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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. Irradiation effects in ferritic steels

    NASA Astrophysics Data System (ADS)

    Lechtenberg, Thomas

    1985-08-01

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

  3. Irradiation embrittlement of neutron-irradiated low activation ferritic steels

    NASA Astrophysics Data System (ADS)

    Kayano, H.; Kimura, A.; Narui, M.; Sasaki, Y.; Suzuki, Y.; Ohta, S.

    1988-07-01

    Effects of neutron irradiation and additions of small amounts of alloying elements on the ductile-brittle transition temperature (DBTT) of three different groups of ferritic steels were investigated by means of the Charpy impact test in order to gain an insight into the development of low-activation ferritic steels suitable for the nuclear fusion reactor. The groups of ferritic steels used in this study were (1) basic 0-5% Cr ferritic steels, (2) low-activation ferritic steels which are FeCrW steels with additions of small amounts of V, Mn, Ta, Ti, Zr, etc. and (3) FeCrMo, Nb or V ferritic steels for comparison. In Fe-0-15% Cr and FeCrMo steels, Fe-3-9% Cr steels showed minimum brittleness and provided good resistance against irradiation embrittlement. Investigations on the effects of additions of trace amounts of alloying elements on the fracture toughness of low-activation ferritic steels made clear the optimum amounts of each alloying element to obtain higher toughness and revealed that the 9Cr-2W-Ta-Ti-B ferritic steel showed the highest toughness. This may result from the refinement of crystal grains and improvement of quenching characteristics caused by the complex effect of Ti and B.

  4. 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.

  5. Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1990-05-01

    Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development.

  6. Irradiation-induced precipitation modelling of ferritic steels

    NASA Astrophysics Data System (ADS)

    Yin, You Fa; Faulkner, Roy G.; Lu, Zheng

    2009-06-01

    In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.

  7. Irradiation creep of low-activation ferritic steels in FFTF/MOTA*1

    NASA Astrophysics Data System (ADS)

    Kohyama, A.; Kohno, Y.; Asakura, K.; Yoshino, M.; Namba, C.; Eiholzer, C. R.

    1994-09-01

    Irradiation creep behavior of low-activation steels, developed as structural materials for fusion reactors (JLF series steels), was investigated to obtain a fundamental understanding of these alloys under fast neutron irradiation in FFTF. (2.25-8)Cr(1-2)W bainitic steels and 12Cr-2W ferritic steels showed superior creep resistance to type-316 stainless steels under fast neutron irradiation up to 520°C. At temperatures below 460°C the creep strain increased with increasing Cr content up to 7 Cr, and further increments of Cr content up to 12% reduced the creep strain. At temperatures between 460 and 600°C, 7-8 Cr ferritic steels showed the largest creep strain. Swelling-enhanced creep, near the peak swelling temperature of 410°C, was also observed. The 9Cr-2W ferritic steel JLF-1 presented excellent properties, suggesting it as a leading candidate alloy for structural components of fusion reactors.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

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

    DOE PAGESBeta

    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 × 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

  13. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2015-11-01

    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 spinodal 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. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Microstructural development under irradiation in European ODS ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Schäublin, R.; Ramar, A.; Baluc, N.; de Castro, V.; Monge, M. A.; Leguey, T.; Schmid, N.; Bonjour, C.

