin; Yu, Xinghua

    2016-01-01

    Two different approaches have been proposed for improvement of cross-weld creep properties of the high temperature ferrous structural materials for fossil-fired energy applications. The traditional creep strength-enhanced ferritic (CSEF) steel weldments suffer from Type IV failures which occur at the fine-grained heat affected zone (FGHAZ). In order to minimize the premature failure at FGHAZ in the existing CSEF steels, such as modified 9Cr-1Mo ferritic-martensitic steels (Grade 91), a thermo-mechanical treatment consisting of aus-forging/rolling and subsequent aus-aging is proposed which promotes the formation of stable MX carbonitrides prior to martensitic transformation. Such MX remains undissolved during welding process, even in FGHAZ, which successfully improves the cross-weld creep properties. Another approach is to develop a new fully ferrtic, creep-resistant FeCrAl alloy which is essentially free from Type IV failure issues. Fe-30Cr-3Al base alloys with minor alloying additions were developed which achieved a combination of good oxidation/corrosion resistance and improved tensile and creep performance comparable or superior to Grade 92 steel.

  11. Tensile Properties of Medium Mn Steel with a Bimodal UFG α + γ and Coarse δ-Ferrite Microstructure

    NASA Astrophysics Data System (ADS)

    Lee, Seonjong; Shin, Sunmi; Kwon, Minhyeok; Lee, Kyooyoung; De Cooman, Bruno C.

    2017-04-01

    While the tensile strength and elongation obtained for medium Mn steel would appear to make it a candidate material in applications which require formable ultra-high strength materials, many secondary aspects of the microstructure-properties relationships have not yet been given enough attention. In this contribution, the microstructural and tensile properties of medium Mn steel with a bimodal microstructure consisting of an ultra-fine grained ferrite + austenite constituent and coarse-grained delta-ferrite are therefore reviewed in detail. The tensile properties of ultra-fine-grained intercritically annealed medium Mn steel reveal a complex dependence on the intercritical annealing temperature. This dependence is related to the influence of the intercritical annealing temperature on the activation of the plasticity-enhancing mechanisms in the microstructure. The kinetics of deformation twinning and strain-induced transformation in the ultra-fine grained austenite play a prominent role in determining the strain hardening of medium Mn steel. While excellent strength-ductility combinations are obtained when deformation twinning and strain-induced transformation occur gradually and in sequence, large elongations are also observed when strain-induced transformation plasticity is not activated. In addition, the localization of plastic flow is observed to occur in samples after intercritical annealing at intermediate temperatures, suggesting that both strain hardening and strain rate sensitivity are influenced by the properties of the ultra-fine-grained austenite.

  12. Microstructural characterization of weld joints of 9Cr reduced activation ferritic martensitic steel fabricated by different joining methods

    SciTech Connect

    Thomas Paul, V.; Saroja, S.; Albert, S.K.; Jayakumar, T.; Rajendra Kumar, E.

    2014-10-15

    This paper presents a detailed electron microscopy study on the microstructure of various regions of weldment fabricated by three welding methods namely tungsten inert gas welding, electron beam welding and laser beam welding in an indigenously developed 9Cr reduced activation ferritic/martensitic steel. Electron back scatter diffraction studies showed a random micro-texture in all the three welds. Microstructural changes during thermal exposures were studied and corroborated with hardness and optimized conditions for the post weld heat treatment have been identified for this steel. Hollomon–Jaffe parameter has been used to estimate the extent of tempering. The activation energy for the tempering process has been evaluated and found to be corresponding to interstitial diffusion of carbon in ferrite matrix. The type and microchemistry of secondary phases in different regions of the weldment have been identified by analytical transmission electron microscopy. - Highlights: • Comparison of microstructural parameters in TIG, electron beam and laser welds of RAFM steel • EBSD studies to illustrate the absence of preferred orientation and identification of prior austenite grain size using phase identification map • Optimization of PWHT conditions for indigenous RAFM steel • Study of kinetics of tempering and estimation of apparent activation energy of the process.

  13. Sensitization Behavior of Type 409 Ferritic Stainless Steel: Confronting DL-EPR Test and Practice W of ASTM A763

    NASA Astrophysics Data System (ADS)

    Scalise, Taís Campos; de Oliveira, Mara Cristina Lopes; Sayeg, Isaac Jamil; Antunes, Renato Altobelli

    2014-06-01

    Stainless steels employed for manufacturing automotive exhaust systems must withstand severe thermal cycles, corrosive environment due to urea decomposition, and welding operations. AISI 409 ferritic stainless steel can be considered a low-cost alternative for this application. However, depending on the manufacturing conditions during welding cycles, this material can be sensitized due to the precipitation of chromium carbides at grain boundaries. In this work, the intergranular corrosion resistances of the AISI 409 ferritic stainless steel were evaluated after annealing at 300, 500, and 700 °C for 2, 4, and 6 h. Solution-annealed samples were also tested for comparison purposes. Two methodologies were used to assess the sensitization behavior of the 409 stainless steel samples: the first one was based on the ASTM A763 (practice W), while the second one was based on the double-loop electrochemical potentiodynamic reactivation test. It was possible to identify that the annealing treatment performed at 500 °C was more critical to the occurrence of intergranular corrosion.

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

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

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

  15. Effect of Coiling Temperature on Microstructure and Tensile Behavior of a Hot-Rolled Ferritic Lightweight Steel

    NASA Astrophysics Data System (ADS)

    Wang, Junfeng; Yang, Qi; Wang, Xiaodong; Wang, Li

    2016-12-01

    Effects of coiling temperature (CT) ranging from 673 K to 973 K (400 °C to 700 °C) on microstructure and tensile property of a hot-rolled ferritic lightweight steel containing 0.35 wt pct C and 4.1 wt pct Al are investigated in the present study. Basically, the microstructure of the hot-rolled steel is composed of δ-ferrite grain bands and secondary phase bands which are originated from the decomposition of antecedent austenite. The secondary phase band is a bainite band at coiling temperatures (CTs) lower than 723 K (450 °C). More specifically, the bainite band mainly consists of lower bainite together with blocky retained austenite at the CT of 673 K (400 °C), while it primarily contains carbide-free bainite being an aggregate of lath-shaped ferrite and austenite at the CT of 723 K (450 °C). The secondary phase band is a carbide band which mainly contains a pearlite structure at CTs higher than 773 K (500 °C). There are three types of carbides in the steel matrix: transitional ɛ-carbide present inside lower bainite, cementite present within carbide bands as well as at the boundaries between carbide bands and δ-ferrite bands, and κ-carbide present at δ-ferrite grain boundaries which is clearly seen at CTs higher than 773 K (500 °C). The volume fraction of retained austenite reaches the peak value of 9.6 pct at the CT of 723 K (450 °C), and abruptly drops to zero when the CTs are higher than 773 K (500 °C). Lath-shaped retained austenite with a higher volume fraction induces significant enhancement of elongation through the TRIP effect, leading to a uniform elongation of 25 pct and an elongation-to-failure of 32 pct at the CT of 723 K (450 °C). Crack initiation and propagation inside the tested specimens are tracked and fracture surface is observed to help understand the deformation and fracture behavior of the hot-rolled steel.

  16. Changes in magnetic properties of neutron irradiated RPV steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Ok, C.I.; Kim, J.W.; Kim, H.C.

    1998-07-01

    Conventional magnetic parameters and Barkhausen noise have been measured in pressure vessel steel samples both as-received and irradiated with doses of up to 10{sup 18} n/cm{sup 2}. The conventional magnetic parameters, i.e., coercive force, remanence and maximum induction did not change significantly with irradiation, whereas the Barkhausen noise amplitude and energy during a magnetization cycle decreased markedly with irradiation dose. A three stage variation of Barkhausen noise with neutron dose was observed in the present work, namely an initial decrease, a near plateau and rapid decrease. The three stage variation with neutron dose is in qualitative agreement with computer simulations of the radiation damage process performed by Beeler. The hardness also varied in three stages in a reverse manner with transition at the same doses.

  17. Residual Stress Analysis in Girth-welded Ferritic and Austenitic Steel Pipes Using Neutron and X-Ray Diffraction

    SciTech Connect

    Hempel, Nico; Bunn, Jeffrey R; Nitschke-Pagel, Thomas; Payzant, E Andrew; Dilger, Klaus

    2016-01-01

    This paper is dedicated to the thorough experimental analysis of the residual stresses in the vicinity of tubular welds and the mechanisms involved in their formation. Pipes made of a ferritic-pearlitic structural steel and an austenitic stainless steel are investigated in this study. The pipes feature a similar geometry and are MAG welded with two passes and comparable parameters. Residual strain mappings are carried out using X-ray and neutron diffraction. The combined use of both techniques permits both near-surface and through-wall analyses of the residual stresses. The findings allow for a consistent interpretation of the mechanisms accounting for the formation of the residual stress fields due to the welding process. Since the results are similar for both materials, it can be concluded that residual stresses induced by phase transformations, which can occur in the structural steel, play a minor role in this regard.

  18. Characteristics and Modification of Non-metallic Inclusions in Titanium-Stabilized AISI 409 Ferritic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Kruger, Dirk; Garbers-Craig, Andrie

    2017-02-01

    This study describes an investigation into the improvement of castability, final surface quality and formability of titanium-stabilized AISI 409 ferritic stainless steel on an industrial scale. Non-metallic inclusions found in this industrially produced stainless steel were first characterized using SEM-EDS analyses through the INCA-Steel software platform. Inclusions were found to consist of a MgO·Al2O3 spinel core, which acted as heterogeneous nucleation site for titanium solubility products. Plant-scale experiments were conducted to either prevent the formation of spinel, or to modify it by calcium treatment. Modification to spherical dual-phase spinel-liquid matrix inclusions was achieved with calcium addition, which eliminated submerged entry nozzle clogging for this grade. Complete modification to homogeneous liquid calcium aluminates was achieved at high levels of dissolved aluminum. A mechanism was suggested to explain the extent of modification achieved.

  19. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    SciTech Connect

    Tan, Lizhen; Yang, Ying; Sridharan, K.

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  20. Neutron Irradiation Effects on the Mechanical Properties of HY-80 Steel

    DTIC Science & Technology

    1986-12-01

    properties of HY-80 steel to 750*F for the indicated irradiation conditions (Ref. [15]) ..................... 52 Figure 26 DBTT changes of HY-80 steels as...to-brittle transition temperature ( DBTT ) and generally it is in the range of -50 to +20 degress Celsius for unirradiated mild steels , [2]. In the...ductility has on the ductile-brittle transition temperature ( DBTT ). For steels irradiated at less than 450-F, a relatively consistent increase in nil

  1. Load partitioning between ferrite/martensite and dispersed nanoparticles of a 9Cr ferritic/martensitic (F/M) ODS steel at high temperatures

    SciTech Connect

    Zhang, Guangming; Mo, Kun; Miao, Yinbin; Liu, Xiang; Almer, Jonathan; Zhou, Zhangjian; Stubbins, James F.

    2015-06-18

    In this study, a high-energy synchrotron radiation X-ray technique was used to investigate the tensile deformation processes of a 9Cr-ODS ferritic/martensitic (F/M) steel at different temperatures. Two minor phases within the 9Cr-ODS F/M steel matrix were identified as Y2Ti2O7 and TiN by the high-energy X-ray diffraction, and confirmed by the analysis using energy dispersive X-ray spectroscopy (EDS) of scanning transmission electron microscope (STEM). The lattice strains of the matrix and particles were measured through the entire tensile deformation process. During the tensile tests, the lattice strains of the ferrite/martensite and the particles (TiN and Y2Ti2O7) showed a strong temperature dependence, decreasing with increasing temperature. Analysis of the internal stress at three temperatures showed that the load partitioning between the ferrite/martensite and the particles (TiN and Y2Ti2O7) was initiated during sample yielding and reached to a peak during sample necking. At three studied temperatures, the internal stress of minor phases (Y2Ti2O7 and TiN) was about 2 times that of F/M matrix at yielding position, while the internal stress of Y2Ti2O7 and TiN reached about 4.5-6 times and 3-3.5 times that of the F/M matrix at necking position, respectively. It indicates that the strengthening of the matrix is due to minor phases (Y2Ti2O7 and TiN), especially Y2Ti2O7 particles. Although the internal stresses of all phases decreased with increasing temperature from RT to 600 degrees C, the ratio of internal stresses of each phase at necking position stayed in a stable range (internal stresses of Y2Ti2O7 and TiN were about 4.5-6 times and 3-3.5 times of that of F/M matrix, respectively). The difference between internal stress of the F/M matrix and the applied stress at 600 degrees C is slightly lower than those at RI and 300 degrees C, indicating that the nanoparticles still have good strengthening effect at 600 degrees C. (C) 2015 Elsevier B.V. All rights reserved.

  2. Effect of Strain-Induced Age Hardening on Yield Strength Improvement in Ferrite-Austenite Duplex Lightweight Steels

    NASA Astrophysics Data System (ADS)

    Song, Hyejin; Lee, Seok Gyu; Sohn, Seok Su; Kwak, Jai-Hyun; Lee, Sunghak

    2016-11-01

    Ferrite-austenite lightweight steels showing TRansformation-induced plasticity were developed by varying the aging temperature with or without prestraining, and their effects on tensile properties were investigated in relation with microstructural evolution of carbide formation. The aged steels contained austenite, pearlite, and martensite in the ferrite matrix, and the austenite volume fraction decreased with the increasing aging temperature because some austenite grains decomposed to pearlites. This austenite decomposition to pearlite was favorable for the improvement of yield strength, but negatively influenced overall tensile properties. The prestraining promoted the austenite decomposition by a diffusion-controlled phase transformation, and changed the morphology of the cementite from a long lamellar shape to a densely agglomerated particle shape. In order to obtain the large increase in yield strength as well as excellent combination of strength and ductility, the strain-induced aging treatment, i.e., prestraining followed by aging, is important like in the prestrained and 673 K (400 °C)-aged steel. This large increase in yield strength, in spite of a reduction of elongation (65 to 43 pct), was basically attributed to an appropriate amount of decomposition of austenite to pearlite ( e.g., 4 vol pct), while having sufficient austenite to martensite transformation ( e.g., 14.5 vol pct martensite).

  3. High Temperature Strengthening in 12Cr-W-Mo Steels by Controlling the Formation of Delta Ferrite

    NASA Astrophysics Data System (ADS)

    Wang, Shushen; Chang, Li; Lin, Deye; Chen, Xiaohua; Hui, Xidong

    2014-09-01

    Novel 12Cr-W-Mo-Co heat resistance steels (HRSs) with excellent mechanical properties have been developed for ultra-supercritical (USC) applications above 923 K (650 °C). The thermal analysis of the present steels indicates that the remelting temperature of secondary phases is increased by Co alloying, resulting in the improvement of microstructural stability. Delta ferrite in these HRSs is completely suppressed as the content of Co is increased up to 5 pct. The room temperature tensile strength (TS), yield strength (YS), and the elongation (EL) of the HRS with 5 pct Co reach 887.9, 652.6 MPa, and 21.07 pct, respectively. At 948 K (675 °C), the TS and YS of the HRS with 5 pct Co attain 360 and 290 MPa, respectively, which are higher than those of T/P122 steel by 27.4 and 22.1 pct, respectively. TEM study of the microstructure confirmed that the strengthening effects for these 12Cr-W-Mo-Co HRSs are attributed to the suppression of delta ferrite, the formation of fine martensitic laths with substructure, dislocation networks and walls, and the precipitation of second nanoscale phases.

  4. The Kinetics of Dislocation Loop Formation in Ferritic Alloys Through the Aggregation of Irradiation Induced Defects

    NASA Astrophysics Data System (ADS)

    Kohnert, Aaron Anthony

    The mechanical properties of materials are often degraded over time by exposure to irradiation environments, a phenomenon that has hindered the development of multiple nuclear reactor design concepts. Such property changes are the result of microstructural changes induced by the collision of high energy particles with the atoms in a material. The lattice defects generated in these recoil events migrate and interact to form extended damage structures. This study has used theoretical models based on the mean field chemical reaction rate theory to analyze the aggregation of isolated lattice defects into larger microstructural features that are responsible for long term property changes, focusing on the development of black dot damage in ferritic iron based alloys. The purpose of such endeavors is two-fold. Primarily, such models explain and quantify the processes through which these microstructures form. Additionally, models provide insight into the behavior and properties of the point defects and defect clusters which drive general microstructural evolution processes. The modeling effort presented in this work has focused on physical fidelity, drawing from a variety of sources of information to characterize the unobservable defect generation and agglomeration processes that give rise to the observable features reported in experimental data. As such, the models are based not solely on isolated point defect creation, as is the case with many older rate theory approaches, but instead on realistic estimates of the defect cluster population produced in high energy cascade damage events. Experimental assessments of the microstructural changes evident in transmission electron microscopy studies provide a means to measure the efficacy of the kinetic models. Using common assumptions of the mobility of defect clusters generated in cascade damage conditions, an unphysically high density of damage features develops at the temperatures of interest with a temperature dependence

  5. The influence of fine ferrite formation on the γ/α interface, fine bainite and retained austenite in a thermomechanically-processed transformation induced plasticity steel

    DOE PAGES

    Timokhina, Ilana B.; Miller, Michael K.; Beladi, Hossein; ...

    2016-03-03

    We subjected a Fe–0.26C–1.96Si–2Mn with 0.31Mo (wt%) steel to a novel thermomechanical processing route to produce fine ferrite with different volume fractions, bainite, and retained austenite. In two types of fine ferrites were found to be: (i) formed along prior austenite grain boundaries, and (ii) formed intragranularly in the interior of austenite grains. An increase in the volume fraction of fine ferrite led to the preferential formation of blocky retained austenite with low stability, and to a decrease in the volume fraction of bainite with stable layers of retained austenite. Moreover, the difference in the morphology of the bainitic ferritemore » and the retained austenite after different isothermal ferrite times was found to be responsible for the deterioration of the mechanical properties. The segregation of Mn, Mo, and C at distances of 2–2.5 nm from the ferrite and retained austenite/martensite interface on the retained austenite/martensite site was observed after 2700 s of isothermal hold. Finally, it was suggested that the segregation occurred during the austenite-to-ferrite transformation, and that this would decrease the interface mobility, which affects the austenite-to-ferrite transformation and ferrite grain size.« less

  6. Variation in Mechanical Properties and Heterogeneity in Microstructure of High-Strength Ferritic Steel During Mill Trial

    NASA Astrophysics Data System (ADS)

    Ghosh, M.; Barat, K.; Das, S. K.; Ravi Kumar, B.; Pramanick, A. K.; Chakraborty, J.; Das, G.; Hadas, S.; Bharathy, S.; Ray, S. K.

    2014-06-01

    HS600 and HS800 are two new generation, high-strength advanced ferritic steels that find widespread application in automobiles. During commercial production of the same grades with different thicknesses, it has been found that mechanical properties like tensile strength and stretchability varied widely and became inconsistent. In the current endeavor, two different thicknesses have been chosen from a mill trial sample of HS600 and HS800. An in-depth structural characterization was carried out for all four alloys to explain the variation in their respective mechanical and shear punch properties. The carbon content was smaller and Ti + Mo quantity was higher in case of HS800 with respect to HS600. The microstructure of both steels consisted of the dispersion of (Ti,Mo)C in a ferrite matrix. The grain size of HS800 was little larger than HS600 due to an increased coiling temperature (CT) of the former in comparison to the latter. It was found that in case of same grade of steel with a different thickness, a variation in microstructure occurred due to change in strain, CT, and cooling rate. The strength and stretch formability of these two alloys were predominantly governed by a microalloyed carbide. In this respect, carbides with a size range above 5 nm were responsible for loosing coherency with ferrite matrix. In case of HS600, both ≤5 and >5-nm size (Ti,Mo)C precipitates shared a nearly equal fraction of microalloyed precipitates. However, for HS800, >5-nm size (Ti,Mo)C carbide was substantially higher than ≤5-nm size alloy carbides. The ultimate tensile strength and yield strength of HS800 was superior to that of HS600 owing to a higher quantity of microalloyed carbide with a decreased column width and interparticle distance. A higher degree of in-coherency of HS800 made the alloy prone to crack formation with low stretchability.

  7. Microstructural study of irradiated isotopically tailored F82H steel

    NASA Astrophysics Data System (ADS)

    Wakai, E.; Miwa, Y.; Hashimoto, N.; Robertson, J. P.; Klueh, R. L.; Shiba, K.; Abiko, K.; Furuno, S.; Jitsukawa, S.