    2006-06-01

    Oxide dispersion strengthened steels based on the ferritic/martensitic steel EUROFER97 are promising candidates for a fusion reactor because of their improved high temperature mechanical properties and their potential higher radiation resistance relative to the base material. Several EUROFER97 based ODS F/M steels are investigated in this study. There are the Plansee ODS steels containing 0.3 wt% yttria, and the CRPP ODS steels, whose production route is described in detail. The reinforcing particles represent 0.3-0.5% weight and are composed of yttria. The effect of 0.3 wt% Ti addition is studied. ODS steel samples have been irradiated with 590 MeV protons to 0.3 and 1.0 dpa at room temperature and 350 °C. Microstructure is investigated by transmission electron microscopy and mechanical properties are assessed by tensile and Charpy tests. While the Plansee ODS presents a ferritic structure, the CRPP ODS material presents a tempered martensitic microstructure and a uniform distribution of the yttria particles. Both materials provide a yield stress higher than the base material, but with reduced elongation and brittle behaviour. Ti additions improve elongation at high temperatures. After irradiation, mechanical properties of the material are only slightly altered with an increase in the yield strength, but without significant decrease in the total elongation, relative to the base material. Samples irradiated at room temperature present radiation induced defects in the form of blacks dots with a size range from 2 to 3 nm, while after irradiation at 350 °C irradiation induced a0<1 0 0>{1 0 0} dislocation loops are clearly visible along with nanocavities. The dispersed yttria particles with an average size of 6-8 nm are found to be stable for all irradiation conditions. The density of the defects and the dispersoid are measured and found to be about 2.3 × 10 22 m -3 and 6.2 × 10 22 m -3, respectively. The weak impact of irradiation on mechanical properties of ODS F

  1. 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.

  2. Effect of boron on post irradiation tensile properties of reduced activation ferritic steel (F-82H) irradiated in HFIR

    SciTech Connect

    Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi; Pawel, J.E.

    1994-12-31

    Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.

  3. 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.

  4. 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.

  5. 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.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    SciTech Connect

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

    1996-04-01

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

  8. 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.

  9. Microstructure of HFIR-irradiated 12-Cr 1 MoVW ferritic steel

    SciTech Connect

    Vitek, J.M.; Klueh, R.L.

    1983-01-01

    As part of the fusion materials development program in the United States, a 12 Cr-1 MoVW ferritic steel was irradiated in the High Flux Isotope Reactor (HFIR) to a damage level of 36 dpa at 300, 400, 500, and 600/sup 0/C. During irradiation in HFIR, a transmutation reaction of nickel results in the production of helium, to a level of 99 at. ppM in the present experiment. The microstructures were evaluated after irradiation and the results are presented. Cavities were found at all temperatures. Small cavities (3 to 9 nm) were observed after irradiation at 300, 500 and 600/sup 0/C. At 500 and 600/sup 0/C, the cavities were found preferentially at dislocations, lath boundaries, and prior austenite grain boundaries. After irradiation at 400/sup 0/C, larger cavities (4 to 30 nm) were observed homogeneously distributed throughout the tempered martensite structure. The maximum swelling was 0.07% after irradiation at 400/sup 0/C. Comparision of the results with other studies in which helium was not present at such high levels indicated helium enhances the swelling of 12 Cr-1 MoVW.

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

    NASA Astrophysics Data System (ADS)

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

    1994-09-01

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

  11. Stability of the strengthening nanoprecipitates in reduced activation ferritic steels under Fe2+ ion irradiation

    NASA Astrophysics Data System (ADS)

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

    2014-02-01

    The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic-martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ∼49 dpa at 500 °C using Fe2+ is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe2W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. THE EFFECTS OF FAST REACTOR IRRADIATION CONDITIONS ON THE TENSILE PROPERTIES OF TWO FERRITIC/MARTENSITIC STEELS

    SciTech Connect

    Maloy, Stuart A.; Toloczko, Mychailo B.; McClellan, K. J.; Romero, T. J.; Kohno, Y.; Garner, Francis A.; Kurtz, Richard J.; Kimura, Akihiko

    2006-09-15

    Tensile testing has been performed at 25 and at ~400°C on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to ~70 dpa at 373 to 433°C. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The percent increase in yield stress decreases with increasing irradiation temperature from 373 to 433ºC. The JFMS alloy, which has 0.7 wt. % silicon, exhibits approximately a factor of two increase in yield strength between tests at 427°C and at 373°C, and shows an increase in hardening with increasing dose. A comparison of the JFMS tensile properties to the properties of other ferritic/martensitic steels suggests that this hardening is due to precipitation of a Si-rich Laves phase in this alloy. The HT-9 alloy, which contains more chromium and more carbon but less silicon (0.2 wt. %), less molybdenum and less nickel, hardens during irradiation at 373°C, but shows less hardening for irradiations performed at 427ºC and no increase in yield stress with increasing dose beyond 10 dpa.