    2002-12-01

    The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250-400 °C to 2.8-51 dpa in HFIR has been examined using isotopes of 54Fe or 10B. Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α '-precipitates on dislocation loops.

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

    NASA Astrophysics Data System (ADS)

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

    2016-07-01

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

  9. Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

    SciTech Connect

    Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

    2012-08-28

    The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to its superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.

  10. Morphological Stability of δ-Ferrite/γ Interphase Boundary in Carbon Steel

    NASA Astrophysics Data System (ADS)

    Chang, Guowei; Chen, Shuying; Yue, Xudong; Li, Qingchun

    2017-01-01

    The morphological changes of the δ-ferrite/γ interphase boundary have been observed in situ with a high-temperature confocal scanning laser microscope (HTCSLM) during δ/γ transformations (δ → γ and γ → δ) of Fe-0.06 wt pct C-0.6 wt pct Mn alloy, and a kinetic equation of morphological stability of δ-ferrite/γ interphase boundary has been established. Thereafter, the criterion expression for morphological stability of δ-ferrite/γ interphase boundary was established and discussed, and the critical migration speeds of δ-ferrite/γ interphase boundaries are calculated in Fe-C, Fe-Ni, and Fe-Cr alloys. The results indicate that the δ-ferrite/γ interphase boundary is very stable and nearly remains absolute planar all the time during γ → δ transformation in Fe-C alloy. The δ-ferrite/γ interphase boundary remains basically planar during δ → γ transformation when the migration speed is lower than 0.88 μm/s, and the interphase boundary will be unstable and exhibit a finger-like morphology when the migration speed is higher than 0.88 μm/s. The morphological stability of δ-ferrite/γ interphase boundary is primarily controlled by the interface energy and the solute concentration gradient at the front of the boundary. During the constant temperature phase transformation, an opposite temperature gradient on both sides of δ-ferrite/γ interphase boundary weakens the steady effect of the temperature gradient on the boundary. The theoretical analysis of the morphological stability of the δ-ferrite/γ interphase boundary is coincident with the observed experimental results utilizing the HTCSLM. There is a good agreement between the theoretical calculation of the critical moving velocities of δ-ferrite/γ interphase boundaries and the experimental results.

  11. Morphological Stability of δ-Ferrite/ γ Interphase Boundary in Carbon Steel

    NASA Astrophysics Data System (ADS)

    Chang, Guowei; Chen, Shuying; Yue, Xudong; Li, Qingchun

    2017-04-01

    The morphological changes of the δ-ferrite/ γ interphase boundary have been observed in situ with a high-temperature confocal scanning laser microscope (HTCSLM) during δ/ γ transformations ( δ → γ and γ → δ) of Fe-0.06 wt pct C-0.6 wt pct Mn alloy, and a kinetic equation of morphological stability of δ-ferrite/ γ interphase boundary has been established. Thereafter, the criterion expression for morphological stability of δ-ferrite/ γ interphase boundary was established and discussed, and the critical migration speeds of δ-ferrite/ γ interphase boundaries are calculated in Fe-C, Fe-Ni, and Fe-Cr alloys. The results indicate that the δ-ferrite/ γ interphase boundary is very stable and nearly remains absolute planar all the time during γ → δ transformation in Fe-C alloy. The δ-ferrite/ γ interphase boundary remains basically planar during δ → γ transformation when the migration speed is lower than 0.88 μm/s, and the interphase boundary will be unstable and exhibit a finger-like morphology when the migration speed is higher than 0.88 μm/s. The morphological stability of δ-ferrite/ γ interphase boundary is primarily controlled by the interface energy and the solute concentration gradient at the front of the boundary. During the constant temperature phase transformation, an opposite temperature gradient on both sides of δ-ferrite/ γ interphase boundary weakens the steady effect of the temperature gradient on the boundary. The theoretical analysis of the morphological stability of the δ-ferrite/ γ interphase boundary is coincident with the observed experimental results utilizing the HTCSLM. There is a good agreement between the theoretical calculation of the critical moving velocities of δ-ferrite/ γ interphase boundaries and the experimental results.

  12. Cr-W-V bainitic/ferritic steel with improved strength and toughness and method of making

    DOEpatents

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

    1994-03-08

    This work describes a high strength, high toughness bainitic/ferritic steel alloy comprising about 2.75% to 4.0% chromium, about 2.0% to 3.5% tungsten, about 0.10% to 0.30% vanadium, and about 0.1% to 0.15% carbon with the balance iron, wherein the percentages are by total weight of the composition, wherein the alloy having been heated to an austenitizing temperature and then cooled at a rate sufficient to produce carbide-free acicular bainite. 15 figures.

  13. Results of crack-arrest tests on irradiated a 508 class 3 steel

    SciTech Connect

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  14. Onset of plasticity of helium-implanted ferritic/martensitic steels during nanoindentation

    NASA Astrophysics Data System (ADS)

    Chen, Siwei; Wang, Yongming; Hashimoto, Naoyuki; Ohnuki, Somei

    2014-07-01

    The onset of plasticity during nanoindentation is a new method to investigate the irradiation damage of structural materials in fission and fusion reactors. In this paper, nanoindentation experiment was carried out to helium implanted F82H-IEA and nano-sized oxide dispersion strengthened F82H-ODS steels for studying the elastic-plastic transition at a constant loading rate. The onset of plasticity shifted after helium implantation. By a statistical thermal activation model, activation volume was extracted to discuss the strength of barrier for dislocation motion. The results reveal an increase in the pinning force and number density of effective obstacles for dislocation motion in He-implanted F82H-IEA, and a decrease in the local pinning force without changing the density of effective obstacles in He-implanted F82H-ODS.

  15. Stress corrosion cracking of neutron irradiated type 304 stainless steels

    SciTech Connect

    Tsukada, Takashi; Miwa, Yukio; Nakajima, Hajime

    1995-12-31

    To study the effect of minor elements on the irradiation assisted stress corrosion cracking (IASCC), a high purity type 304L stainless steel and its heats doped minor elements, Si, P, S, C and Ti were irradiated at 513 K to 6.7 {times} 10{sup 24} n/m{sup 2} (E>1 MeV). After irradiation, susceptibility to the stress corrosion cracking (SCC) was evaluated by the slow strain rate tensile (SSRT) test in an oxygenated high purity water at 573 K, and the fracture surface of the specimens was examined by the scanning electron microscopy (SEM). The specimens showed high susceptibilities to SCC. Specimens without addition of C showed the intergranular type SCC (IGSCC), while C doped specimens generally failed by the transgranular type SCC (TGSCC). Addition of C into the hi purity alloy caused an enhancement of radiation hardening and a remarkable increase in maximum stress during SSRT test. Enrichment of Si changed specifically tensile properties after irradiation and decreased maximum stress and improved total elongation. Addition of S greatly enhanced the IASCC susceptibility and addition of P seemed to be beneficial for suppressing it. An effect of Ti was not prominent in the alloy with a high C concentration.

  16. Effect of deformation on the austenite-to-ferrite transformation in a plain carbon and two microalloyed steels

    NASA Astrophysics Data System (ADS)

    Essadiqi, E.; Jonas, J. J.

    1988-03-01

    Isothermal compression tests were carried out on three steels: (i) a plain C, (ii) a Mo, and (iii) a Mo-Nb-V microalloyed grade in order to study the effect of deformation on the austenite-to-ferrite transformation. Dynamic TTT (DTTT) curves were determined, which show clearly the extent to which deformation accelerates the decomposition of austenite. The latter effect is diminished by the addition of the alloying elements Mo, Nb, and V and is further reduced as the temperature is increased. The substitutional elements Mo, Nb, and V appear to reduce the nucleation rate through reduction of the austenite grain boundary energy. The growth rate is also reduced by these elements, apparently through the solute drag-like effect. The microstructural results indicate that the ferrite formed under dynamic conditions becomes more homogeneous and finer when the strain rate or the temperature is increased. Under static conditions, increasing the prestrain or the prestraining strain rate accelerates the γ-to-α transformation and reduces the mean grain size of the ferrite, although the highest transformation rate is still associated with the dynamic case.

  17. Corrosion of ferritic steels by molten lithium: Influence of competing thermal gradient mass transfer and surface product reactions

    SciTech Connect

    Tortorelli, P.F.

    1987-10-01

    An Fe-12Cr-1MoVW steel was exposed to thermally convective lithium for 6962 h. Results showed that the weight change profile of Fe-12Cr-1MoVW steel changed substantially as the maximum loop temperature was raised from 500 to 600/sup 0/C. Furthermore, for a particular loop experiment, changes in the structure and composition of the exposed surfaces did not reflect typical thermal gradient mass transfer effects for all elements: the surface concentration of chromium was often a maximum at intermediate temperatures, while nickel (present at low concentrations in the starting material) tended to be transported to the coldest part of the loop. Such data were interpreted in terms of a qualitative model in which there are different dominant reactions or the various constituents of the ferritic steels (surface product formation involving nitrogen and/or carbon and solubility-driven elemental transport). This competition among different reactions is important in evaluating overall corrosion behavior and the effects of temperature. The overall corrosion rate of the 12Cr-1MoVW steel was relatively low when compared to that for austenitic stainless steel exposed under similar conditions.

  18. Influence of grain refinement on the electrochemical behavior of AISI 430 ferritic stainless steel in an alkaline solution

    NASA Astrophysics Data System (ADS)

    Fattah-alhosseini, A.; Vafaeian, S.

    2016-01-01

    In this paper, the effect of grain refinement on the electrochemical behavior of AISI 430 ferritic stainless steel in 0.1 M NaOH solution was investigated. Potentiodynamic polarization curves showed that fine-grained samples have less corrosion potential, higher corrosion current density, and less protective passive film in comparison to coarse-grained samples. Electrochemical impedance spectroscopy (EIS) analysis revealed that implementing the thermomechanical operation led to lower polarization resistance. Also, Mott-Schottky analysis revealed that the passive films on both fine-grained and coarse-grained samples behave as n-type and p-type semiconductors and the semiconductor character of the passive films did not change by grain refinement. Moreover, it was found that the calculated donor and acceptor densities increased with grain refinement. Thus, the presented results indicated that grain refinement weakens the corrosion and passivation behavior of AISI 430 stainless steel in this alkaline solution.

  19. Microchemical characterization of grain boundaries in irradiated steels

    SciTech Connect

    Walmsley, J.; Spellward, P.; Fisher, S.; Jenssen, A.

    1995-12-31

    Field Emission Gun Scanning Transmission Electron Microscopy and Auger Electron Spectroscopy have been used to characterize grain boundaries in unirradiated and neutron-irradiated type 304 stainless steel. Both techniques are used to give compositional information with nanometer-scale spatial resolution at and around grain boundaries. Irradiation induced changes in grain boundary nanochemistry from the solution treated starting condition are described. Initial segregation of Cr at boundaries is seen to develop through an intermediate ``side-lobe`` distribution, seen clearly at an intermediate dose of {approximately}10{sup 21}n/cm{sup 2}, to Cr depletion at higher dose of {approximately} 10{sup 22}n/cm{sup 2}. Thin foil analysis suggests a considerably higher grain boundary phosphorus level in the intermediate dose material than is measured by fracture surface analysis. For the high dose material the two techniques produce consistent phosphorus levels when comparison is made using experience gained from dual examinations of other steels. It is suggested that in the medium dose material fracture occurs along the plane of minimum chromium arising from the ``side-lobe`` Cr distribution so that the surface exposed by fracture is several nanometers away from the true grain boundary.

  20. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Hashimoto, N.; Sokolov, M. A.; Shiba, K.; Jitsukawa, S.

    2006-10-01

    Tensile and Charpy specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 °C in the High Flux Isotope Reactor (HFIR) up to ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n, α) reactions of thermal neutrons with 58Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile-brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 °C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2-4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

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

    SciTech Connect

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

    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.

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

    SciTech Connect

    Stoller, R.E.

    1987-12-01

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

  3. High temperature deformation mechanism of 15CrODS ferritic steels at cold-rolled and recrystallized conditions

    NASA Astrophysics Data System (ADS)

    Sugino, Yoshito; Ukai, Shigeharu; Oono, Naoko; Hayashi, Shigenari; Kaito, Takeji; Ohtsuka, Satoshi; Masuda, Hiroshi; Taniguchi, Satoshi; Sato, Eiichi

    2015-11-01

    The ODS ferritic steels realize potentially higher operating temperature due to structural stability by the dispersed nano-size oxide particles. The deformation process and mechanism of 15CrODS ferritic steels were investigated at 1073 K and 1173 K for the cold-rolled and recrystallized conditions. Tensile and creep tests were conducted at the stress in parallel (LD) and perpendicular (TD) directions to the grain boundaries. Strain rate varied from 10-1 to 10-9 s-1. For the LD specimens, deformation in the cold rolled and recrystallized conditions is reinforced by finely dispersed oxide particles. The dominant deformation process for the recrystallized TD specimen is controlled through the grain boundary sliding and stress accommodation via diffusional creep at temperature of 1173 K and lower strain rate less than 10-4 s-1. The grain boundary sliding couldn't be rate-controlling process at 1073 K for the as-cold rolled TD specimen, where a dynamic recovery of the dislocation produced by cold-rolling is related to the deformation process.

  4. Parametric Optimization Of Gas Metal Arc Welding Process By Using Grey Based Taguchi Method On Aisi 409 Ferritic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Ghosh, Nabendu; Kumar, Pradip; Nandi, Goutam

    2016-10-01

    Welding input process parameters play a very significant role in determining the quality of the welded joint. Only by properly controlling every element of the process can product quality be controlled. For better quality of MIG welding of Ferritic stainless steel AISI 409, precise control of process parameters, parametric optimization of the process parameters, prediction and control of the desired responses (quality indices) etc., continued and elaborate experiments, analysis and modeling are needed. A data of knowledge - base may thus be generated which may be utilized by the practicing engineers and technicians to produce good quality weld more precisely, reliably and predictively. In the present work, X-ray radiographic test has been conducted in order to detect surface and sub-surface defects of weld specimens made of Ferritic stainless steel. The quality of the weld has been evaluated in terms of yield strength, ultimate tensile strength and percentage of elongation of the welded specimens. The observed data have been interpreted, discussed and analyzed by considering ultimate tensile strength ,yield strength and percentage elongation combined with use of Grey-Taguchi methodology.

  5. Long-term oxidation behavior of spinel-coated ferritic stainless steel for solid oxide fuel cell interconnect applications

    SciTech Connect

    Stevenson, Jeffry W.; Yang, Zhenguo; Xia, Guanguang; Nie, Zimin; Templeton, Joshua D.

    2013-06-01

    Long-term tests (>8,000 hours) indicate that AISI 441 ferritic stainless steel coated with a Mn-Co spinel protection layer is a promising candidate material system for IT-SOFC interconnect applications. While uncoated AISI 441 showed a substantial increase in area-specific electrical resistance (ASR), spinel-coated AISI 441 exhibited much lower ASR values (11-13 mOhm-cm2). Formation of an insulating silica sublayer beneath the native chromia-based scale was not observed, and the spinel coatings reduced the oxide scale growth rate and blocked outward diffusion of Cr from the alloy substrate. The structure of the scale formed under the spinel coatings during the long term tests differed from that typically observed on ferritic stainless steels after short term oxidation tests. While short term tests typically indicate a dual layer scale structure consisting of a chromia layer covered by a layer of Mn-Cr spinel, the scale grown during the long term tests consisted of a chromia matrix with discrete regions of Mn-Cr spinel distributed throughout the matrix. The presence of Ti in the chromia scale matrix and/or the presence of regions of Mn-Cr spinel within the scale may have increased the scale electrical conductivity, which would explain the fact that the observed ASR in the tests was lower than would be expected if the scale consisted of pure chromia.

  6. Factors Affecting the Inclusion Potency for Acicular Ferrite Nucleation in High-Strength Steel Welds

    NASA Astrophysics Data System (ADS)

    Kang, Yongjoon; Jeong, Seonghoon; Kang, Joo-Hee; Lee, Changhee

    2016-06-01

    Factors affecting the inclusion potency for acicular ferrite nucleation in high-strength weld metals were investigated and the contribution of each factor was qualitatively evaluated. Two kinds of weld metals with different hardenabilities were prepared, in both, MnTi2O4-rich spinel formed as the predominant inclusion phase. To evaluate the factors determining the inclusion potency, the inclusion characteristics of size, phase distribution in the multiphase inclusion, orientation relationship with ferrite, and Mn distribution near the inclusion were analyzed. Three factors affecting the ferrite nucleation potency of inclusions were evaluated: the Baker-Nutting (B-N) orientation relationship between ferrite and the inclusion; the formation of an Mn-depleted zone (MDZ) near the inclusion; and the strain energy around the inclusion. Among these, the first two factors were found to be the most important. In addition, it was concluded that the increased chemical driving force brought about by the formation of an MDZ contributed more to the formation of acicular ferrite in higher-strength weld metals, because the B-N orientation relationship between ferrite and the inclusion was less likely to form as the transformation temperature decreased.

  7. Studies on oxidation and deuterium permeation behavior of a low temperature α-Al2O3-forming Fesbnd Crsbnd Al ferritic steel

    NASA Astrophysics Data System (ADS)

    Xu, Yu-Ping; Zhao, Si-Xiang; Liu, Feng; Li, Xiao-Chun; Zhao, Ming-Zhong; Wang, Jing; Lu, Tao; Hong, Suk-Ho; Zhou, Hai-Shan; Luo, Guang-Nan

    2016-08-01

    To evaluate the capability of Fesbnd Crsbnd Al ferritic steels as tritium permeation barrier in fusion systems, the oxidation behavior together with the permeation behavior of a Fesbnd Crsbnd Al steel was investigated. Gas driven permeation experiments were performed. The permeability of the oxidized Fesbnd Crsbnd Al steel was obtained and a reduced activation ferritic/martensitic steel CLF-1 was used as a comparison. In order to characterize the oxide layer, SEM, XPS, TEM, HRTEM were used. Al2O3 was detected in the oxide film by XPS, and HRTEM showed that Al2O3 in the α phase was found. The formation of α-Al2O3 layer at a relatively low temperature may result from the formation of Cr2O3 nuclei.

  8. Dislocation structures in the bands of localized cyclic plastic strain in austenitic 316L and austenitic-ferritic duplex stainless steels

    SciTech Connect

    Kruml, T.; Polak, J.; Obrtlik, K.; Degallaix, S.

    1997-12-01

    Dislocation structures in bands corresponding to cyclic strain localization have been studied in two types of stainless steels, single phase austenitic 316L steel and two-phase austenitic-ferritic duplex steel. Dislocation structures are documented in thin foils oriented approximately perpendicular to the active slip plane of individual grains and parallel to the primary Burgers vector. Persistent slip bands, with the structure more or less reminiscent of the well-known ladder structure, were found in austenitic grains of both steels. These bands can be correlated with the distinct surface relief consisting of extrusions, intrusions and shallow surface cracks in austenitic grains were found. The distribution of the wall and labyrinth structure embedded in the matrix structure in ferritic grains, which was proposed to be responsible for the localization of the cyclic strain, however, does not correspond to the distribution of the distinct surface slip lines on the surface.

  9. ENABLING THE PRACTICAL APPLICATION OF OXIDE DISPERSION-STRENGTHENED FERRITIC STEELS

    SciTech Connect

    Wright, Ian G; Pint, Bruce A; Dyadko, Dr. Eugene G.; Bornstein, Norman S.; Tatlock, Gordon J

    2007-01-01

    Effort has continued to evaluate joints made in oxide dispersion-strengthened (ODS) FeCrAl by (i) pulsed plasma-assisted diffusion (PPAD) bonding, and (ii) transient liquid phase (TLP) bonding. Creep tests of PPAD-bonded butt joints in air at 1000 C, using small, shoulder-loaded, dog bone-shaped specimens and an incrementally-loaded test technique, indicated that failure occurred at loads of up to 82% of that required to fail the parent alloy in the same test. For high creep-strength ferritic steels joined by conventional welding methods, strength reduction factors of 50-80% are considered to be acceptable. The failures apparently did not initiate along the joints; the observed mode of failure of the joined specimens was the same as observed for monolithic specimens of this alloy, by crack-initiated transgranular brittle fracture, followed by ductile overload failure. The progress of TLP bonding has been slower, with the major effort focused on understanding the behavior of the transient liquid phase and its interaction with the alloy microstructure during the various stages of bonding. Creep testing using the same procedures also has been used to evaluate changes resulting from torsional deformation of ODS-FeCrAl tubes in an attempt to modify their microstructures and increase their hoop strength. Interpretation of the results so far has not shown a clear trend, largely due to difficulties in measuring the effective angle of twist in the specimen gauge lengths. Other issues that have been addressed are the refinement of an approach for prediction of the oxidation-limited service lifetime of alumina scale-forming ODS alloys, and alternative routes for ODS alloy powder processing. Analysis of alloy specimens oxidized to failure (in some cases involving exposures for many thousands of hours) over a range of temperatures has provided an improved basis for calculating the values of parameters required in the lifing model (minimum Al content for protective behavior

  10. Evaluation by the Double Loop Electrochemical Potentiokinetic Reactivation Test of Aged Ferritic Stainless Steel Intergranular Corrosion Susceptibility

    NASA Astrophysics Data System (ADS)

    Sidhom, H.; Amadou, T.; Braham, C.

    2010-12-01

    An experimental design method was used to determine the effect of factors that significantly affect the response of the double loop-electrochemical potentiokinetic reactivation (DL-EPR) test in controlling the susceptibility to intergranular corrosion (IGC) of UNS S43000 (AISI 430) ferritic stainless steel. The test response is expressed in terms of the reactivation/activation current ratio ( I r / I a pct). Test results analysed by the analysis of variance (ANOVA) method show that the molarity of the H2SO4 electrolyte and the potential scanning rate have a more significant effect on the DL-EPR test response than the temperature and the depassivator agent concentration. On the basis of these results, a study was conducted in order to determine the optimal operating conditions of the test as a nondestructive technique for evaluating IGC resistance of ferritic stainless steel components. Three different heat treatments are considered in this study: solution annealing (nonsensitized), aging during 3 hours at 773 K (500 °C) (slightly sensitized), and aging during 2 hours at 873 K (600 °C) (highly sensitized). The aim is to find the operating conditions that simultaneously ensure the selectivity of the attack (intergranular and chromium depleted zone) and are able to detect the effect of low dechromization. It is found that a potential scanning rate of 2.5 mV/s in an electrolyte composed of H2SO4 3 M solution without depassivator, at a temperature around 293 K (20 °C), is the optimal operating condition for the DL-EPR test. Using this condition, it is possible to assess the degree of sensitization (DOS) to the IGC of products manufactured in ferritic stainless steels rapidly, reliably, and quantitatively. A time-temperature-start of sensitization (TTS) diagram for the UNS S43000 (France Inox, Villepinte, France) stainless steel was obtained with acceptable accuracy by this method when the IGC sensitization criterion was set to I r / I a > 1 pct. This diagram is in

  11. Investigation of iron-chromium-niobium-titanium ferritic stainless steel for solid oxide fuel cell interconnect applications

    NASA Astrophysics Data System (ADS)

    Yang, Zhenguo; Xia, Guan-Guang; Wang, Chong-Min; Nie, Zimin; Templeton, Joshua; Stevenson, Jeffry W.; Singh, Prabhakar

    As part of an effort to develop cost-effective ferritic stainless steel-based interconnects for solid oxide fuel cell (SOFC) stacks, both bare AISI441 and AISI441 coated with (Mn,Co) 3O 4 protection layers were studied in terms of its metallurgical characteristics, oxidation behavior, and electrical performance. The addition of minor alloying elements, in particular Nb, led to formation of Laves phases both inside grains and along grain boundaries. In particular, the Laves phase which precipitated out along grain boundaries during exposure at intermediate SOFC operating temperatures was found to be rich in both Nb and Si. The capture of Si in the Laves phase minimized the Si activity in the alloy matrix and prevented formation of an insulating silica layer at the scale/metal interface, resulting in a reduction in area-specific electrical resistance (ASR). However, the relatively high oxidation rate of the steel, which leads to increasing ASR over time, and the need to prevent volatilization of chromium from the steel necessitates the application of a conductive protection layer on the steel. In particular, the application of a Mn 1.5Co 1.5O 4 spinel protection layer substantially improved the electrical performance of the 441 by reducing the oxidation rate.

  12. Oxidation behavior of ferritic-martensitic and ODS steels in supercritical water

    NASA Astrophysics Data System (ADS)

    Bischoff, Jeremy

    water corroded much faster than those in steam (1.5 to 2 times faster). Additionally, during these corrosion tests a marker experiment was performed with the deposition of micrometric palladium markers on the surface of some samples prior to oxidation. The markers were found at the outer-inner layer interface, consistent with a corrosion mechanism of outward migration of iron to form the outer layer and inward migration of oxygen to form the inner layer. The discrepancy between the SCW and steam environments suggests that the outward migration of iron may be the rate-limiting step. A detailed study of the oxide advancement was performed using the TEM by analyzing the inner and diffusion layer structure. Energy-filtered TEM images were acquired to analyze the micrometric and nanometric distribution of elements in these layers. Such images from the inner layer revealed the presence of localized chromium enrichment regions associated with the presence of pores. Additionally, an iron-chromium nanometric segregation was observed and may be associated with the mixture of Fe3O4 and FeCr2O4. In the diffusion layer, small nanometric chromium-rich oxide particles were seen within metal grains. The (Fe,Cr)3O4 spinel oxide has an inverse spinel structure as Fe3O4 but becomes normal spinel as FeCr 2O4, thus the structure changes depending on the chromium content. Additionally, the spinel structure was analyzed using the ligand theory and showed that chromium does not migrate and that the main diffusing species is the Fe2+ ion. Calculations of the amount of iron leaving the inner layer showed that this amount accounted for the amount of iron necessary to form the outer layer, thus no dissolution of oxide in SCW is observed. Additionally, the differences in oxidation behavior in steam and SCW suggest that the rate-limiting step for the corrosion of ferritic-martensitic steels is the iron outward migration. The iron migration is driven by the gradient in the Fe2+/Fe 3+ ratio and is

  13. Stress corrosion cracking susceptibility of irradiated Type 304 stainless steels

    SciTech Connect

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

    1992-08-01

    Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 [times] 10[sup 2l] n[center dot]cm[sup [minus]2] (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of the CP absorber tubes or the CP control blade sheath. IGSCC susceptibilities of the BWR components could not be correlated to segregation impurities on grain boundaries. However for comparable fluence levels, Cr on grain-boundaries.

  14. Stress corrosion cracking susceptibility of irradiated Type 304 stainless steels

    SciTech Connect

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

    1992-08-01

    Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 {times} 10{sup 2l} n{center_dot}cm{sup {minus}2} (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of the CP absorber tubes or the CP control blade sheath. IGSCC susceptibilities of the BWR components could not be correlated to segregation impurities on grain boundaries. However for comparable fluence levels, Cr on grain-boundaries.

  15. Stability Of Nanoclusters In 14YWT Oxide Dispersion Strengthened Steel Under Heavy Ion-irradiation By Atom Probe Tomography

    SciTech Connect

    He, Jianchao; Wan, F.; Sridharan, Kumar; Allen, Todd R.; Certain, Alicia G.; Shutthanandan, V.; Wu, Yaqiao

    2014-12-01

    14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 °C, 450 °C, and 600 °C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 °C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 × 1023 to 3.6 × 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.

  16. Suppression effect of nano-sized oxide particles on helium irradiation hardening in F82H-ODS steel

    NASA Astrophysics Data System (ADS)

    Chen, S.; Wang, Y.; Tadaki, K.; Hashimoto, N.; Ohnuki, S.

    2014-12-01

    Helium implantation was performed to investigate irradiation hardening in ferritic/martensitic steels. Depth dependence of nano-hardness was obtained using a Berkovich nano-indenter, and then nano-hardness was extracted from Nix-Gao model. The correlation between irradiation hardening and the concentration 500-2000 appm of helium was plotted. Nano-hardness increases as a function of helium concentration. F82H-ODS with a higher nano-hardness provides a lower irradiation hardening than F82H-IEA. Cross-sectional transmission electron microscopy (XTEM) revealed that cavities with a uniform distribution were formed after helium implantation at 2000 appm helium concentration, showing a mean size of 1.1 nm with an average number density of 4.9 × 1023 m-3 in F82H-IEA and 1.3 nm with 7.4 × 1023 m-3 in F82H-ODS. Orowan model was applied to evaluate the hardening from dispersed cavities. The significant difference of hardening between calculation and nano-indentation result of F82H-ODS indicates that oxide particles may shield the hardening effect from cavities because of the complex multi-interaction.

  17. High-resolution electron microscopy observation and dislocation reaction mechanism of fivefold twinning in a Cu-rich precipitate in a cold rolled ferritic steel containing copper

    SciTech Connect

    Wang, Ling; Wang, Wei; Chen, Bolin; Zhou, Xiying; Li, Zhongwen; Zhou, Bangxin; Wang, Lumin

    2014-09-15

    Ferritic steels containing copper have been studied as model systems for clusters/precipitate formation in reactor pressure vessel steels. The samples were aged at 400 °C for 4000 h and subsequently cold rolled to 30% reduction at room temperature. The microstructural characteristics of the samples were analyzed using high-resolution transmission electron microscopy. Direct evidence was found that the fivefold twinning occurs via simultaneous emission of two Shockley partial dislocations from two particular α-Fe/Cu interfaces, and then the pileup tips of the twofold twin. - Highlights: • Fivefold twin is observed in a Cu-rich precipitate in cold rolled ferritic steels. • A dislocation reaction mechanism for the fivefold twin formation is proposed. • Two particular mismatching α-Fe/Cu-rich precipitate interfaces play a critical role.

  18. Deformation and cracking of irradiated austenitic stainless steels

    SciTech Connect

    Carter, R.D.; Atzmon, M.; Was, G.S.

    1995-12-31

    Samples of proton-irradiated 304L stainless steel were deformed by constant extension rate tensile tests at strain rates of 3 {times} 10{sup {minus}7} s{sup {minus}1} and 3 {times} 10{sup {minus}8} s{sup {minus}1} to strains of up to 10% at 288--350 C in argon. Minor cracking was observed in and around spinel inclusions in the material, however no intergranular cracking of the type observed in water environments was found. Thus intergranular cracking cannot occur by a radiation hardening mechanism alone. The microstructures that resulted from irradiation and deformation were characterized using electron microscopy. Surface slip band formation is observed on one or two {l_brace}111{r_brace} slip systems in each grain. The slip bands correspond to dislocation channels in the material as identified by transmission electron microscopy. The channels form by activation of grain-boundary dislocation sources, with the emitted dislocations sweeping through the grain interior to the opposing rain boundaries. During this process, the dislocations remove the radiation-produced defects. Slip band and dislocation channel densities increase with increasing strain in the samples. These results are used to interpret stress corrosion cracking behavior in this material.

  19. Development of A New Class of Fe-3Cr-W(V)Ferritic Steels for Industrial Process Applications

    SciTech Connect

    Sikka, V.J.; Jawad, M.H.

    2005-06-15

    The project, 'Development of a New Class of Fe-Cr-W(V) Ferritic Steels for Industrial Process Applications', was a Cooperative Research and Development Agreement (CRADA) between Oak Ridge National Laboratory (ORNL) and Nooter Corporation. This project dealt with improving the materials performance and fabrication for the hydrotreating reactor vessels, heat recovery systems, and other components for the petroleum and chemical industries. The petroleum and chemical industries use reactor vessels that can approach the ship weights of approximately 300 tons with vessel wall thicknesses of 3 to 8 in. These vessels are typically fabricated from Fe-Cr-Mo steels with chromium ranging from 1.25 to 12% and molybdenum from 1 to 2%. Steels in this composition have great advantages of high thermal conductivity, low thermal expansion, low cost, and properties obtainable by heat treatment. With all of the advantages of Fe-Cr-Mo steels, several issues are faced in design and fabrication of vessels and related components. These issues include the following: (1) low strength properties of current alloys require thicker sections; (2) increased thickness causes heat-treatment issues related to nonuniformity across the thickness and thus not achieving the optimum properties; (3) fracture toughness (ductile-to-brittle transition ) is a critical safety issue for these vessels, and it is affected in thick sections due to nonuniformity of microstructure; (4) PWHT needed after welding and makes fabrication more time-consuming with increased cost; and (5) PWHT needed after welding also limits any modifications of the large vessels in service. The goal of this project was to reduce the weight of large-pressure vessel components (ranging from 100 to 300 tons) by approximately 25% and reduce fabrication cost and improve in-service modification feasibility through development of Fe-3Cr-W(V) steels with combination of nearly a 50% higher strength, a lower DBTT and a higher upper-shelf energy

  20. End Closure Joining of Ferritic-Martensitic and Oxide-Dispersion Strengthened Steel Cladding Tubes by Magnetic Pulse Welding

    NASA Astrophysics Data System (ADS)

    Lee, Jung-Gu; Park, Jin-Ju; Lee, Min-Ku; Rhee, Chang-Kyu; Kim, Tae-Kyu; Spirin, Alexey; Krutikov, Vasiliy; Paranin, Sergey

    2015-07-01

    The magnetic pulse welding (MPW) technique was employed for the end closure joining of fuel pin cladding tubes made of ferritic-martensitic (FM) steel and oxide-dispersion strengthened (ODS) steel. The technique is a solid-state impact joining process based on the electromagnetic force, similar to explosive welding. For a given set of optimal process parameters, e.g., the end-plug geometry, the rigid metallurgical bonding between the tube and end plug was obtained by high-velocity impact collision accompanied with surface jetting. The joint region showed a typical wavy morphology with a narrow grain boundary-like bonding interface. There was no evidence of even local melting, and only the limited grain refinement was observed in the vicinity of the bonding interface without destructing the original reinforcement microstructure of the FM-ODS steel, i.e., a fine grain structure with oxide dispersion. No leaks were detected during helium leakage test, and moreover, the rupture occurred in the cladding tube section without leaving any joint damage during internal pressure burst test. All of the results proved the integrity and durability of the MPWed joints and signified the great potential of this method of end closure joining for advanced fast reactor fuel pin fabrication.

  1. Gas atomized precursor alloy powder for oxide dispersion strengthened ferritic stainless steel

    SciTech Connect

    Rieken, Joel

    2011-12-13

    Gas atomization reaction synthesis (GARS) was employed as a simplified method for producing precursor powders for oxide dispersion strengthened (ODS) ferritic stainless steels (e.g., Fe-Cr-Y-(Ti,Hf)-O), departing from the conventional mechanical alloying (MA) process. During GARS processing a reactive atomization gas (i.e., Ar-O2) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 150 nm) metastable Cr-enriched oxide layer that was used as a vehicle for solid-state transport of O into the consolidated microstructure. In an attempt to better understand the kinetics of this GARS reaction, theoretical cooling curves for the atomized droplets were calculated and used to establish an oxidation model for this process. Subsequent elevated temperature heat treatments, which were derived from Rhines pack measurements using an internal oxidation model, were used to promote thermodynamically driven O exchange reactions between trapped films of the initial Cr-enriched surface oxide and internal Y-enriched intermetallic precipitates. This novel microstructural evolution process resulted in the successful formation of nano-metric Y-enriched dispersoids, as confirmed using high energy X-ray diffraction and transmission electron microscopy (TEM), equivalent to conventional ODS alloys from MA powders. The thermal stability of these Y-enriched dispersoids was evaluated using high temperature (1200°C) annealing treatments ranging from 2.5 to 1,000 hrs of exposure. In a further departure from current ODS practice, replacing Ti with additions of Hf appeared to improve the Y-enriched dispersoid thermal stability by means of crystal structure modification. Additionally, the spatial distribution of the dispersoids was found to depend strongly on the original rapidly solidified microstructure. To exploit this, ODS microstructures were engineered from

  2. Gas atomized precursor alloy powder for oxide dispersion strengthened ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Rieken, Joel Rodney

    Gas atomization reaction synthesis (GARS) was employed as a simplified method for producing precursor powders for oxide dispersion strengthened (ODS) ferritic stainless steels (e.g., Fe-Cr-Y-(Ti,Hf)-O), departing from the conventional mechanical alloying (MA) process. During GARS processing a reactive atomization gas (i.e., Ar-O2) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 150 nm) metastable Cr-enriched oxide layer that was used as a vehicle for solid-state transport of O into the consolidated microstructure. In an attempt to better understand the kinetics of this GARS reaction, theoretical cooling curves for the atomized droplets were calculated and used to establish an oxidation model for this process. Subsequent elevated temperature heat treatments, which were derived from Rhines pack measurements using an internal oxidation model, were used to promote thermodynamically driven O exchange reactions between trapped films of the initial Cr-enriched surface oxide and internal Y-enriched intermetallic precipitates. This novel microstructural evolution process resulted in the successful formation of nano-metric Y-enriched dispersoids, as confirmed using high energy X-ray diffraction and transmission electron microscopy (TEM), equivalent to conventional ODS alloys from MA powders. The thermal stability of these Y-enriched dispersoids was evaluated using high temperature (1200°C) annealing treatments ranging from 2.5 to 1,000 hrs of exposure. In a further departure from current ODS practice, replacing Ti with additions of Hf appeared to improve the Y-enriched dispersoid thermal stability by means of crystal structure modification. Additionally, the spatial distribution of the dispersoids was found to depend strongly on the original rapidly solidified microstructure. To exploit this, ODS microstructures were engineered from different

  3. Microstructure and nanoindentation of the CLAM steel with nanocrystalline grains under Xe irradiation

    NASA Astrophysics Data System (ADS)

    Chang, Yongqin; Zhang, Jing; Li, Xiaolin; Guo, Qiang; Wan, Farong; Long, Yi

    2014-12-01

    This work presents an early look at irradiation effects on China low activation martensitic (CLAM) steel with nanocrystalline grains (NC-CLAM steels) under 500 keV Xe-ion bombardment at room temperature to doses up to 5.3 displacements per atom (dpa). The microstructure in the topmost region of the steel is composed of nanocrystalline grains with an average diameter of 13 nm. As the samples were implanted at low dose, the nanocrystalline grains had martensite lath structure, and many dislocations and high density bubbles were introduced into the NC-CLAM steels. As the irradiation dose up to 5.3 dpa, a tangled dislocation network exists in the lath region, and the size of the bubbles increases. X-ray diffraction results show that the crystal quality decreases after irradiation, although the nanocrystals obviously coarsen. Grain growth under irradiation may be ascribed to the direct impact of the thermal spike on grain boundaries in the NC-CLAM steels. In irradiated samples, a compressive stress exists in the surface layer because of grain growth and irradiation-introduced defects, while the irradiation introduced grain-size coarsening and defects gradients from the surface to matrix result in a tensile stress in the irradiated NC-CLAM steels. Nanoindentation was used to estimate changes in mechanical properties during irradiation, and the results show that the hardness of the NC-CLAM steels increases with increasing irradiation dose, which was ascribed to the competition between the grain boundaries and the irradiation-introduced defects.

  4. Residual stresses in a ferritic steel welded pipe: an experimental comparison between reactor and pulsed neuron sources

    NASA Astrophysics Data System (ADS)

    Albertini, Gianni; Brugnami, D.; Bruno, Giovanni; Ceretti, M.; Cernuschi, F. M.; Edwards, Lyndon

    1997-02-01

    Residual stresses induced by welding processes can affect the integrity of structural components like tubes and pipes of thermoelectric power plants. In order to reduce or cancel these stresses, welded components are often heat treated. The residual stress field in an arc-welded 2.25Cr1Mo ferritic steel pipe was measured using neutron diffraction both before and after stress relieving heat treatment. In the first stage stresses were measured using two different kinds of neutron sources: a reactor and a pulsed one. A comparison between results obtained using G5.2 diffractometer at LLB and ENGIN equipment at ISIS will be outlined and the effectiveness of heat treatment will be shown.

  5. AFM and TEM study of cyclic slip localization in fatigued ferritic X10CrAl24 stainless steel

    SciTech Connect

    Man, J. . E-mail: man@ipm.cz; Petrenec, M.; Obrtlik, K.; Polak, J.