  20. The effects of fast reactor irradiation conditions on the tensile properties of two ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Maloy, Stuart A.; Toloczko, M. B.; McClellan, K. J.; Romero, T.; Kohno, Y.; Garner, F. A.; Kurtz, R. J.; Kimura, A.

    2006-09-01

    Tensile testing has been performed at 25 and at ˜400 °C on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to ˜70 dpa at 373-433 °C. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The percent increase in yield stress decreases with increasing irradiation temperature from 373 to 433 °C. The JFMS alloy, which has 0.7 wt% silicon, exhibits approximately a factor of two increase in yield strength between tests at 427 and at 373 °C, and shows an increase in hardening with increasing dose. A comparison of the JFMS tensile properties to the properties of other ferritic/martensitic steels suggests that this hardening is due to precipitation of a Si-rich laves phase in this alloy. The HT-9 alloy, which contains more chromium and more carbon but less silicon (0.2 wt%), less molybdenum and less nickel, hardens during irradiation at 373 °C, but shows less hardening for irradiations performed at 427 °C and no increase in yield stress with increasing dose beyond 10 dpa.

  1. 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.

  2. IRRADIATION CREEP AND SWELLING OF RUSSIAN FERRITIC-MARTENSITIC STEELS IRRADIATED TO VERY HIGH EXPOSURES IN THE BN-350 FAST REACTOR AT 305-335 DEGREES C

    SciTech Connect

    Konobeev, Yu V.; Dvoraishin, A. M.; Porollo, S. I.; Shulepin, S. V.; Budylkin, N. I.; Mironova, E. G.; Garner, Francis A.; Toloczko, Mychailo B.

    2003-09-03

    Russian ferritic martensitic (F(slash)M) steels EP(dash)450, EP(dash)852 and EP(dash)823 were irradiated in the BN(dash)350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb(dash)Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP(dash)450 and EP(dash) 823 at temperatures between 390 and 520 degrees C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP(dash)450 and EP(dash)852 at temperatures between 305 and 335 degrees C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, 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. These irradiation creep studies confirm that the creep compliance of F(slash)M steels is about one half that of austenitic steels.

  3. Low cycle fatigue properties of reduced activation ferritic/martensitic steels after high-dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Gaganidze, E.; Petersen, C.; Aktaa, J.; Povstyanko, A.; Prokhorov, V.; Diegele, E.; Lässer, R.

    2011-08-01

    This paper focuses on the low cycle fatigue (LCF) behaviour of reduced activation ferritic/martensitic steels irradiated to a displacement damage dose of up to 70 dpa at 330-337 °C in the BOR 60 reactor within the ARBOR 2 irradiation programme. The influence of neutron irradiation on the fatigue behaviour was determined for the as-received EUROFER97, pre-irradiation heat-treated EUROFER97 HT and F82H-mod steels. Strain-controlled push-pull loading was performed using miniaturized cylindrical specimens at a constant temperature of 330 °C with total strain ranges between 0.8% and 1.1%. Comparison of the LCF behaviour of irradiated and reference unirradiated specimens was performed for both the adequate total and inelastic strains. Neutron irradiation-induced hardening may have various effects on the fatigue behaviour of the steels. The reduction of inelastic strain in the irradiated state compared with the reference unirradiated state at common total strain amplitudes may increase fatigue lifetime. The increase in the stress at the adequate inelastic strain, by contrast, may accelerate fatigue damage accumulation. Depending on which of the two effects mentioned dominates, neutron irradiation may either extend or reduce the fatigue lifetime compared with the reference unirradiated state. The results obtained for EUROFER97 and EUROFER97 HT confirm these considerations. Most of the irradiated specimens show fatigue lifetimes comparable to those of the reference unirradiated state at adequate inelastic strains. Some irradiated specimens, however, show lifetime reduction or increase in comparison with the reference state at adequate inelastic strains.

  4. 55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

    SciTech Connect

    Liu, Haibo; Abdou, Mohamed A.; Greenwood, Lawrence R.

    2013-11-01

    How to increase the ferritic steel He(appm)/dpa ratio in a fission reactor neutron spectrum is an important question for fusion reactor material testing. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched 54Fe in iron with 458.2 effective full power days (EFPD) irradiation in the High Flux Isotope Reactor (HFIR), ORNL. Greenwood suggested that the transmutation produced 55Fe has a thermal neutron helium production cross section which may have an effect on this result. In the current work, the ferritic steel He(appm)/dpa ratio is studied in the neutron spectrum of HFIR with 55Fe thermal neutron helium production taken into account. The available ENDF-b format 55Fe incident neutron cross section file from TENDL, Netherlands, is first input into the calculation model. A benchmark calculation for the same sample as used in the aforementioned experiment was used to adjust and evaluate the TENDL 55Fe (n, a) cross section values. The analysis shows a decrease of a factor of 6700 for the TENDL 55Fe (n, a) cross section in the intermediate and low energy regions is required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy is about 27 mb. With the adjusted 55Fe (n, a) cross sections, calculation show that the 54Fe and 55Fe isotopes can be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. The results show that a 70% enriched 54Fe and 30% enriched 55Fe ferritic steel sample would produce a He(appm)/dpa ratio of about 13 initially in the HFIR peripheral target position (PTP). After one year irradiation, the ratio decreases to about 10. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the 55Fe (n, a) cross section

  5. Influence of Ar-ions irradiation on the oxidation behavior of ferritic-martensitic steel P92 in supercritical water

    NASA Astrophysics Data System (ADS)

    Huang, Xi; Shen, Yinzhong; Zhu, Jun

    2015-02-01

    The corrosion behavior of ferritic-marensitic steel P92 with and without Ar-ions irradiation in supercritical water at 823 K(550 °C)/25 MPa for different exposure times was investigated by a variety of characterization techniques. A distinct difference in oxidation morphology between irradiated and unirradiated samples was observed. The oxide morphology of samples with a relatively moderate radiation intensity was similar with that of samples without irradiation. Many small oxide particles were observed in the region with a relatively high radiation intensity but their size was increased gradually with increasing exposure times. Exfoliation of oxide layer occurred for irradiated samples exposed for 100 h. Chromium-rich oxide layer with a chromium content of more than 20 wt pct along with a small-scale three-layer oxide structures were observed in Ar-ions irradiated samples, arising from the microstructural change in steel samples after the irradiation. Mechanism for the exfoliation of oxide layer is also discussed.

  6. Shear Punch Properties of Low Activation Ferritic Steels Following Irradiation in ORR

    SciTech Connect

    Ermi, Ruby M.; Hamilton, Margaret L.; Gelles, David S.; Ermi, August M.

    2001-10-01

    Shear punch post-irradiation test results are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 degrees centigrade to {approx}10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5% and W to 1.0%. Comparison of results with tensile test results showed good correlations with previously observed trends except where disks were improperly manufactured because they were too thin or because engraving was faulty.

  7. Characterization of neutron-irradiated ferritic model alloys and a RPV steel from combined APT, SANS, TEM and PAS analyses

    NASA Astrophysics Data System (ADS)

    Meslin, E.; Lambrecht, M.; Hernández-Mayoral, M.; Bergner, F.; Malerba, L.; Pareige, P.; Radiguet, B.; Barbu, A.; Gómez-Briceño, D.; Ulbricht, A.; Almazouzi, A.

    2010-11-01

    Understanding the behavior of reactor pressure vessel (RPV) steels under irradiation is a mandatory task that has to be elucidated in order to be able to operate safely a nuclear power plant or to extend its lifetime. To build up predictive tools, a substantial experimental data base is needed at the nanometre scale to extract quantitative information on neutron-irradiated materials and to validate the theoretical models. To reach this experimental goal, ferritic model alloys and French RPV steel were neutron irradiated in a test reactor at an irradiation flux of 9 × 10 17 nm -2 s, doses from 0.18 to 1.3 × 10 24 nm -2 and 300 °C. The main goal of this paper is to report the characterization of the radiation-induced microstructural change in the materials by using the state-of-the-art of characterization techniques available in Europe at the nanometre scale. Possibilities, limitations and complementarities of the techniques to each other are highlighted.