    2004-11-08

    Atomic force microscopy and high resolution scanning electron microscopy were applied to the study of surface relief evolution at emerging persistent slip bands (PSBs) in individual grains of ferritic X10CrAl24 stainless steel cycled with constant plastic strain amplitude. Only the combination of both methods can reveal the true shape and fine details of extrusions and intrusions. Quantitative data on the changes of the surface topography of persistent slip markings and on the kinetics of extrusion growth during the fatigue life were obtained. Transmission electron microscopy of surface foils revealed PSBs with the typical, well-known ladder structure. Experimental data on cyclic slip localization in PSBs are compared with those in fcc metals and discussed in terms of vacancy models of surface relief evolution and fatigue crack initiation.

  6. Study for corrosion characteristics of ferritic stainless steel weld metal with respect to added contents of Ti and Nb

    NASA Astrophysics Data System (ADS)

    Kim, JongMin; Lee, HaeWoo

    2014-03-01

    This paper identified the effects of Ti and Nb on pitting and intergranular corrosion resistance in a ferritic stainless steel weld metal of the automobile exhaust system. We fabricated 4 flux cored wires designed with 0-0.2 wt% Ti and 0-1.0 wt% Nb and performed Flux Cored Arc Welding. Through the potentiodynamic polarization test in 0.5M NaCl, we evaluated pitting resistance. And in order to evaluate the intergranular corrosion resistance, we observed microstructure after we performed DL-EPR test in 0.5M H2SO4+0.01M KSCN. As a result of the test, the specimen added with 0.2%Ti+1.0%Nb showed the highest pitting resistance. From observing the degree of sensitization and microstructure, the intergranular corrosion resistance was higher as the contents of Ti and Nb increased. And through EBSD we observed Cr carbide which affects the corrosion resistance.

  7. New nano-particle-strengthened ferritic/martensitic steels by conventional thermo-mechanical treatment

    NASA Astrophysics Data System (ADS)

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

    2007-08-01

    For increased fusion power plant efficiency, steels for operation at 650 °C and higher are sought. Based on the science of precipitate strengthening, a thermo-mechanical treatment (TMT) was developed that increased the strength from room temperature to 700 °C of commercial nitrogen-containing steels and new steels designed for the TMT. At 700 °C increases in yield stress of 80 and 200% were observed for a commercial steel and a new steel, respectively, compared to commercial normalized-and-tempered steels. Creep-rupture strength was similarly improved. Depending on the TMT, precipitates were up to eight-times smaller at a number density four orders of magnitude greater than those in a conventionally heat treated steel of similar composition.

  8. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

    PubMed Central

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-01

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density. PMID:28079191

  9. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

    NASA Astrophysics Data System (ADS)

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-01

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  10. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation.

    PubMed

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-12

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  11. Irradiation effects on magnetic properties in neutron and proton irradiated reactor pressure vessel steel

    SciTech Connect

    Park, D.G.; Hong, J.H.; Kim, I.S.; Kim, H.C.

    1999-09-01

    The effects of neutron and proton dose on the magnetic properties of a reactor pressure vessel (RPV) steel were investigated. The coercivity and maximum induction increased in two stages with respect to neutron dose, being nearly constant up to a dose of 1.5 x 10{sup {minus}7} dpa, followed by a rapid increase up to a dose of 1.5 x 10{sup {minus}5} dpa. The coercivity and maximum induction in the proton irradiated specimens also showed a two stage variation with respect to proton dose, namely a rapid increase up to a dose of 0.2 x 10{sup {minus}2} dpa, then a decrease up to 1.2 x 10{sup {minus}2} dpa. The Barkhausen noise (BN) amplitude in neutron irradiated specimens also varied in two stages in a reverse manner, the transition at the same dose of 1.5 x 10{sup {minus}7} dpa. The BN amplitude in proton irradiated specimens decreased by 60% up to 0.2 x 10{sup {minus}2} dpa followed by an increase up to 1.2 x 10{sup {minus}2} dpa. The results were in good accord with the one dimensional domain wall model considering the density of defects and wall energy.

  12. Influence of alloy content and a cerium surface treatment on the oxidation behavior of Fe-Cr ferritic stainless steels

    SciTech Connect

    Alman, D.E.; Jablonski, P.D.

    2006-01-01

    The cost of solid oxide fuel cells (SOFC) can be significantly reduced by using interconnects made from ferritic stainless steels. In fact, several alloys have been developed specifically for this application (Crofer 22APU and Hitachi ZMG323). However, these steels lack environmental stability in SOFC environments, and as a result, degrade the performance of the SOFC. A steel interconnect can contribute to performance degradation through: (i) Cr poisoning of electrochemically active sites within the cathode; (ii) formation of non-conductive oxides, such as SiO2 or Al2O3 from residual or minor alloying elements, at the base metal-oxide scale interface; and/or (iii) excessive oxide scale growth, which may also retard electrical conductivity. Consequently, there has been considerable attention on developing coatings to protect steel interconnects in SOFC environments and controlling trace elements during alloy production. Recently, we have reported on the development of a Cerium surface treatment that improves the oxidation behavior of a variety alloys, including Crofer 22APU [1-5]. Initial results indicated that the treatment may improve the performance of Crofer 22APU for SOFC application by: (i) retarding scale growth resulting in a thinner oxide scale; and (ii) suppressing the formation of a deleterious continuous SiO2 layer that can form at the metal-oxide scale interface in materials with high residual Si content [5]. Crofer 22 APU contains Fe-22Cr-0.5Mn-0.1Ti (weight percent). Depending on current market prices and the purity of raw materials utilized for ingot production, Cr can contribute upwards of 90 percent of the raw materials cost. The present research was undertaken to determine the influence of Cr content and minor element additions, especially Ti, on the effectiveness of the Ce surface treatment. Particular emphasis is placed on the behavior of low Cr alloys.

  13. Impact behavior of reduced-activation steels irradiated to 24 dpa*1

    NASA Astrophysics Data System (ADS)

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

    1996-10-01

    Charpy impact tests were previously conducted on eight chromium-tungsten steels after irradiation at 365°C to 6-8 and 15-17 dpa in the Fast Flux Test Facility. These same steels, which range in concentration from 2.25 to 12 wt% (all steels contained 0.1%C), have now been irradiated to 20-24 dpa under the same conditions. Post-irradiation Charpy impact tests after 20-24 dpa showed that the loss of impact toughness, as measured by an increase in the ductile—brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, remained relatively unchanged from the values after 15-17 dpa. As before, the most irradiation-resistant steels were two 9% Cr steels: the DBTT of a 9Cr2W0.25V steel increased 59°C, and for the same composition with an addition of 0.07% Ta, the DBTT increased only 21°C. The other steels developed shifts in DBTT of 100 to 300°C. A 2.25% Cr steel with 2% W and 0.25% V was less severely affected by irradiation than 2.25% Cr steels with 0.25% V and no tungsten, 2% W and no vanadium, and with 1% W and 0.25% V. Steels with 5 and 12% Cr, 2% W, and 0.25% V had properties between those of the 2.25Cr and 9Cr steels.

  14. Microstructural investigation, using small-angle neutron scattering (SANS), of Optifer steel after low dose neutron irradiation and subsequent high temperature tempering

    NASA Astrophysics Data System (ADS)

    Coppola, R.; Lindau, R.; Magnani, M.; May, R. P.; Möslang, A.; Valli, M.

    2007-08-01

    The microstructural effect of low dose neutron irradiation and subsequent high temperature tempering in the reduced activation ferritic/martensitic steel Optifer (9.3 Cr, 0.1 C, 0.50 Mn, 0.26 V, 0.96 W, 0.66 Ta, Fe bal wt%) has been studied using small-angle neutron scattering (SANS). The investigated Optifer samples had been neutron irradiated, at 250 °C, to dose levels of 0.8 dpa and 2.4 dpa. Some of them underwent 2 h tempering at 770 °C after the irradiation. The SANS measurements were carried out at the D22 instrument of the High Flux Reactor at the Institut Max von Laue - Paul Langevin, Grenoble, France. The differences observed in nuclear and magnetic SANS cross-sections after subtraction of the reference sample from the irradiated one suggest that the irradiation and the subsequent post-irradiation tempering produce the growth of non-magnetic defects, tentatively identified as microvoids.

  15. Elevated temperature tensile properties of P9 steel towards ferritic steel wrapper development for sodium cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Choudhary, B. K.; Mathew, M. D.; Isaac Samuel, E.; Christopher, J.; Jayakumar, T.

    2013-11-01

    Tensile deformation and fracture behaviour of the three developmental heats of P9 steel for wrapper applications containing varying silicon in the range 0.24-0.60% have been examined in the temperature range 300-873 K. Yield and ultimate tensile strengths in all the three heats exhibited gradual decrease with increase in temperature from room to intermediate temperatures followed by rapid decrease at high temperatures. A gradual decrease in ductility to a minimum at intermediate temperatures followed by an increase at high temperatures has been observed. The fracture mode remained transgranular ductile. The steel displayed signatures of dynamic strain ageing at intermediate temperatures and dominance of recovery at high temperatures. No significant difference in the strength and ductility values was observed for varying silicon in the range 0.24-0.60% in P9 steel. P9 steel for wrapper application displayed strength and ductility values comparable to those reported in the literature.

  16. Static Recrystallization Kinetics and Crystallographic Texture of Nb-Stabilized Ferritic Stainless Steel Based on Orientation Imaging Microscopy

    NASA Astrophysics Data System (ADS)

    Malta, Paula Oliveira; Alves, Davi Silva; Ferreira, Aline Oliveira Vasconcelos; Moutinho, Iane Dutra; Dias, Carolina Arriel Pedroso; Santos, Dagoberto Brandão

    2017-01-01

    In the present study, Nb-stabilized ferritic stainless steel was prepared with annealing (430-A) and without annealing (430-NA) annealing, and the microstructure of the resulting samples was examined. The steel was then subjected to cold rolling and isothermal annealing in order to analyze its recrystallization kinetics and texture evolution. Microstructural characterization was performed by scanning and transmission electron microscopies. Recrystallization kinetics were evaluated by measuring the microhardness of the samples, and analyzing their kernel average misorientation and grain orientation spread via electron backscatter diffraction. The Avrami exponent data revealed that one-dimensional grain growth occurred owing to the migration of high-angle grain boundaries. The mean activation energies for recrystallization for 430-NA and 430-A was found to be 365 and 419 kJ mol-1, respectively. The recrystallization texture was influenced by oriented nucleation and selected growth mechanisms, as well as by the Nb carbonitride distribution and grain boundary energy. The recrystallized and growing grains with the {554}<225> orientation showed a dimensional advantage over the other recrystallized components. The coincident site lattice boundaries were attributed to the progression of recrystallization since the CSL numeric fraction increased as the temperature increased. The {554}<225> component was associated with the ∑19a boundary, which exerted a significant control on the selective growth during the recrystallization.

  17. Static Recrystallization Kinetics and Crystallographic Texture of Nb-Stabilized Ferritic Stainless Steel Based on Orientation Imaging Microscopy

    NASA Astrophysics Data System (ADS)

    Malta, Paula Oliveira; Alves, Davi Silva; Ferreira, Aline Oliveira Vasconcelos; Moutinho, Iane Dutra; Dias, Carolina Arriel Pedroso; Santos, Dagoberto Brandão

    2017-03-01

    In the present study, Nb-stabilized ferritic stainless steel was prepared with annealing (430-A) and without annealing (430-NA) annealing, and the microstructure of the resulting samples was examined. The steel was then subjected to cold rolling and isothermal annealing in order to analyze its recrystallization kinetics and texture evolution. Microstructural characterization was performed by scanning and transmission electron microscopies. Recrystallization kinetics were evaluated by measuring the microhardness of the samples, and analyzing their kernel average misorientation and grain orientation spread via electron backscatter diffraction. The Avrami exponent data revealed that one-dimensional grain growth occurred owing to the migration of high-angle grain boundaries. The mean activation energies for recrystallization for 430-NA and 430-A was found to be 365 and 419 kJ mol-1, respectively. The recrystallization texture was influenced by oriented nucleation and selected growth mechanisms, as well as by the Nb carbonitride distribution and grain boundary energy. The recrystallized and growing grains with the {554}<225> orientation showed a dimensional advantage over the other recrystallized components. The coincident site lattice boundaries were attributed to the progression of recrystallization since the CSL numeric fraction increased as the temperature increased. The {554}<225> component was associated with the ∑19a boundary, which exerted a significant control on the selective growth during the recrystallization.

  18. Impact of the use of the ferritic/martensitic ODS steels cladding on the fuel reprocessing PUREX process

    NASA Astrophysics Data System (ADS)

    Gwinner, B.; Auroy, M.; Mas, D.; Saint-Jevin, A.; Pasquier-Tilliette, S.

    2012-09-01

    Some ferritic/martensitic oxide dispersed strengthened (F/M ODS) steels are presently developed at CEA for the fuel cladding of the next generation of sodium fast nuclear reactors. The objective of this work is to study if this change of cladding could have any consequences on the spent fuel reprocessing PUREX process. During the fuel dissolution stage the cladding can actually be corroded by nitric acid. But some process specifications impose not to exceed a limit concentration of the corrosion products such as iron and chromium in the dissolution medium. For that purpose the corrosion behavior of these F/M ODS steels is studied in hot and concentrated nitric acid. The influence of some metallurgical parameters such as the chromium content, the elaboration process and the presence of the yttrium oxides is first discussed. The influence of environmental parameters such as the nitric acid concentration, the temperature and the presence of oxidizing species coming from the fuel is then analyzed. The corrosion rate is characterized by mass loss measurements and electrochemical tests. Analyses of the corroded surface are carried out by X-ray photoelectron spectroscopy.

  19. New Ferritic Steels with Combined Optimal Creep Resistance and Ductility by Coupling Thermodynamic Calculations with Focused Experiments

    SciTech Connect

    Teng, Zhenke; Zhang, F; Miller, Michael K; Liu, Chain T; Huang, Shenyan; Chou, Y; Tien, R; Chang, Y A; Liaw, Peter K

    2012-01-01

    Two critical issues restricting the applications of NiAl precipitate-strengthened ferritic steels are their poor room temperature ductility and insufficient creep resistance at temperatures higher than 600 C. In this study, a thermodynamic modeling approach is integrated with experiments focused on investigating the ductility and creep resistance of steel alloys based on the Fe-Ni-Al-Cr-Mo multi-component system. The mechanical property studies showed that the creep resistance increases with increasing the volume fraction of B2-ordered precipitates, while the opposite trend was observed for the ductility. Low solubility of Al in the {alpha}-Fe matrix was found to favor a ductility increase. Thermodynamic calculations were used to predict the volume fraction of B2-ordered precipitate and the elemental partitioning to guide the selection of alloy compositions that might exhibit the balanced creep resistance and ductility. Key experiments were then conducted to validate the prediction. This integrated approach was found to be very effective in the alloy development.

  20. Extending the boundaries of mechanical properties of Ti-Nb low-carbon steel via combination of ultrafast cooling and deformation during austenite-to-ferrite transformation

    NASA Astrophysics Data System (ADS)

    Deng, Xiangtao; Fu, Tianliang; Wang, Zhaodong; Liu, Guohuai; Wang, Guodong; Misra, R. D. K.

    2017-01-01

    We underscore here a novel approach to extend the boundaries of mechanical properties of Ti-Nb low-carbon steel via combination of ultrafast cooling and deformation during austenite-to-ferrite transformation. The proposed approach yields a refined microstructure and high density nano-sized precipitates, with consequent increase in strength. Steels subjected to ultra-fast cooling during austenite-to-ferrite transformation led to 145 MPa increase in yield strength, while the small deformation after ultra-fast cooling process led to increase in strength of 275 MPa. The ultra-fast cooling refined the ferrite and pearlite constituents and enabled uniform dispersion, while the deformation after ultra-fast cooling promoted precipitation and broke the lamellar pearlite to spherical cementite and long thin strips of FexC. The contribution of nano-sized precipitates to yield strength was estimated to be 247.9 MPa and 358.3 MPa for ultrafast cooling and deformation plus ultrafast cooling processes. The nano precipitates carbides were identified to be (Ti, Nb)C and had a NaCl-type crystal structure, and obeyed the Baker-Nutting orientation relationship with the ferrite matrix.

  1. Effect of particle size, fraction and carbide banding on deformation and damage behavior of ferrite-cementite steel under tensile/shear loads

    NASA Astrophysics Data System (ADS)

    Zhuang, Xincun; Ma, Siming; Zhao, Zhen

    2017-01-01

    Deformation and damage behavior of ferrite-cementite steel was investigated using microstructure-based representative volume element (RVE) methodology. A series of automatically generated 2D RVEs with ferrite matrix and globular cementite particles were generated as representative of various microstructures. The geometrically necessary dislocation (GND) accumulation at the ferrite-cementite interphase was also studied by introducing an intermediate layer around the cementite particles. Damage mechanisms such as ductile fracture in ferrite matrix, brittle fracture in cementite, and decohesion at ferrite-cementite interphase were considered to study the fracture modes. The relationships between interface strength and particle size were estimated according to the modified Argon criterion and showed satisfactory agreement with related works. The influences of microstructural features, such as particle size, particle fraction and carbide banding, on deformation and damage evolution were investigated under tensile and shear loads. Simulation results indicated that small particle size and particle fraction could postpone the initial decohesion under both tensile and shear loads, while carbide banding can lead to early fracture due to local stress concentration, which has potential to cause the loss of ductility and premature failure. These adverse effects become more severe when more cementite particles remain in the band or the gather density of the cementite particles in the band becomes higher.

  2. Effect of Initial Heat Treatment on DBTT of F82H Steel Irradiated by Neutrons

    SciTech Connect

    Wakai, E.; Ando, M.; Matsukawa, S.; Taguchi, T.; Yamamoto, T.; Tomita, H.; Takada, F.

    2005-05-15

    The dependence of ductile-brittle transition temperature (DBTT) on tempering time and temperature was examined for a martensitic steel F82H irradiated at 150 and 250 deg. C to a neutron dose of 1.9 dpa in the JMTR. The heat treatment was performed at 750 and 780 deg. C for 0.5 h after the normalizing at 1040 deg. C for 0.5 h. The tempering time at 750 deg. C was varied from 0.5 to 10 h. 1/3CVN specimens were used in this study, and the absorbed energies in the impact tests were measured as a function of temperature. DBTT of F82H steels irradiated at 250 deg. C to 1.9 dpa was ranged from -23 to 25 deg. C, and DBTT of F82H steels irradiated at 150 deg. C to 1.9 dpa was ranged from 0 to 15 deg. C. DBTT of F82H steels irradiated at 250 deg. C depended strongly on temperature and time of tempering, and it tended to decrease with increasing yield stress. The effect of tempering conditions on DBTT was smaller in the specimens irradiated at 150 deg. C. DBTT due to irradiation in the F82H steels irradiated at 250 deg. C tended to decrease with increasing time and temperature of tempering.

  3. Mn1.5Co1.5O4 Spinel Protection Layers on Ferritic Stainless Steels for SOFC Interconnect Applications

    SciTech Connect

    Yang, Z Gary; Xia, Gordon; Stevenson, Jeffry W.

    2005-01-26

    In intermediate solid oxide fuel cells, the use of cost effective chromia forming alloy interconnects such as ferritic stainless steels can lead to severe degradation in cell performance due to chromium migration into the cells at the cathode side. To protect cells from chromium poisoning and improve their performance, a Mn1.5Co1.5O4 spinel barrier layer has been developed and tested on the ferritic stainless steel Crofer22 APU. Thermal and electrical tests confirmed the effectiveness of the spinel protection layer as a means of stopping chromium migration and decreasing oxidation, while promoting electrical contact and minimizing cathode/interconnect interfacial resistance. The thermally grown spinel protection layer was well-bonded to the Crofer22 APU substrate and demonstrated stable performance under thermal cycling.

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

    SciTech Connect

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

    1993-10-01

    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 ({Delta}TT4{sub 41J} or {Delta}TT{sub 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 {Delta}TT) 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.

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

    NASA Astrophysics Data System (ADS)

    Voyevodin, V. N.; Karpov, S. A.; Kopanets, I. E.; Ruzhytskyi, V. V.; Tolstolutskaya, G. D.; Garner, F. A.

    2016-01-01

    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(3He,p)4He, 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.

  6. Fracture toughness of the IEA heat of F82H ferritic/martensitic stainless steel as a function of loading mode

    SciTech Connect

    Li, Huaxin; Gelles, D.S.; Hirth, J.P.

    1997-04-01

    Mode I and mixed-mode I/III fracture toughness tests were performed for the IEA heat of the reduced activation ferritic/martensitic stainless steel F82H at ambient temperature in order to provide comparison with previous measurements on a small heat given a different heat treatment. The results showed that heat to heat variations and heat treatment had negligible consequences on Mode I fracture toughness, but behavior during mixed-mode testing showed unexpected instabilities.

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

    SciTech Connect

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

    2007-01-01

    In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. Exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). We look at the results of a study of simulated light-water reactor coolants, material chemistry, and irradiation damage and their effects on the susceptibility to stress-corrosion cracking of various commercially available and laboratory-melted stainless steels.

  8. Fatigue crack propagation in dual-phase steels: Effects of ferritic-martensitic microstructures on crack path morphology

    NASA Astrophysics Data System (ADS)

    Dutta, V. B.; Suresh, S.; Ritchie, R. O.

    1984-06-01

    microstructures with maximum resistance to fatigue crack extension while maintaining high strength levels. A wide range of crack growth rates has been examined, from ~10-8 to 10-3 mm per cycle, in a series of duplex microstructures of comparable yield strength and prior austenite grain size where intercritical heat treatments were used to vary the proportion, morphology, and distribution of the ferrite and martensite phases. Results of fatigue crack propagation tests, conducted on “long cracks” in room temperature moist air environments, revealed a very large influence of microstructure over the entire spectrum of growth rates at low load ratios. Similar trends were observed at high load ratio, although the extent of the microstructural effects on crack growth behavior was significantly less marked. Specifically, microstructures containing fine globular or coarse martensite in a coarse-grained ferritic matrix demonstrated exceptionally high resistance to crack growth without loss in strength properties. To our knowledge, these microstructures yielded the highest ambient temperature fatigue threshold stress intensity range ΔK0 values reported to date, and certainly the highest combination of strength and ΔK0 for steels ( i.e., ΔK0 values above 19 MPa√m with yield strengths in excess of 600 MPa). Such unusually high crack growth resistance is attributed primarily to a tortuous morphology of crack path which results in a reduction in the crack driving force from crack deflection and roughness-induced crack closure mechanisms. Quantitative metallography and experimental crack closure measurements, applied to currently available analytical models for the deflection and closure processes, are presented to substantiate such interpretations.

  9. The Effect of H and He on Irradiation Performance of Fe and Ferritic Alloys

    SciTech Connect

    James F. Stubbins

    2010-01-22

    This research program was designed to look at basic radiation damage and effects and mechanical properties in Fe and ferritic alloys. The program scope included a number of materials ranging from pure single crystal Fe to more complex Fe-Cr-C alloys. The range of materials was designed to examine materials response and performance on ideal/model systems and gradually move to more complex systems. The experimental program was coordinated with a modeling effort. The use of pure and model alloys also facilitated the ability to develop and employ atomistic-scale modeling techniques to understand the inherent physics underlying materials performance

  10. Embrittlement of Cr-Mo steels after low fluence irradiation in HFIR

    SciTech Connect

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

    1995-04-01

    The goal of this work is the determination of the possible effect of the simultaneous formation of helium and displacement damage during irradiation on the Charpy impact behavior. Subsize Charpy impact specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and 12Cr-1MoVW with 2%Ni (12Cr-1MOVW-2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400{degree}C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toghness. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr-1MoVW-2Ni steel irradiated at 400{degree}C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behaviour of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

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

    SciTech Connect

    Ermi, A.M.; Gelles, D.S.

    1995-04-01

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

  12. Nondestructive Evaluation of Irradiation Embrittlement of SQV2A Steel by Using Magnetic Method

    NASA Astrophysics Data System (ADS)

    Shiwa, Mitsuharu; Weiying, Cheng; Nakahigashi, Shigeo; Komura, Ichiro; Fujiwara, Koji; Takahashi, Norio

    2006-03-01

    Irradiation embrittlement of SQV2A steel was evaluated by magnetic methods. Thermal aging (TA) and electron irradiation (EI) specimens were prepared to evaluate the thermal aging and the irradiation damage effects separately. B-H loops changed with TA and EI. Higher harmonics of AC magnetization signals were sensitive to micro-structure changing of specimens. The intensity of the 3rd harmonics increased linearly with over 100 years of equivalent operation time by Larson-Miller parameter of nuclear power plants.

  13. Nondestructive Evaluation of Irradiation Embrittlement of SQV2A Steel by Using Magnetic Method

    SciTech Connect

    Shiwa, Mitsuharu; Cheng Weiying; Nakahigashi, Shigeo; Komura, Ichiro; Fujiwara, Koji; Takahashi, Norio

    2006-03-06

    Irradiation embrittlement of SQV2A steel was evaluated by magnetic methods. Thermal aging (TA) and electron irradiation (EI) specimens were prepared to evaluate the thermal aging and the irradiation damage effects separately. B-H loops changed with TA and EI. Higher harmonics of AC magnetization signals were sensitive to micro-structure changing of specimens. The intensity of the 3rd harmonics increased linearly with over 100 years of equivalent operation time by Larson-Miller parameter of nuclear power plants.

  14. Applicability of the fracture toughness master curve to irradiated highly embrittled steel and intergranular fracture

    SciTech Connect

    Nanstad, Randy K; Sokolov, Mikhail A; McCabe, Donald E

    2008-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory has evaluated a submerged-arc (SA) weld irradiated to a high level of embrittlement and a temper embrittled base metal that exhibits significant intergranular fracture (IGF) relative to representation by the Master Curve. The temper embrittled steel revealed that the intergranular mechanism significantly extended the transition temperature range up to 150 C above To. For the irradiated highly embrittled SA weld study, a total of 21 1T compact specimens were tested at five different temperatures and showed the Master Curve to be nonconservative relative to the results, although that observation is uncertain due to evidence of intergranular fracture.

  15. JRQ and JPA irradiated and annealed reactor pressure vessel steels studied by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, Vladimír; Gokhman, Oleksandr; Pecko, Stanislav; Sojak, Stanislav; Bergner, Frank

    2016-03-01

    The paper is focused on a comprehensive study of JRQ and JPA reactor pressure vessel steels from the positron annihilation lifetime spectroscopy (PALS) point of view. Based on our more than 20 years' experience with characterization of irradiated reactor steels, we confirmed that defects after irradiation start to grow and/or merge into bigger clusters. Experimental results shown that JPA steel is more sensitive to the creation of irradiation-induced defects than JRQ steel. It is most probably due to high copper content (0.29 wt.% in JPA) and copper precipitation has a major impact on neutron-induced defect creation at the beginning of the irradiation. Based on current PALS results, no large vacancy clusters were formed during irradiation, which could cause dangerous embrittlement concerning operation safety of nuclear power plant. The combined PALS, small angle neutron scattering and atomic probe tomography studies support the model for JRQ and JPA steels describing the structure of irradiation-induced clusters as agglomerations of vacancy clusters (consisting of 2-6 vacancies each) and are separated from each other by a distribution of atoms.

  16. Effect of aging temperature on the microstructures and mechanical properties of ZG12Cr9Mo1Co1NiVNbNB ferritic heat-resistant steel

    NASA Astrophysics Data System (ADS)

    Yang, Xue; Sun, Lan; Xiong, Ji; Zhou, Ping; Fan, Hong-yuan; Liu, Jian-yong

    2016-02-01

    The effect of aging on the mechanical properties and microstructures of a new ZG12Cr9Mo1Co1NiVNbNB ferritic heat resistant steel was investigated in this work to satisfy the high steam parameters of the ultra-supercritical power plant. The results show that the main precipitates during aging are Fe(Cr, Mo)23C6, V(Nb)C, and (Fe2Mo) Laves in the steel. The amounts of the precipitated phases increase during aging, and correspondingly, the morphologies of phases are similar to be round. Fe(Cr, Mo)23C6 appears along boundaries and grows with increasing temperature. In addition, it is revealed that the martensitic laths are coarsened and eventually happen to be polygonization. The hardness and strength decrease gradually, whereas the plasticity of the steel increases. What's more, the hardness of this steel after creep is similar to that of other 9%-12%Cr ferritic steels. Thus, ZG12Cr9Mo1Co1NiVNbNB can be used in the project.

  17. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  18. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    SciTech Connect

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  19. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    SciTech Connect

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  20. Neutron-irradiated model alloys and pressure-vessel steels studied using positron spectroscopy

    NASA Astrophysics Data System (ADS)

    Cumblidge, Stephen Eric

    We have used positron-annihilation-lifetime spectroscopies to examine microstructural evolution of pressure vessel steels and model alloys that have systematically varied amounts of copper, nickel, and phosphorus during neutron irradiation and post-irradiation annealing. The objective of this work was to characterize the neutron-irradiation induced microstructural features that cause the embrittlement of nuclear reactor pressure-vessel steel. We used positron annihilation lifetime spectroscopy and Doppler-broadening spectroscopy to examine the model alloys and pressure-vessel steels before and after irradiation and after post-irradiation annealing. We followed the changes in the mechanical properties of the materials using Rockwell 15N hardness measurements. The results show that in both the model alloys and pressure-vessel steels neutron irradiation causes the formation of vacancy-type defect clusters and a fine distribution of copper- and nickel-enriched metallic precipitates. The vacancy clusters are small in size and were present in all samples, and disappear upon annealing at 450°C. The metallic precipitates are present only in the model alloy samples with either high Cu or a combination of medium Cu and high Ni, and they remain in the microstructure after annealing up to 550°C, starting to anneal possibly at 600°C. The neutron-irradiated pressure vessel steels behave similarly to the high Cu samples, indicating that neutron irradiation induced precipitation occurs in these alloys as well. This work provides independent evidence for the irradiation-induced metallic precipitates seen by other techniques, gives evidence for the exact nature of the matrix damage, and is significant to understanding the in-service degradation of pressure vessel materials.

  1. The influence of ferrite volume fraction on Rayleigh wave propagation in A572 grade 50 steel

    NASA Astrophysics Data System (ADS)

    Abbasi, Zeynab; Tehrani, Niloofar; Ozevin, Didem; Indacochea, J. E.

    2017-02-01

    The acoustoelastic effect is the interaction between ultrasonic wave velocity and stress. To estimate the stress a perturbation signal is introduced and the shift in time of flight is measured at the receiving location. In addition to the stress, the wave velocity can be affected by the volume fraction of the phases in the material's microstructure. This study investigates the changes in Rayleigh wave velocity as a function of stress and microstructure obtained in A572 grade 50 steel following heat treatments. The steel was heat treated to homogenize the microstructure of as-received steel that showed banding; the samples are heat treated at 970 °C for 0.5, 1, and 4 hours, furnace cooled and metallographically characterized. The acoustoelastic coefficient for 1 MHz perturbation frequency is calculated by uniaxial loading of each heat treated plate while measuring ultrasonic wave velocity. The results are discussed in relation to the reduction of banding obtained from optical microscopy.

  2. Tensile Properties, Ferrite Contents, and Specimen Heating of Stainless Steels in Cryogenic Gas Tests

    SciTech Connect

    Ogata, T.; Yuri, T.; Ono, Y.

    2006-03-31

    We performed tensile tests at cryogenic temperatures below 77 K and in helium gas environment for SUS 304L and SUS 316L in order to obtain basic data of mechanical properties of the materials for liquid hydrogen tank service. We evaluate tensile curves, tensile properties, ferrite contents, mode of deformation and/or fracture, and specimen heating during the testing at 4 to 77 K. For both SUS 304L and 316L, tensile strength shows a small peak around 10 K, and specimen heating decreases above 30 K. The volume fraction of {alpha}-phase increases continuously up to 70 % with plastic strain, at approximately 15 % plastic strain for 304L and up to 35 % for 316L. There was almost no clear influence of testing temperature on strain-induced martensitic transformation at the cryogenic temperatures.

  3. Summary Report of Summer Work: High Purity Single Crystal Growth & Microstructure of Ferritic-Martensitic Steels

    SciTech Connect

    Pestovich, Kimberly Shay

    2015-08-18

    Harnessing the power of the nuclear sciences for national security and to benefit others is one of Los Alamos National Laboratory’s missions. MST-8 focuses on manipulating and studying how the structure, processing, properties, and performance of materials interact at the atomic level under nuclear conditions. Within this group, single crystal scintillators contribute to the safety and reliability of weapons, provide global security safeguards, and build on scientific principles that carry over to medical fields for cancer detection. Improved cladding materials made of ferritic-martensitic alloys support the mission of DOE-NE’s Fuel Cycle Research and Development program to close the nuclear fuel cycle, aiming to solve nuclear waste management challenges and thereby increase the performance and safety of current and future reactors.

  4. Microstructure and fracture behavior of F82H steel under different irradiation and tensile test conditions

    NASA Astrophysics Data System (ADS)

    Wang, K.; Dai, Y.; Spätig, P.

    2016-01-01

    Specimens of martensitic steel F82H were irradiated to doses ranging from 10.7 dpa/850 appm He to 19.6 dpa/1740 appm He at temperatures between 165 and 305 °C in the second experiment of SINQ Target Irradiation Program (STIP-II). Tensile tests were conducted at different temperatures and various fracture modes were observed. Microstructural changes including irradiation-induced defect clusters, dislocation loops and helium bubbles under different irradiation conditions were investigated using transmission electron microscopy (TEM). The deformation microstructures of tensile tested specimens were carefully examined to understand the underlying deformation mechanisms. Deformation twinning was for the first time observed in irradiated martensitic steels. A change of deformation mechanism from dislocation channeling to deformation twinning was observed when the fracture mode changed from rather ductile (quasi-cleavage) to brittle (intergranular or cleavage and intergranular mixed).

  5. Embrittlement of CrMo steels after low fluence irradiation in HFIR

    NASA Astrophysics Data System (ADS)

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

    1995-02-01

    Subsize Charpy impact specimens of 9Cr1MoVNb (modified 9Cr1Mo) and 12Cr1MoVW (Sandvik HT9) steels and 12Cr1MoVW with 2% Ni (12Cr1MoVW2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400°C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toughness. Displacement damage was produced by fast neutrons, and helium was formed by the reaction of 58Ni with thermal neutrons in the mixed-neutron spectrum of HFIR. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr1MoVW2Ni steel irradiated at 400°C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behavior of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

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

    NASA Astrophysics Data System (ADS)

    Margolin, B.; Sorokin, A.; Shvetsova, V.; Minkin, A.; Potapova, V.; Smirnov, V.

    2016-11-01

    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.

  7. Short-term in vitro responses of human peripheral blood monocytes to ferritic stainless steel fiber networks.

    PubMed

    Spear, Rose L; Brooks, Roger A; Markaki, Athina E

    2013-05-01

    Beneficial effects on bone-implant bonding may accrue from ferromagnetic fiber networks on implants which can deform in vivo inducing controlled levels of mechanical strain directly in growing bone. This approach requires ferromagnetic fibers that can be implanted in vivo without stimulating undue inflammatory cell responses or cytotoxicity. This study examines the short-term in vitro responses, including attachment, viability, and inflammatory stimulation, of human peripheral blood monocytes to 444 ferritic stainless steel fiber networks. Two types of 444 networks, differing in fiber cross section and thus surface area, were considered alongside austenitic stainless steel fiber networks, made of 316L, a widely established implant material. Similar high percent seeding efficiencies were measured by CyQuant® on all fiber networks after 48 h of cell culture. Extensive cell attachment was confirmed by fluorescence and scanning electron microscopy, which showed round monocytes attached at various depths into the fiber networks. Medium concentrations of lactate dehydrogenase (LDH) and tumor necrosis factor alpha (TNF-α) were determined as indicators of viability and inflammatory responses, respectively. Percent LDH concentrations were similar for both 444 fiber networks at all time points, whereas significantly lower than those of 316L control networks at 24 h. All networks elicited low-level secretions of TNF-α, which were significantly lower than that of the positive control wells containing zymosan. Collectively, the results indicate that 444 networks produce comparable responses to medical implant grade 316L networks and are able to support human peripheral blood monocytes in short-term in vitro cultures without inducing significant inflammatory or cytotoxic effects.

  8. Effects of activating fluxes on the weld penetration and corrosion resistant property of laser welded joint of ferritic stainless steel

    NASA Astrophysics Data System (ADS)

    Wang, Yonghui; Hu, Shengsun; Shen, Junqi

    2015-10-01

    This study was based on the ferritic stainless steel SUS430. Under the parallel welding conditions, the critical penetration power values (CPPV) of 3mm steel plates with different surface-coating activating fluxes were tested. Results showed that, after coating with activating fluxes, such as ZrO2, CaCO3, CaF2 and CaO, the CPPV could reduce 100~250 W, which indicating the increases of the weld penetrations (WP). Nevertheless, the variation range of WP with or without activating fluxes was less than 16.7%. Compared with single-component ones, a multi-component activating flux composed of 50% ZrO2, 12.09% CaCO3, 10.43% CaO, and 27.49% MgO was testified to be much more efficient, the WP of which was about 2.3-fold of that without any activating fluxes. Furthermore, a FeCl3 spot corrosion experiment was carried out with samples cut from weld zone to test the effects of different activating fluxes on the corrosion resistant (CR) property of the laser welded joints. It was found that all kinds of activating fluxes could improve the CR of the welded joints. And, it was interesting to find that the effect of the mixed activating fluxes was inferior to those single-component ones. Among all the activating fluxes, the single-component of CaCO3 seemed to be the best in resisting corrosion. By means of Energy Dispersive Spectrometer (EDS) testing, it was found that the use of activating fluxes could effectively restrain the loss of Cr element of weld zone in the process of laser welding, thus greatly improving the CR of welded joints.

  9. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  10. Effects of alloying elements and heat treatments on mechanical properties of Korean reduced-activation ferritic-martensitic steel

    NASA Astrophysics Data System (ADS)

    Chun, Y. B.; Kang, S. H.; Noh, S.; Kim, T. K.; Lee, D. W.; Cho, S.; Jeong, Y. H.

    2014-12-01

    As part of an alloy development program for Korean reduced-activation ferritic-martensitic (RAFM) steel, a total of 37 program alloys were designed and their mechanical properties were evaluated with special attention being paid to the effects of alloying elements and heat treatments. A reduction of the normalizing temperature from 1050 °C to 980 °C was found to have a positive effect on the impact resistance, resulting in a decrease in ductile-brittle transition-temperature (DBTT) of the program alloys by an average of 30 °C. The yield strength and creep rupture time are affected strongly by the tempering time at 760 °C but at the expense of ductility. Regarding the effects of the alloying elements, the addition of trace amounts of Zr enhances both the creep and impact resistance: the lowest DBTT was observed for the alloys containing 0.005 wt.% Zr, whereas the addition of 0.01 wt.% Zr extends the creep rupture-time under an accelerated condition. The enhanced impact resistance owing to the normalizing at lower temperature is attributed to a more refined grain structure, which provides more barriers to the propagation of cleavage cracks. Solution softening by Zr addition is suggested as a possible mechanism for enhanced resistance to both impact and creep of the program alloys.

  11. Probing Formability Improvement of Ultra-thin Ferritic Stainless Steel Bipolar Plate of PEMFC in Non-conventional Forming Process

    NASA Astrophysics Data System (ADS)

    Bong, Hyuk Jong; Barlat, Frédéric; Lee, Myoung-Gyu

    2016-08-01

    Formability increase in non-conventional forming profiles programmed in the servo-press was investigated using finite element analysis. As an application, forming experiment on a 0.15-mm-thick ferritic stainless steel sheet for a bipolar plate, a primary component of a proton exchange membrane fuel cell, was conducted. Four different forming profiles were considered to investigate the effects of forming profiles on formability and shape accuracy. The four motions included conventional V motion, holding motion, W motion, and oscillating motion. Among the four motions, the holding motion, in which the slide was held for a certain period at the bottom dead point, led to the best formability. Finite element simulations were conducted to validate the experimental results and to probe the formability improvement in the non-conventional forming profiles. A creep model to address stress relaxation effect along with tool elastic recovery was implemented using a user-material subroutine, CREEP in ABAQUS finite element software. The stress relaxation and variable contact conditions during the holding and oscillating profiles were found to be the main mechanism of formability improvement.

  12. Microstructural evolution in a ferritic-martensitic stainless steel and its relation to high-temperature deformation and rupture models

    SciTech Connect

    DiMelfi, R.J.; Gruber, E.E.; Kramer, J.M.

    1991-01-01

    The ferritic-martensitic stainless steel HT-9 exhibits an anomalously high creep strength in comparison to its high-temperature flow strength from tensile tests performed at moderate rates. A constitutive relation describing its high-temperature tensile behavior over a wide range of conditions has been developed. When applied to creep conditions the model predicts deformation rates orders of magnitude higher than observed. To account for the observed creep strength, a fine distribution of precipitates is postulated to evolve over time during creep. The precipitate density is calculated at each temperature and stress to give the observed creep rate. The apparent precipitation kinetics thereby extracted from this analysis is used in a model for the rupture-time kinetics that compares favorably with observation. Properly austenitized and tempered material was aged over times comparable to creep conditions, and in a way consistent with the precipitation kinetics from the model. Microstructural observations support the postulates and results of the model system. 16 refs., 10 figs.

  13. Microstructure and mechanical property of ferritic-martensitic steel cladding under a 650 °C liquid sodium environment

    NASA Astrophysics Data System (ADS)

    Kim, Jun Hwan; Kim, Sung Ho

    2013-11-01

    A study was carried out to investigate the effect of liquid sodium on the microstructural and mechanical property of ferritic-martensitic steel (FMS) used for a Sodium-cooled Fast Reactor (SFR) cladding tube. A quasi-dynamic device characterized by natural circulation was constructed and a compatibility test between FMS and liquid sodium was performed. HT9 (12Cr-1MoWVN) and Gr.92 (9Cr-2WNbVNB) coupons as well as a Gr.92 cladding tube were immersed in the 650 °C liquid sodium up to 3095 h and a microstructural observation, a mechanical property evaluation such as nanoindentation, and a ring tension test were also done in this study. The results showed that both HT9 and Gr.92 exhibited macroscopic weight loss behavior where pitting and decarburization took place. Weight loss as well as the decarburization process decreased as the chromium content increased. A compatibility test over the cladding tube revealed that a decrease of the mechanical property caused by the aging process governed the whole mechanical property of the cladding tube.

  14. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    NASA Astrophysics Data System (ADS)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  15. Fracture and the formation of sigma phase, M[sub 23]C[sub 6], and austenite from delta-ferrite in an AISI 304L stainless steel

    SciTech Connect

    Tseng, C.C.; Shen, Y.; Thompson, S.W.; Krauss, G. . Dept. of Metallurgical and Materials Engineering); Mataya, M.C. )

    1994-06-01

    The decomposition of delta-ferrite and its effects on tensile properties and fracture of a hot-rolled AISI 304L stainless steel plate were studied. Magnetic response measurements of annealed specimens showed that the transformation rate of delta-ferrite was highest at 720 C. Transformation behavior was characterized by light microscopy, transmission electron microscopy, scanning electron microscopy, and energy-dispersive spectroscopy on thin foils. The initial transformation of delta-ferrite ([delta]) to austenite ([gamma]) and a chromium-rich carbide (M[sub 23]C[sub 6]) occurred by a lamellar eutectoid reaction, [sigma] [r reversible] M[sub 23]C[sub 6] + [gamma]. The extent of the reaction was limited by the low carbon content of the 304L plate, and the numerous, fine M[sub 23]C[sub 6] particles of the eutectoid structure provide microvoid nucleation sites in tensile specimens annealed at 720 C for short times. Sigma phase ([sigma]) formed as a result of a second eutectoid reaction, [delta] [r reversible] [sigma] + [gamma]. Brittle fracture associated with the plate-shaped sigma phase of the second eutectoid structure resulted in a significant decrease in reduction of area (RA) in the transverse tensile specimens. The RA for longitudinal specimens was not affected by the formation of sigma phase. Tensile strengths were little affected by delta-ferrite decomposition products in either longitudinal or transverse orientations.

  16. Effect of cold work on tensile behavior of irradiated type 316 stainless steel

    SciTech Connect

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

    1986-01-01

    Tensile specimens were irradiated in ORR at 250, 290, 450, and 500/sup 0/C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290/sup 0/C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450/sup 0/C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500/sup 0/C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500/sup 0/C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes.

  17. Preparation and performances of Co-Mn spinel coating on a ferritic stainless steel interconnect material for solid oxide fuel cell application

    NASA Astrophysics Data System (ADS)

    Zhang, H. H.; Zeng, C. L.

    2014-04-01

    Ferritic stainless steels have become the candidate materials for interconnects of intermediate temperature solid oxide fuel cell (SOFC). The present issues to be solved urgently for the application of ferritic stainless steel interconnects are their rapid increase in contact resistance and Cr poisoning. In the present study, a chloride electrolyte suspension has been developed to electro-deposit a Co-Mn alloy on a type 430 stainless steel, followed by heat treatment at 750 °C in argon and at 800 °C in air to obtain Co-Mn spinel coatings. The experimental results indicate that an adhesive and compact Co-Mn alloy layer can be deposited in the chloride solution. After heat treatment, a complex coating composed of an external MnCo2O4 layer and an inner Cr-rich oxide layer has been formed on 430SS. The coating improves the oxidation resistance of the steel at 800 °C in air, especially in wet air, and inhibits the outward diffusion of Cr from the Cr-rich scale. Moreover, a low contact resistance has been achieved with the application of the spinel coatings.

  18. Comparison of the effects of long-term thermal aging and HFIR irradiation on the microstructural evolution of 9Cr-1MoVNb steel

    SciTech Connect

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

    1990-01-01

    Both thermal aging at 482--704{degree}C for up to 25,000h and HFIR irradiation at 300--600{degree}C for up to 39 dpa produce substantial changes in the as-tempered microstructure of 9Cr-1MoVNb martensitic/ferritic steel. However, the changes in the dislocation/subgrain boundary and the precipitate structures caused by thermal aging or neutron irradiation are quite different in nature. During thermal aging, the as-tempered lath/subgrain boundary and carbide precipitate structures remain stable below 650{degree}C, but coarsen and recover somewhat at 650--704{degree}C. The formation of abundant intergranular Laves phase, intra-lath dislocation networks, and fine dispersions of VC needles are thermal aging effects that are superimposed upon the as-tempered microstructure at 482--593{degree}C. HFIR irradiation produces dense dispersions of very small black-dot'' dislocations loops at 300{degree}C and produces helium bubbles and voids at 400{degree}C At 300--500{degree}C, there is considerable recovery of the as-tempered lath/subgrain boundary structure and microstructural/microcompositional instability of the as-tempered carbide precipitates during irradiation. By contrast, the as-tempered microstructure remains essentially unchanged during irradiation at 600{degree}C. Comparison of thermally aged with irradiation material suggests that the instabilities of the as-tempered lath/subgrain boundary and precipitate structures at lower irradiation temperatures are radiation-induced effects, whereas the absence of both Laves phase and fine VC needles during irradiation is a radiation-retarded thermal effect.

  19. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    SciTech Connect

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  20. Evaluation of stress corrosion cracking of irradiated 304L stainless steel in PWR environment using heavy ion irradiation

    NASA Astrophysics Data System (ADS)

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

    2016-08-01

    IASCC has been a major concern regarding the structural and functional integrity of core internals of PWR's, especially baffle-to-former bolts. Despite numerous studies over the past few decades, additional evaluation of the parameters influencing IASCC is still needed for an accurate understanding and modeling of this phenomenon. In this study, Fe irradiation at 450 °C was used to study the cracking susceptibility of 304 L austenitic stainless steel. After 10 MeV Fe irradiation to 5 dpa, irradiation-induced damage in the microstructure was characterized and quantified along with nano-hardness measurements. After 4% plastic strain in a PWR environment, quantitative information on the degree of strain localization, as determined by slip-line spacing, was obtained using SEM. Fe-irradiated material strained to 4% in a PWR environment exhibited crack initiation sites that were similar to those that occur in neutron- and proton-irradiated materials, which suggests that Fe irradiation may be a representative means for studying IASCC susceptibility. Fe-irradiated material subjected to 4% plastic strain in an inert argon environment did not exhibit any cracking, which suggests that localized deformation is not in itself sufficient for initiating cracking for the irradiation conditions used in this study.

  1. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    SciTech Connect

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  2. Fatigue performance and cyclic softening of F82H, a ferritic martensic steel

    SciTech Connect

    Stubbins, J.F.; Gelles, D.S.

    1996-04-01

    The room temperature fatigue performance of F82H has been examined. The fatigue life was determined in a series of strain-controlled tests where the stress level was monitored as a function of the number of accrued cycles. Fatigue lives in the range of 10{sup 3} to 10{sup 6} cycles to failure were examined. The fatigue performance was found to be controlled primarily by the elastic strain range over most of the range of fatigue lives examined. Only at low fatigue lives did the plastic strain range contribute to the response. However, when the significant plastic strain did contribute, the material showed a tendency to cyclically soften. That is the load carrying capability of the material degrades with accumulated fatigue cycles. The overall fatigue performance of the F82H alloy was found to be similiar to other advanced martensitic steels, but lower than more common low alloy steels which possess lower yield strengths.

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

    SciTech Connect

    Ashdown, B.G.

    1980-04-01

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

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

    SciTech Connect

    Not Available

    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)

  5. A simplified LBB evaluation procedure for austenitic and ferritic steel piping

    SciTech Connect

    Gamble, R.M.; Wichman, K.R.

    1997-04-01

    The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

  6. Boron effect on the microstructure of 9% Cr ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Klimenkov, M.; Materna-Morris, E.; Möslang, A.

    2015-07-01

    The microstructure of reduces-activation 9Cr-WTaV steel alloyed with 83 and 1160 wt. ppm 10B was detailed analysed using transmission electron microscopy. The influence of boron content on the precipitation behaviour of M23C6 and MX (VN and TaC) phases and, hence, on the formation process of steel's grain and lath structure was studied. VN precipitates, which play an important role in the stabilisation of the lath structure, exhibit most sensitive reaction on presence of boron. Their spatial density significantly reduces in the alloy with 83 ppm boron. In the steel with 1160 wt. ppm boron, no formation of VN was detected, whereas TaC particles precipitate at the lath and grain boundaries. These changes in the structure stabilisation mechanism lead to an increasing lath width and a decreasing thermal stability of laths and grains. Analytical investigations of several BN particles reveal their complex multi-phase structure and allow conclusions to be drawn with respect to their precipitation sequence.

  7. Void formation and helium effects in 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated in HFIR and FFTF at 400/degree/C

    SciTech Connect

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

    1988-01-01

    Martensitic/ferritic 9Cr-1MoVNb and 12Cr-1MoVW steels doped with up to 2 wt% Ni have up to 450 appm He after HFIR irradiation to /approximately/38 dpa, but only 5 appm He after 47 dpa in FFTF. No fine He bubbles and few or no larger voids were observable in any of these steels after FFTF irradiation at 407/degree/C. By contrast, many voids were found in the undoped steels (30-90 appm He) irradiated in HFIR at 400/degree/C, while voids plus many more fine He bubbles were found in the Ni-doped steels (400-450 appm He). Irradiation in both reactors at /approximately/400/degree/C produced significant changes in the as-tempered lath/subgrain boundary, dislocation, and precipitation structures that were sensitive to alloy composition, including doping with Ni. However, for each specific alloy the irradiation-produced changes were exactly the same comparing samples irradiated in FFTF and HFIR, particularly the Ni-doped steels. Therefore, the increased void formation appears solely due to the increased helium generation found in HFIR. While the levels of void swelling are relatively low after 37-39 dpa in HFIR (0.1-0.4%), details of the microstructural evolution suggest that void nucleation is still progressing, and swelling could increase with dose. The effect of helium on void swelling remains a valid concern for fusion application that requires higher dose experiments. 15 refs., 14 figs., 8 tabs.

  8. Thermally activated deformation of irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  9. Evaluation of irradiated pressure vessel steel by mechanical tests and positron annihilation lineshape analysis

    SciTech Connect

    Nakamura, Noriko; Ohta, Yoshio; Yoshida, Kazuo; Maeda, Noriyoshi

    1999-10-01

    Mechanical test and positron annihilation lineshape analysis have been performed on neutron irradiated pressure vessel steels, A533B1 steel and the weld metal. Marked changes in the mechanical properties were observed for both metals after the neutron exposure. S-parameters, the positron annihilation parameters, also increased after the neutron irradiation but only the small change was observed in the different levels of neutron fluence. The change in S-parameter and the mechanical properties were well correlated. It is concluded that changes in embrittlement induced by radiation can be monitored by positron annihilation lineshape analysis but detectability is dependent on the materials.

  10. Characterization of TiN, TiC and Ti(C,N) in titanium-alloyed ferritic chromium steels focusing on the significance of different particle morphologies

    SciTech Connect

    Michelic, S.K.; Loder, D.; Reip, T.; Ardehali Barani, A.; Bernhard, C.

    2015-02-15

    Titanium-alloyed ferritic chromium steels are a competitive option to classical austenitic stainless steels owing to their similar corrosion resistance. The addition of titanium significantly influences their final steel cleanliness. The present contribution focuses on the detailed metallographic characterization of titanium nitrides, titanium carbides and titanium carbonitrides with regard to their size, morphology and composition. The methods used are manual and automated Scanning Electron Microscopy with Energy Dispersive X-ray Spectroscopy as well as optical microscopy. Additional thermodynamic calculations are performed to explain the precipitation procedure of the analyzed titanium nitrides. The analyses showed that homogeneous nucleation is decisive at an early process stage after the addition of titanium. Heterogeneous nucleation gets crucial with ongoing process time and essentially influences the final inclusion size of titanium nitrides. A detailed investigation of the nuclei for heterogeneous nucleation with automated Scanning Electron Microscopy proved to be difficult due to their small size. Manual Scanning Electron Microscopy and optical microscopy have to be applied. Furthermore, it was found that during solidification an additional layer around an existing titanium nitride can be formed which changes the final inclusion morphology significantly. These layers are also characterized in detail. Based on these different inclusion morphologies, in combination with thermodynamic results, tendencies regarding the formation and modification time of titanium containing inclusions in ferritic chromium steels are derived. - Graphical abstract: Display Omitted - Highlights: • The formation and modification of TiN in the steel 1.4520 was examined. • Heterogeneous nucleation essentially influences the final steel cleanliness. • In most cases heterogeneous nuclei in TiN inclusions are magnesium based. • Particle morphology provides important information

  11. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    NASA Astrophysics Data System (ADS)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  12. He and H irradiation effects on the nanoindentation hardness of CLAM steel

    NASA Astrophysics Data System (ADS)

    Jiang, Siben; Peng, Lei; Ge, Hongen; Huang, Qunying; Xin, Jingping; Zhao, Ziqiang

    2014-12-01

    In this study, He and H ion irradiation induced hardening behavior of China Low Activation Martensitic (CLAM) steel was investigated, and the influence of Si on irradiation hardening was also examined. CLAM steel with different Si contents, Heat 0912 and Heat 0408D, were irradiated with single He (He concentration range from 0 to 2150 appm) ion beam and He/H dual ion beams. Then nanoindentation tests were applied to evaluate the ion irradiation induced hardening effect. The result of Heat 0912 showed hardening effect would be more serious with higher He concentration, and the trend saturated when He concentration reach 1000 appm. Comparing the result of Heat 0912 and Heat 0408D, higher Si content might improve the resistance to hardening.

  13. Temperature dependence of the deformation behavior of 316 stainless steel after low temperature neutron irradiation

    SciTech Connect

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

    1996-10-01

    The effects of low temperature neutron irradiation on the tensile behavior of 316 stainless steel have been investigated. A single heat of solution annealed 316 was irradiated to 7 and 18 dpa at 60, 200, 330, and 400{degrees}C. The tensile properties as a function of dose and as a function of temperature were examined. Large changes in yield strength, deformation mode, strain to necking, and strain hardening capacity were seen in this irradiation experiment. The magnitudes of the changes are dependent on both irradiation temperature and neutron dose. Irradiation can more than triple the yield strength over the unirradiated value and decrease the strain to necking (STN) to less than 0.5% under certain conditions. A maximum increase in yield strength and a minimum in the STN occur after irradiation at 330{degrees}C but the failure mode remains ductile.

  14. Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

    SciTech Connect

    Grossbeck, M.L.; Liu, K.C.

    1980-01-01

    Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by a factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.

  15. The microstructure of neutron irradiated type-348 stainless steel and its relation to creep and hardening

    NASA Astrophysics Data System (ADS)

    Thomas, L. E.; Beeston, J. M.

    1982-06-01

    Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350°C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1-2 nm diameter helium bubbles, 2-20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel suicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding.

  16. Effect of irradiation on the steels 316L/LN type to 12 dpa at 400 °C

    NASA Astrophysics Data System (ADS)

    Bulanova, T.; Fedoseev, A.; Kalinin, G.; Rodchenkov, B.; Shamardin, V.

    2004-08-01

    The 316L type stainless steel is widely used as a structural material for the fission reactors internal structures (core, core supports, etc.) and for experimental irradiation facilities. The 316L(N)-IG type steel is proposed as a main structural material for the ITER reactor (first wall, blanket, vacuum vessel, cooling pipe lines). It is obvious that different steel grades should exhibit different reaction to neutron irradiation. The main objective of this work was to study of irradiation behaviour of three different commercial steels: AISI 316LN, AISI 316L (US grades) and 02X17H14M2 (Russian steel grade that is similar to 316L). Irradiation effect on the three commercial steels of 316L family to ˜12 dpa at the temperature ˜370-400 °C on the tensile properties, microstructure, swelling and susceptibility to SCC are described in the paper.

  17. Neutron Irradiation Effects on Fatigue Crack Propagation in Type 316 Stainless Steels at 649 C.

    DTIC Science & Technology

    1980-08-01

    maintaining the maximum tensile load constant for selected time periods duriig each cycle. Induction heating was employed to achieve a test temperature of...7 AD-A OB 052 NAVAL RESEAR ICH LAS WASHINGTON DC F/6 11/BNEUTRON IRRADIATION EFFECTS ON FATIGUE CRACK PROPAGATION IN TYP-EC(UlU UNL AUG 80 0 .J...and Identify by block number) Radiation Microstruture B Irradiation Fatigue Stainless steels Crack propagation Radiation effects High temperature V

  18. Swift heavy ion irradiated spinel ferrite: A cheap radiation resistant material

    NASA Astrophysics Data System (ADS)

    Satalkar, M.; Kane, S. N.; Kulriya, P. K.; Avasthi, D. K.

    2016-07-01

    Effect of (80 MeV) 16O 6+ ion irradiation on the structural properties and cation distribution of the as-burnt samples (i.e. the samples are without any thermal/sintering treatment) with the following compositions: MnFe2O4, Mn0.5Zn0.5Fe2O4 and ZnFe2O4 prepared by sol-gel auto-combustion technique have been studied through in-situ and ex-situ X-ray diffraction (XRD) technique. Well characterized single phase MnFe2O4 and Mn0.5Zn0.5Fe2O4 samples were irradiated at fluence 1 × 1011, 1 × 1012, 1 × 1013 and 1 × 1014 ions/cm2 to see the effect of the electronic energy loss induced changes in the structural properties and in cation distribution monitored through ex-situ XRD. ZnFe2O4 samples were irradiated with ion fluence values ranging between 1 × 1011 - 2 × 1014 ions/cm2 to observe the effect of in-situ XRD on structural properties and cation distribution. Results very clearly depict the redistribution of cations in the samples, which show noticeable changes in: ionic radii of A-site (rA) and B-site (rB), experimental and theoretical lattice parameter (aexp.,ath.), unit cell volume (V), Scherrer's Grain diameter (D), oxygen positional parameter (u), tetrahedral and octahedral bond length (RA, RB), shared tetrahedral and octahedral edge (dAE,dBE) and bond angles (θ1, θ2, θ3, θ4, θ5). Results are interpreted in terms of irradiation induced changes in the above mentioned parameters.

  19. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale J.; Was, Gary S.

    2014-01-01

    The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

  20. Temperature dependence of fracture toughness in HT9 steel neutron-irradiated up to 145 dpa

    SciTech Connect

    Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, S; Toloczko, M

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to high doses was investigated using miniature three-point bend (TPB) fracture specimens. These specimens were from the ACO-3 fuel duct wall of the Fast Flux Test Facility (FFTF), in which irradiation doses were in the range of 3.2 144.8 dpa and irradiation temperatures in the range of 380.4 502.6 oC. A miniature specimen reuse technique has been established for this investigation: the specimens used were the tested halves of miniature Charpy impact specimens (~13 3 4 mm) with diamond-saw cut in the middle. The fatigue precracking for specimens and fracture resistance (J-R) tests were carried out in a MTS servo-hydraulic testing machine with a vacuum furnace following the standard procedure described in the ASTM Standard E 1820-09. For each of five irradiated and one archive conditions, 7 to 9 J-R tests were performed at selected temperatures ranging from 22 C to 600 C. The fracture toughness of the irradiated HT9 steel was strongly dependent on irradiation temperatures rather than irradiation dose. When the irradiation temperature was below about 430 C, the fracture toughness of irradiated HT9 increased with test temperature, reached an upper shelf of 180 200 MPa m at 350 450 C and then decreased with test temperature. When the irradiation temperature 430 C, the fracture toughness was nearly unchanged until about 450 C and decreased with test temperature in higher temperature range. Similar test temperature dependence was observed for the archive material although the highest toughness values are lower after irradiation. Ductile stable crack growth occurred except for a few cases where both the irradiation temperature and test temperature are relatively low.

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

    SciTech Connect

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

    1998-03-01

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

  2. Heavy-section steel irradiation program. Progress report, October 1992--March 1993

    SciTech Connect

    Corwin, W.R.

    1998-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is one of only two more safety-related components of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV`s fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established at Oak Ridge National Laboratory (ORNL) under sponsorship of the Nuclear Regulatory Commission (NRC). The primary goal of this major safety program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (in particular, the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program centers on experimental assessments of irradiation-induced embrittlement (including the completion of certain irradiation studies previously conducted by the Heavy-Section Steel Technology Program) augmented by detailed examinations and modeling of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties.

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

    NASA Astrophysics Data System (ADS)

    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. On the shape of the magnetic Barkhausen noise profile for better revelation of the effect of microstructures on the magnetisation process in ferritic steels

    NASA Astrophysics Data System (ADS)

    Vashista, M.; Moorthy, V.

    2015-11-01

    The shape of the Magnetic Barkhausen Noise (MBN) profiles has been compared for two different methods of MBN measurements in order to reveal the true extent of the influence of different carbon-content related microstructures on the magnetisation process. The MBN profiles were measured using high frequency and low frequency MBN measurement systems on samples from low carbon 18CrNiMo5 steel and high carbon 42CrMo4 steel heat treated by isothermal annealing, spheroidising annealing and quenching and tempering processes. The high frequency MBN (HFMBN) profile shows only a single peak for all the samples due to insufficient applied magnetic field strength and shallow skin-depth of detection of HFMBN signals. The low frequency MBN (LFMBN) profile shows two peaks for all the samples due to larger magnetisation range revealing the difference in the interaction of domain walls with different microstructural features such as ferrite, pearlite, martensite and carbides. The shape of the LFMBN profile shows systematic and distinct variation in the magnetisation process with respect to carbon content and different microstructures. This study shows that the LFMBN profile reveals distinct changes in shape which could be successfully used for characterisation of different microstructural phases in ferritic steels.

  5. High-temperature degradation and protection of ferritic and austenitic steels in steam generators

    NASA Astrophysics Data System (ADS)

    Martínez-Villafañe, A.; Almeraya-Calderón, M. F.; Gaona-Tiburcio, C.; Gonzalez-Rodriguez, J. G.; Porcayo-Calderón, J.

    1998-02-01

    The useful life of superheaters and reheaters of power stations which use heavy fuel oil is shortened and their continuous service is inhibited by corrosion (fireside) and creep-type problems. The increase of corrosion attack on boilers is caused by the presence of fuel ash deposits containing mainly vanadium, sodium, and sulfur which form low-melting-point compounds. The tubes are exposed to the action of high stresses and high temperatures, producing the so-called “creep damage.” In this work, two kinds of results are reported: lab and field studies using a 2.25Cr-1Mo steel. The laboratory work was in turn divided into two parts. In the first, the steel was exposed to the action of natural ash deposits in oxidant atmospheres at 600 ° for 24 h. In the second part, tensile specimens were creep tested in Na2SO4, V2O5, and their mixture over a temperature range of 580 to 620 °. In the field work, components of a power station were coated with different types of nickel-and iron-base coatings containing chromium, Fe-Cr, and Fe-Si using the powder flame spraying technique. After testing, the coated tubes were analyzed using electron microscopy. The results showed that all the coating systems had good corrosion resistance, especially those containing silicon or chromium.

  6. Structural and magnetic properties of nano-sized NiCuZn ferrites synthesized by co-precipitation method with ultrasound irradiation

    NASA Astrophysics Data System (ADS)

    Harzali, Hassen; Saida, Fairouz; Marzouki, Arij; Megriche, Adel; Baillon, Fabien; Espitalier, Fabienne; Mgaidi, Arbi

    2016-12-01

    Sonochemically assisted co-precipitation has been used to prepare nano-sized Ni-Cu-Zn-ferrite powders. A suspension of constituent hydroxides was ultrasonically irradiated for various times at different temperatures with high intensity ultrasound radiation using a direct immersion titanium horn. Structural and magnetic properties were investigated using X-diffraction (XRD), FT-IR spectroscopy, transmission electron microscopy (TEM), Nitrogen adsorption at 77 K (BET) and Vibrating sample magnetometer (VSM). Preliminary experimental results relative to optimal parameters showed that reaction time t=2 h, temperature θ=90 °C and dissipated Power Pdiss=46.27 W. At these conditions, this work shows the formation of nanocrystalline single-phase structure with particle size 10-25 nm. Also, ours magnetic measurements proved that the sonochemistry method has a great influence on enhancing the magnetic properties of the ferrite.

  7. In Situ Observation of Phase Transformation and Structure Evolution of a 12 pct Cr Ferritic Stainless Steel

    NASA Astrophysics Data System (ADS)

    Song, Changjiang; Guo, Yuanyi; Li, Kefeng; Sun, Fengmei; Han, Qingyou; Zhai, Qijie

    2012-10-01

    This work focuses on an in situ observation of phase transformation of a 12 pct Cr ferritic stainless steel using high-temperature laser scanning confocal microscopy. α→ γ→ δ phase transformation temperatures are determined to be approximately 1073 K and 1423 K (800 °C and 1150 °C), respectively. The onset of phase transformation is found to occur at grain boundaries. When the temperature is beyond 1518 K (1245 °C), the grain growth rate suddenly becomes very high, and the grain growth is related to the self-organizing of adjacent grains. δ→ γ phase transformation has been mostly restrained when cooling rates are in the range of 22.4 K/s to 13.3 K/s (22.4 °C/s to 13.3 °C/s) except for at grain boundaries. Martensitic phase transformation, rather than γ→ α phase transformation, occurs when the cooling rates are in the range of 8.5 K/s to 2.2 K/s (8.5 °C/s to 2.2 °C/s). The starting temperature of martensitic phase transformation is approximately 697 K to 728 K (424 °C to 455 °C) for specimens heated to 1373 K (1100 °C) ( i.e., γ phase field), which is 50 K to 100 K (50 °C to 100 °C) higher than that of specimens heated to 1723 K (1450 °C) ( i.e., δ phase field). Many bulges remain on the surfaces of the specimen heated to 1723 K (1450 °C), and their formation mechanism is analyzed.

  8. Formation Mechanism of Type IV Failure in High Cr Ferritic Heat-Resistant Steel-Welded Joint

    NASA Astrophysics Data System (ADS)

    Liu, Y.; Tsukamoto, S.; Shirane, T.; Abe, F.

    2013-10-01

    The mechanism of type IV failure has been investigated by using a conventional 9Cr ferritic heat-resistant steel Gr.92. In order to clarify the main cause of type IV failure, different heat treatments were performed on the base metal in order to change the prior austenite grain (PAG) size and precipitate distribution after applying the heat-affected zone (HAZ) simulated thermal cycle at the peak temperature of around A c3 ( A c3 HAZ thermal cycle) and postweld heat treatment (PWHT). The microstructural evolution during the A c3 HAZ thermal cycle and PWHT was investigated by means of scanning electron microscope (SEM), electron backscatter diffraction (EBSD), electron probe microanalysis (EPMA), and transmission electron microscope (TEM). It was found that M23C6 carbides were scarcely precipitated at the newly formed fine PAG, block, and lath boundaries in A c3 HAZ-simulated Gr.92, because the carbide forming elements such as Cr and C were segregated at the former PAG and block boundaries of the base metal. On the other hand, if all the boundaries were covered by sufficient M23C6 carbides by homogenization of the alloying elements prior to applying the HAZ thermal cycle, the creep strength was much improved even if the fine PAG was formed. From these results, it is concluded that fine-grained microstructure cannot account for the occurrence of type IV failure, and it only has a small effect during long-term creep. The most important factor is the precipitate formation behavior at various boundaries. Without sufficient boundary strengthening by precipitates, the microstructure of A c3 HAZ undergoes severe changes even during PWHT and causes premature failure during creep.

  9. High-Temperature Tensile Properties of Nano-Oxide Dispersion Strengthened Ferritic Steels Produced by Mechanical Alloying and Spark Plasma Sintering

    NASA Astrophysics Data System (ADS)

    Boulnat, Xavier; Fabregue, Damien; Perez, Michel; Mathon, Marie-Hélène; de Carlan, Yann

    2013-06-01

    Oxide-dispersion strengthened (ODS) ferritic steels were produced by mechanical alloying and subsequent spark plasma sintering. Very fast heating rates were used to minimize porosity when controlling grain size and precipitation of dispersoids within a compacted material. Sintering cycles performed at 1373 K (1100 °C) induced heterogeneous, but fine grain size distribution and high density of nano-oxides. Yield strengths at room temperature and at 923 K (650 °C) are 975 MPa and 298 MPa, respectively. Furthermore, high-temperature ductility is much increased: total strain of 28 pct at 923 K (650 °C).

  10. The role of dislocation channeling in IASCC initiation of neutron irradiated austenitic stainless steel

    NASA Astrophysics Data System (ADS)

    Stephenson, Kale Jennings

    The objective of this study was to understand the role of dislocation channeling in the initiation of irradiation-assisted stress corrosion cracking (IASCC) of neutron irradiated austenitic stainless steel using a novel four-point bend test. Stainless steels used in this study were irradiated in the BOR-60 fast reactor at 320 °C, and included a commercial purity 304L stainless steel irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity stainless steels, Fe-18Cr-12Ni and Fe-18Cr-25Ni, irradiated to ~10 dpa. The four-point bend test produced the same relative IASCC susceptibility as constant extension rate tensile (CERT) experiments performed on the same irradiated alloys in boiling water reactor normal water chemistry. The cracking susceptibility of the CP 304L alloy was high at all irradiation dose levels, enhanced by the presence of MnS inclusions in the alloy microstructure, which dissolve in the NWC environment. Dissolution of the MnS inclusion results in formation of an oxide cap that occludes the inclusion site, creating a crevice condition with a high propensity for crack initiation. Crack initiation at these locations was induced by stress concentration at the intersecting grain boundary, resulting from the intersection of a discontinuous dislocation channels (DC). Stress to initiate an IASCC crack decreased with dose due earlier DC initiation. The HP Fe-18Cr-12Ni alloy had low susceptibility to IASCC, while the high Ni alloy exhibited no cracking susceptibility. The difference in susceptibility among these conditions was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC -- grain boundary intersections.

  11. Evolution of microstructure at hot band annealing of ferritic FeSi steels

    NASA Astrophysics Data System (ADS)

    Schneider, Jürgen; Li, Guangqiang; Franke, Armin; Zhou, Bowen

    2017-02-01

    The magnetic properties of the finally fabricated nonoriented FeSi steels critically depend on the microstructure and on the occurring crystallographic texture. The fabrication route comprises hot rolling, coiling and cooling, hot band annealing before cold rolling (optional), cold rolling and the final thermal treatment. As well known there is an interplay between the microstructure and texture during the various processing steps. For that reason, it is of interest to know more on the evolution of the microstructure at hot band annealing of hot band prepared in different ways. In this paper we will summarize our recent results on the evolution of microstructure during thermal annealing of hot band: thermal treatment following immediately the last pass of hot rolling or a hot band annealing as a separate processing step before cold rolling.

  12. Room temperature ductility of NiAl-strengthened ferritic steels: Effects of precipitate microstructure

    SciTech Connect

    Teng, Z.K.; Liu, C.T.; Miller, M.K.; Ghosh, G.; Kenik, E.A.; Huang, S.; Liaw, P.K.

    2012-04-11

    The effects of precipitate microstructure on the room temperature ductility of a series of carefully designed Fe-Al-Ni-Cr-Mo steels were investigated. Transmission electron microscopy (TEM), ultra small angle X-ray scattering (USAXS), and atom probe tomography (APT) were conducted to quantify the nano-scaled precipitates. The accuracy of the characterization results was verified by a numerical analysis. Three point bending tests results demonstrated that ductility was a function of the precipitate volume fraction and the Al and Ni concentrations in the Fe matrix, these relationships were discussed in terms of possible mechanisms. The ductility was also found to be independent of the precipitate size and inter-particle spacing in the studied range, which was validated by a theoretical model.

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

    SciTech Connect

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

    1995-04-01

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

  14. Effect of nickel content on the neutron irradiation embrittlement of Ni-Mo-Cr steels

    NASA Astrophysics Data System (ADS)

    Lee, Chang-Hoon; Kasada, R.; Kimura, A.; Lee, Bong-Sang; Suh, Dong-Woo; Lee, Hu-Chul

    2013-11-01

    The influence of nickel on the neutron irradiation embrittlement of Ni-Mo-Cr reactor pressure vessel (RPV) steels was investigated using alloys containing nickel in the range of 0.9-3.5 wt%. In all investigated alloys, the neutron irradiation with two dose conditions of 4.5 × 1019 neutron/cm2 at 290 °C and 9.0 × 1019 neutron/cm2 at 290 °C, respectively, increased the hardness and ductile-to-brittle transition temperature (DBTT). However, the increases of the hardness and DBTT resulting from the neutron irradiation were primarily affected by the irradiation dose that is closely related to the generation of irradiation defects, but not by the nickel content. In addition, a linear relationship between the changes in the hardness and DBTT subjected to the irradiation was confirmed. These results demonstrate that increasing the nickel content up to 3.5 wt% does not have a harmful effect on the irradiation embrittlement of Ni-Mo-Cr reactor pressure vessel (RPV) steels.

  15. Hardness of Carburized Surfaces in 316LN Stainless Steel after Low Temperature Neutron Irradiation

    SciTech Connect

    Byun, TS

    2005-01-31

    A proprietary surface carburization treatment is being considered to minimize possible cavitation pitting of the inner surfaces of the stainless steel target vessel of the SNS. The treatment gives a large supersaturation of carbon in the surface layers and causes substantial hardening of the surface. To answer the question of whether such a hardened layer will remain hard and stable during neutron irradiation, specimens of the candidate materials were irradiated in the High Flux Isotope Reactor (HFIR) to an atomic displacement level of 1 dpa. Considerable radiation hardening occurred in annealed 316LN stainless steel and 20% cold rolled 316LN stainless steel, and lesser radiation hardening in Kolsterised layers on these materials. These observations coupled with optical microscopy examinations indicate that the carbon-supersaturated layers did not suffer radiation-induced decomposition and softening.

  16. Microstructure and microhardness of CLAM steel irradiated up to 20.8 dpa in STIP-V

    NASA Astrophysics Data System (ADS)

    Peng, Lei; Ge, Hongen; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic (CLAM) steel were irradiated in the fifth experiment of SINQ target irradiation program (STIP-V) up to 20.8 dpa/1564 appm He. Microhardness measurements and transmission electron microscope (TEM) observations have been performed to investigate irradiation induced hardening effects. The results of CLAM steel specimens show similar trend in microhardness and microstructure changes with irradiation dose, compared to F82H/Optimax-A steels irradiated in STIP-I/II. Defects and helium bubbles were observed in all specimens, even at a very low dose of 5.4 dpa. For defects and bubbles, the mean size and number density increased with increasing irradiation dose to 13 dpa, and then the mean size increased and number density decreased with the increasing irradiation dose to 20.8 dpa.

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

    SciTech Connect

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

    2008-01-21

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

  18. Investigation of the Kinetics of the Ferrite/Austenite Phase Transformation in the HAZ of a 2205 Duplex Stainless Steel Weldment

    SciTech Connect

    Palmer, T A; Elmer, J W; Wong, J; Babu, S S; Vitek, J M

    2002-03-14

    A semi-quantitative map based on a series of spatially resolved X-ray diffraction (SRXRD) scans shows the progression of the ferrite ({delta})/austenite ({gamma}) phase balance throughout the HAZ during GTA welding of a 2205 duplex stainless steel (DSS). This map shows an unexpected decrease in the ferrite fraction on heating, followed by a recovery to the original ferrite fraction on cooling at locations within the HAZ. Even though such behavior is supported by thermodynamic calculations, it has not been confirmed by either experimental methods or have the kinetics been evaluated. Both Gleeble thermal simulations and time resolved x-ray diffraction measurements on spot welds in the 2205 DSS provide further evidence for this rather low-temperature transformation. On the other hand, calculations of the diffusion of alloying elements across the 6/y interface under a variety of conditions shed no further light on the driving force for this transformation. Further work on the mechanisms and driving forces for this transformation is on-going.

  19. Irradiation-induced embrittlement of a 2.25Cr1Mo steel

    NASA Astrophysics Data System (ADS)

    Song, S.-H.; Faulkner, R. G.; Flewitt, P. E. J.; Smith, R. F.; Marmy, P.; Victoria, M.

    2000-07-01

    Irradiation-induced embrittlement of a 2.25Cr1Mo is investigated by means of small punch testing and scanning electron microscopy (SEM). The ductile-brittle transition temperature (DBTT) determined by the small punch test is much lower than that determined by the standard Charpy test. There are some irradiation-induced embrittlement effects after the steel is irradiated at about 270°C for 46 days with a neutron dose rate of 1.05×10 -8 dpa s -1 and at about 400°C for 86 days with a neutron dose rate of 1.75×10 -8 dpa s -1. In addition, there is some temper embrittlement after the steel is aged at about 400°C for 86 days.

  20. Silicon-containing ferritic/martensitic steel after exposure to oxygen-containing flowing lead-bismuth eutectic at 450 and 550 °C

    NASA Astrophysics Data System (ADS)

    Schroer, Carsten; Koch, Verena; Wedemeyer, Olaf; Skrypnik, Aleksandr; Konys, Jürgen

    2016-02-01

    A ferritic/martensitic (f/m) steel with 9 and 3 mass% of chromium (Cr) and silicon (Si), respectively, was tested on performance in flowing lead-bismuth eutectic (LBE) at 450 and 550 °C, each at concentrations of solved oxygen of both 10-7 and 10-6 mass%. The 9Cr-3Si steel generally exhibits the same basic corrosion modes as other f/m materials with 9 mass% Cr and typically lower Si content, namely Steel T91. The Si-rich steel shows an overall improved performance in comparison to T91 at 450 °C and 10-7 mass% solved oxygen, but especially at 450 °C and 10-6 mass% solved oxygen. The advantage of higher Si-content in 9Cr steel is less clear at 550 °C. Especially high oxygen content in flowing LBE at 550 °C, between >10-6 mass% and oxygen saturation, seems detrimental for the high-Si material in respect of the initiation and progress of a solution-based corrosion.

  1. Effect of Tungsten on Long-Term Microstructural Evolution and Impression Creep Behavior of 9Cr Reduced Activation Ferritic/Martensitic Steel

    NASA Astrophysics Data System (ADS)

    Thomas Paul, V.; Vijayanand, V. D.; Sudha, C.; Saroja, S.

    2017-01-01

    The present study describes the changes in the creep properties associated with microstructural evolution during thermal exposures to near service temperatures in indigenously developed reduced activation ferritic-martensitic steels with varying tungsten (1 and 1.4 wt pct W) contents. The creep behavior has been studied employing impression creep (IC) test, and the changes in impression creep behavior with tungsten content have been correlated with the observed microstructures. The results of IC test showed that an increase in 0.4 pct W decreases the creep rate to nearly half the value. Creep strength of 1.4 pct W steel showed an increase in steels aged for short durations which decreased as aging time increased. The microstructural changes include coarsening of precipitates, reduction in dislocation density, changes in microchemistry, and formation of new phases. The formation of various phases and their volume fractions have been predicted using the JMatPro software for the two steels and validated by experimental methods. Detailed transmission electron microscopy analysis shows coarsening of precipitates and formation of a discontinuous network of Laves phase in 1.4 W steel aged for 10,000 hours at 823 K (550 °C) which is in agreement with the JMatPro simulation results.

  2. Effects of helium on ductile-brittle transition behavior of reduced-activation ferritic steels after high-concentration helium implantation at high temperature

    NASA Astrophysics Data System (ADS)

    Hasegawa, A.; Ejiri, M.; Nogami, S.; Ishiga, M.; Kasada, R.; Kimura, A.; Abe, K.; Jitsukawa, S.

    2009-04-01

    The effects of He on the fracture behavior of reduced-activation ferritic/martensitic steels, including oxide dispersion-strengthened (ODS) steels and F82H, was determined by characterizing the microstructural evolution in and fracture behavior of these steels after He implantation up to 1000 appm at around 550 °C. He implantation was carried out by a cyclotron with a beam of 50 MeV α-particles. In the case of F82H, the ductile-to-brittle transition temperature (DBTT) increase induced by He implantation was about 70 °C and the grain boundary fracture surface was only observed in the He-implanted area of all the ruptured specimens in brittle manner. By contrast, no DBTT shift or fracture mode change was observed in He-implanted 9Cr-ODS and 14Cr-ODS steels. Microstructural characterization suggested that the difference in the bubble formation behavior of F82H and ODS steels might be attributed to the grain boundary rupture of He-implanted F82H.

  3. Fabrication of 13Cr-2Mo Ferritic/Martensitic Oxide-Dispersion-Strengthened Steel Components by Mechanical Alloying and Spark-Plasma Sintering

    NASA Astrophysics Data System (ADS)

    Bogachev, I.; Grigoryev, E.; Khasanov, O. L.; Olevsky, E.

    2014-06-01

    The outcomes of the mechanical alloying of 13Cr-2Mo ferritic/martensitic steel and yttria (oxide-dispersion-strengthened steel) powders in a ball mill are reported in terms of the powder particle size and morphology evolution. The optimal ball mill rotation speed and the milling time are discussed. The densification kinetics of the mechanically alloyed powder during the process of spark-plasma sintering is analyzed. An optimal set of the compaction processing parameters, including the maximum temperature, the dwell time, and the heating rate, is determined. The specifics of the densification are discussed in terms of the impact of major spark-plasma sintering parameters as well as the possible phase transformations occurring during compaction processing.

  4. Saturation of the DBTT shift of irradiated 12Cr-1MoVW with increasing fluence

    SciTech Connect

    Vitek, J.M.; Corwin, W.R.; Klueh, R.L.; Hawthorne, J.R.

    1986-01-01

    Ferritic/martensitic steels are considered one of the prime candidate materials for structural applications in fusion reactors. This class of steel provides many advantages including excellent swelling resistance and high thermal conductivity. A reiew of current results on the suitability of ferritic/martensitic steels reveals their corrosion behavior and postirradiation properties are, in general, quite promising. However, a major concern is the existence in these steels of a ductile-to-brittle transition temperature (DBTT), below which the steel has limited ductility and fails in a brittle mode. Although the DBTT, as measured in a Charpy impact test, may be below room temperature in the unirradiated condition, irradiation can increase the DBTT significantly, and the increase depends on the irradiation temperature. Increases in the DBTT of greater than 100/sup 0/C have been measured following irradiation at 400/sup 0/C and below.

  5. Ferrite Formation Dynamics and Microstructure Due to Inclusion Engineering in Low-Alloy Steels by Ti2O3 and TiN Addition

    NASA Astrophysics Data System (ADS)

    Mu, Wangzhong; Shibata, Hiroyuki; Hedström, Peter; Jönsson, Pär Göran; Nakajima, Keiji

    2016-08-01

    The dynamics of intragranular ferrite (IGF) formation in inclusion engineered steels with either Ti2O3 or TiN addition were investigated using in situ high temperature confocal laser scanning microscopy. Furthermore, the chemical composition of the inclusions and the final microstructure after continuous cooling transformation was investigated using electron probe microanalysis and electron backscatter diffraction, respectively. It was found that there is a significant effect of the chemical composition of the inclusions, the cooling rate, and the prior austenite grain size on the phase fractions and the starting temperatures of IGF and grain boundary ferrite (GBF). The fraction of IGF is larger in the steel with Ti2O3 addition compared to the steel with TiN addition after the same thermal cycle has been imposed. The reason for this difference is the higher potency of the TiO x phase as nucleation sites for IGF formation compared to the TiN phase, which was supported by calculations using classical nucleation theory. The IGF fraction increases with increasing prior austenite grain size, while the fraction of IGF in both steels was the highest for the intermediate cooling rate of 70 °C/min, since competing phase transformations were avoided, the structure of the IGF was though refined with increasing cooling rate. Finally, regarding the starting temperatures of IGF and GBF, they decrease with increasing cooling rate and the starting temperature of GBF decreases with increasing grain size, while the starting temperature of IGF remains constant irrespective of grain size.

  6. Monitoring microstructural evolution in irradiated steel with second harmonic generation

    SciTech Connect

    Matlack, Kathryn H.; Kim, Jin-Yeon; Jacobs, Laurence J.; Wall, James J.; Qu, Jianmin

    2015-03-31

    Material damage in structural components is driven by microstructural evolution that occurs at low length scales and begins early in component life. In metals, these microstructural features are known to cause measurable changes in the acoustic nonlinearity parameter. Physically, the interaction of a monochromatic ultrasonic wave with microstructural features such as dislocations, precipitates, and vacancies, generates a second harmonic wave that is proportional to the acoustic nonlinearity parameter. These nonlinear ultrasonic techniques thus have the capability to evaluate initial material damage, particularly before crack initiation and propagation occur. This paper discusses how the nonlinear ultrasonic technique of second harmonic generation can be used as a nondestructive evaluation tool to monitor microstructural changes in steel, focusing on characterizing neutron radiation embrittlement in nuclear reactor pressure vessel steels. Current experimental evidence and analytical models linking microstructural evolution with changes in the acoustic nonlinearity parameter are summarized.

  7. Enabling Inexpensive Metallic Alloys as SOFC Interconnects: An Investigation into Hybrid Coating Technologies to Deposit Nanocomposite Functional Coatings on Ferritic Stainless Steel

    SciTech Connect

    Gannon, Paul; Gorokhovsky, Vladimir I.; Deibert, Max; Smith, Richard J.; Kayani, Asghar N.; White, P T.; Sofie, Stephen W.; Yang, Z Gary; Mccready, David E.; Visco, S.; Jacobson, C.; Kurokawa, H.

    2007-11-01

    Reduced operating temperatures (600-800°C) of Solid Oxide Fuel Cells (SOFCs) may enable the use of inexpensive ferritic steels as interconnects. Due to the demanding SOFC interconnect operating environment, protective coatings are required to increase long-term stability. In this study, large area filtered arc deposition (LAFAD) and hybrid filtered arc-assisted electron beam physical vapor deposition (FA-EBPVD) technologies were used to deposit two-segment coatings with Cr-Al-Y-O nanocomposite bottom segments and Mn-Co-O spinel-based top segments. Coatings were deposited on ferritic steels and subsequently annealed in air for various times. Surface oxidation was investigated using SEM/EDS, XRD and RBS analyses. Cr-volatilization was evaluated by transpiration and ICP-MS analysis of the resultant condensate. Time dependent Area Specific Resistance (ASR) was studied using the four-point technique. The oxidation behavior, Cr volatilization rate, and ASR of coated and uncoated samples are reported. Significant long-term (>1,000 hours) surface stability, low ASR, and dramatically reduced Cr-volatility were observed with the coated specimens. Improvement mechanisms, including the coating diffusion barrier properties and electrical conductivity are discussed.

  8. Charpy impact toughness of martensitic steels irradiated in FFTF: Effect of heat treatment

    SciTech Connect

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

    1992-12-31

    Plates of 9Cr-1MoVNb and 12Cr-1 MoVW steels were normalized and then tempered at two different tempering conditions. One-third-size Charpy specimens from each steel were irradiated to 7.4-8{times}10{sup 26} n/m{sup 2} (about {approximately}35 dpa) at 420{degrees}C in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility. Specimens were also thermally aged to 20,000 h at 400{degrees}C to compare the effect of aging and irradiation. Previous work on the steels irradiated to 4-5 dpa at 365{degrees}C in MOTA were reexamined in light of the new results. The tests indicated that prior-austenite grain size, which was varied by different normalizing treatments, had an effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize properties.

  9. Charpy impact toughness of martensitic steels irradiated in FFTF: Effect of heat treatment

    SciTech Connect

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

    1992-12-31

    Charpy tests were made on plates of 9Cr-1MoVNb and 12Cr-1MoVW steels given four different normalizing-and-tempering treatments. One-third-size Charpy specimens from each steel were irradiated to 7.4-8 {times} 10{sup 26} n/m{sup 2} (about 34--37 dpa) at 420C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4--5 dpa at 365C in MOTA were reexamined in light of the new results. The tests indicated that prior austenite grain size, which was varied by different normalizing treatments, had an effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. conclusions are presented on how heat treatment can be used to optimize properties.

  10. Charpy impact toughness of martensitic steels irradiated in FFTF: Effect of heat treatment

    SciTech Connect

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

    1992-01-01

    Charpy tests were made on plates of 9Cr-1MoVNb and 12Cr-1MoVW steels given four different normalizing-and-tempering treatments. One-third-size Charpy specimens from each steel were irradiated to 7.4-8 [times] 10[sup 26] n/m[sup 2] (about 34--37 dpa) at 420C in the Materials Open Test Assembly of the Fast Flux Test Facility. Specimens were also thermally aged to 20000 h at 400C to determine the effect of aging during irradiation. Previous work on these steels irradiated to 4--5 dpa at 365C in MOTA were reexamined in light of the new results. The tests indicated that prior austenite grain size, which was varied by different normalizing treatments, had an effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. conclusions are presented on how heat treatment can be used to optimize properties.

  11. Effect of γ-rays irradiation on the structural, magnetic, and electrical properties of Mg-Cu-Zn and Ni-Cu-Zn ferrites

    NASA Astrophysics Data System (ADS)

    Assar, S. T.; Abosheiasha, H. F.; El Sayed, A. R.

    2017-01-01

    Nanoparticles of Ni0.35Cu0.15Zn0.5Fe2O4 and Mg0.35Cu0.15Zn0.5Fe2O4, have been synthesized by citrate precursor method. Then some of the prepared samples have been irradiated by γ-rays of 60Co radioactive source at room temperature with doses of 1 Mrad and 2 Mrad, at a dose rate of 0.1 Mrad/h to study the effect of γ-rays irradiation on some structural, magnetic and electrical properties of the samples. The X-ray diffraction analysis (XRD), transmission electron microscopy, Fourier transform infrared spectroscopy and vibrating sample magnetometer measurements have been used to investigate the samples. The XRD results show that the irradiation has caused a decrease in the crystallite size and the measured density and an increase in the porosity, specific surface area, and microstrain in the case of Ni-Cu-Zn ferrite whereas in the case of Mg-Cu-Zn ferrite the reverse trend has been noticed. The lattice constant of the investigated samples has been increased with the increase of irradiation due to the conversion of Fe3+ (0.67 Å) to Fe2+ (0.76 Å). The magnetization results show an increase in saturation and remnant magnetizations for the two prepared ferrites after γ-rays irradiation. The main reason of this behavior is most probably due to the redistribution of the cations between A and B sites. The cation distribution has been proposed such that the values of theoretical and experimental magnetic moment are identical and increase as the magnetization increases. Moreover, a theoretical estimation of the lattice constant has been calculated on the basis of the proposed cation distribution for each sample and compared with the corresponding experimental values obtained by XRD analysis; where they have been found in a good agreement with each other. This can be considered as another confirmation of the validity of the cation distribution. Moreover, the cation distribution is thought to play an important role in increasing the values of dc conductivity of all samples

  12. Modifications of the magnetic properties of ferrites by swift heavy ion irradiations

    SciTech Connect

    Costantini, Jean-Marc; Studer, Francis; Peuzin, Jean-Claude

    2001-07-01

    Single crystal plates of ferrimagnetic yttrium iron garnet (111)-YIG:Si (Y{sub 3}Fe{sub 4.94}Si{sub 0.06}O{sub 12}) and barium hexaferrite (00.1)-BaM (BaFe{sub 12}O{sub 19}) or (00.1)-BaM:Co,Ti (BaFe{sub 9.1}Co{sub 1.4}Ti{sub 1.5}O{sub 19}) are irradiated with swift heavy ions (3.8 GeV {sup 129}Xe or 6.0 GeV {sup 208}Pb) in the electronic slowing down regime, above the threshold ({approximately}20 keV nm{minus}1) of formation of continuous and homogeneous cylindrical amorphous tracks. The modifications of the magnetic properties are studied by {sup 57}Fe Moessbauer spectroscopy and ac magnetic permeability measurements versus ion fluence. In the doped crystals having a planar magnetic anisotropy (YIG:Si and BaM:Co,Ti), the room-temperature Moessbauer spectra show that the magnetization is flipped perpendicularly to the sample plane at a critical amorphous fraction around 30% in both compounds. This corresponds to a 90% drop of the measured in-plane magnetic permeability. No such effect is seen in the undoped BaM samples with the axial [00.1] anisotropy. These data are interpreted by a magnetomechanical effect generated by the stress field induced by the amorphous tracks in the sample plane which flips the magnetization along the track-axis direction when the stress-induced anisotropy constant surpasses the pristine crystal anisotropy constant at the critical amorphous fraction. In the case of YIG:Si single crystal, a track-induced anisotropy field around 0.1 T is deduced from the Moessbauer spectra under a magnetic field applied in the sample (111) plane which rotates the magnetization back to the easy {l_angle}111{r_angle} magnetization axis lying near the sample (111) plane in a reversible manner. The magnetic ordering of amorphous YIG:Si below 70 K is also studied by Moessbauer spectroscopy under high magnetic field (5 T). A two-dimensional Bruggeman model used for the calculation of the permeability of the crystal+amorphous track composites yields track

  13. Microstructural characterization of irradiated Fe-Cu-Ni-P model steels

    SciTech Connect

    Miller, M.K.; Hoelzer, D.T.; Ebrahimi, F.; Hawthorne, J.R.; Burke, M.G.

    1987-01-01

    The microstructure of Fe-Cu-Ni-P model pressure vessel steels after neutron irradiation and thermal aging has been characterized by atom probe field-ion microscopy and augmented by transmission electron microscopy. High densities of small, roughly spherical or disc shaped copper clusters/precipitates were observed in the neutron irradiated alloys that contained copper. Small spherical phosphorus clusters were observed in the irradiated copper-free alloys, and copper phosphides were observed in a high phosphorus Fe-Cu-Ni-P alloy. None of these clusters/precipitates were observed in the thermally aged materials. The increases in the tensile and yield strengths that were observed after neutron irradiation resulted from these clusters and other lattice defects. 14 refs., 8 figs., 2 tabs.

  14. Nondestructive and Localized Measurements of Stress-Strain Curves and Fracture Toughness of Ferritic Steels at Various Temperatures Using Innovative Stress-Strain Microprobe Technology. Final Report for Period 8/13/1996--06/16/1999

    SciTech Connect

    Fahmy M. Haggag

    1999-10-29

    The results presented in this report demonstrate the capabilities of Advanced Technology Corporation's patented Portable/In Situ Stress-Strain Microprobe (TM) (SSM) System and its Automated Ball Indentation (ABI) test techniques to nondestructively measure the yield strength, the stress-strain curve, and the fracture toughness of ferritic steel samples and components in a reliable and accurate manner.

  15. Irradiation embrittlement of reactor pressure vessel steel at very high neutron fluence

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Debarberis, L.; von Estorff, U.; Gillemot, F.; Oszvald, F.

    2012-03-01

    For the prediction of radiation embrittlement of RPV materials beyond the NPP design time the analysis of research data and extended surveillance data up to a fluence ˜23 × 1020 cm-2 (E > 0.5 MeV) has been carried out. The experimental data used for the analysis are extracted from the International Database of RPV materials. Key irradiation embrittlement mechanisms, direct matrix damage, precipitation and element segregation have been considered. The essential part of the analysis concerns the assessment of irradiation embrittlement of WWER-440 steel irradiated with very high neutron fluence. The analysis of several surveillance sets irradiated at a fluence up to 23 × 1020 cm-2 (E > 0.5 MeV) has been performed. The effect of the main influencing chemical elements phosphorus and copper has been verified up to a fluence of 4.6 × 1020 cm-2 (E > 0.5 MeV). The data are indicating good radiation stability, in terms of the Charpy transition temperature shift and yield strength increase for steels with relatively low concentrations of copper and phosphorus. The linear dependence between ΔTk and ΔRp0.2 can be an evidence of strengthening mechanisms of irradiation embrittlement and absence of non-hardening embrittlement even at very high neutron fluence.

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

    NASA Astrophysics Data System (ADS)

    Margolin, B.; Minkin, A.; Smirnov, V.; Sorokin, A.; Shvetsova, V.; Potapova, V.

    2016-11-01

    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.

  17. Effect of 200 MeV Ag+15 ion irradiation on magnetic and dielectric properties of nanocrystalline Zn-Cr ferrite

    NASA Astrophysics Data System (ADS)

    Dolia, S. N.; Pareek, S. P.; Samariya, Arvind; Sharma, P. K.; Prasad, Arun S.; Dhawan, M. S.; Kumar, Sudhish; Sharma, K. B.; Asokan, K.

    2013-08-01

    Nanocrystalline samples of ZnCr0.4Fe1.6O4 ferrite were synthesized by the advanced sol-gel method to investigate the effect of 200 MeV Ag+15 ion irradiation on the cation distribution, magnetic and dielectric properties. Rietveld profile refinement of the X-ray diffraction (XRD) patterns confirms the single-phase cubic spinel structure of the specimens. The irradiated sample retains the cubic spinel structure with a slight increase in the lattice parameters and the average crystallite size. Temperature- and field-dependent dc magnetization studies show an appreciable enhancement in the saturation magnetization and blocking temperature of the irradiated samples, which could be attributed to the slight increase in the particle size due to the heat evolved during irradiation. Subsequently, the rearrangement of cations in the lattice structure and the ion-induced modifications on the surface states of the nanoparticles could be accountable. The room temperature dielectric constant and the loss tangent in the frequency range 75 kHz-10 MHz revealed the normal frequency dispersion. The increase in ɛr and tan δ on irradiation could be attributed to the slight crystal growth and hence the availability of the sufficient number of Fe2+ and/or Zn3+ ions particularly at the octahedral site on the grain boundaries, showing a fair agreement with the magnetization results.

  18. Grafting of HEMA onto dopamine coated stainless steel by 60Co-γ irradiation method

    NASA Astrophysics Data System (ADS)

    Jin, Wanqin; Yang, Liming; Yang, Wei; Chen, Bin; Chen, Jie

    2014-12-01

    A novel method for grafting of 2-hydroxyethyl methacrylate (HEMA) onto the surface of stainless steel (SS) was explored by using 60Co-γ irradiation. The surface of SS was modified by coating of dopamine before radiation grafting. The grafting reaction was performed in a simultaneous irradiation condition. The chemical structures change of the surface before and after grafting was demonstrated by Fourier transform infrared (FTIR) spectrometer. The hydrophilicity of the samples was determined by water contact angle measurement in the comparison of the stainless steel in the conditions of pristine, dopamine coated and HEMA grafted. Surface morphology of the samples was characterized by atomic force microscope (AFM) and scanning electron microscope (SEM). The corrosion resistance properties of the samples were evaluated by Tafel polarization curve. The hemocompatibility of the samples were tested by platelet adhesion assay.

  19. Properties of copper?stainless steel HIP joints before and after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Tähtinen, S.; Laukkanen, A.; Singh, B. N.; Toft, P.

    2002-12-01

    The tensile and fracture behaviour of CuCrZr and CuAl25 IG0 alloys joint to 316L(N) stainless steel by hot isostatic pressing (HIP) have been determined in unirradiated and neutron-irradiated conditions. The tensile and fracture behaviour of copper alloy HIP joint specimens are dominated by the properties of the copper alloys, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloys and stainless steel. The test temperature, neutron irradiation and thermal cycles primarily affect the copper alloy HIP joint properties through changing the strength mismatch between the base alloys. Changes in the loading conditions i.e. tensile, bend ( JI) and mixed-mode bend ( JI/ JII) lead to different fracture modes in the copper alloy HIP joint specimens.

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

    SciTech Connect

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

    1996-10-01

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