  8. Microstructural evolution of P92 ferritic/martensitic steel under Ar{sup +} ion irradiation at elevated temperature

    SciTech Connect

    Jin Shuoxue; Guo Liping; Li Tiecheng; Chen Jihong; Yang Zheng; Luo Fengfeng; Tang Rui; Qiao Yanxin; Liu Feihua

    2012-06-15

    Irradiation damage in P92 ferritic/martensitic steel irradiated by Ar{sup +} ion beams to 7 and 12 dpa at elevated temperatures of 290 Degree-Sign C, 390 Degree-Sign C and 550 Degree-Sign C has been investigated by transmission electron microscopy, scanning electron microscopy and atomic force microscopy. The precipitate periphery (the matrix/carbide interface) was amorphized only at 290 Degree-Sign C, while higher irradiation temperature could prevent the amorphization. The formation of the small re-precipitates occurred at 290 Degree-Sign C after irradiation to 12 dpa. With the increase of irradiation temperature and dose, the phenomenon of re-precipitation became more severe. The voids induced by irradiation were observed after irradiation to 7 dpa at 550 Degree-Sign C, showing that high irradiation temperature ({>=} 550 Degree-Sign C) was a crucial factor which promoted the irradiation swelling. Energy dispersive X-ray analysis revealed that segregation of Cr and W in the voids occurred under irradiation, which may influence mechanical properties of P92 F/M steel. - Graphical Abstract: High density of small voids, about 2.5 nm in diameter, was observed after irradiation to 12 dpa at 550 Degree-Sign C, which was shown in panel a (TEM micrograph). As shown in panel b (SEM image), a large number of nanometer-sized hillocks were formed in the surface irradiated at 550 Degree-Sign C, and the mean size was {approx} 30 nm. The formation of the nanometer-sized hillocks might be due to the voids that appeared as shown in TEM images (panel a). High irradiation temperature ({>=} 550 Degree-Sign C) was a crucial factor for the formation of void swelling. Highlights: Black-Right-Pointing-Pointer Small carbides re-precipitated in P92 matrix irradiated to 12 dpa at 290 Degree-Sign C. Black-Right-Pointing-Pointer High density of voids was observed at 550 Degree-Sign C. Black-Right-Pointing-Pointer Segregation of Cr and W in voids occurred under irradiation.

  9. Articles comprising ferritic stainless steels

    DOEpatents

    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.

  10. Helium behavior in ferritic/martensitic steels irradiated in spallation target

    NASA Astrophysics Data System (ADS)

    Krsjak, Vladimir; Kuriplach, Jan; Shen, Tielong; Sabelova, Veronika; Sato, Koichi; Dai, Yong

    2015-01-01

    Two positron annihilation spectroscopy (PAS) techniques have been used for the investigation of helium behavior in STIP samples. Positron lifetime measurements and coincidence Doppler broadening spectroscopy have been employed together in a complex PAS characterization of RAFM steel irradiated in a mixed neutron-proton spectrum up to 20 dpa and 1800 appm He. Both techniques show an increase of the He-to-dpa ratio up to ∼10 dpa. At higher irradiation loads, the ratio is decreasing, which was attributed to the formation and growth of helium bubbles.

  11. Deformation Microstructure of a Reduced-Activation Ferritic/Martensitic Steel Irradiated in HFIR

    SciTech Connect

    Hashimoto, N.; Klueh, R.L.; Ando, M.; Tanigawa, H.; Sawai, T.; Shiba, K.

    2003-09-15

    In order to determine the contributions of different microstructural features to strength and to deformation mode, microstructure of deformed flat tensile specimens of irradiated reduced activation F82H (IEA heat) base metal (BM) and its tungsten inert-gas (TIG) weldments (weld metal and weld joint) were investigated by transmission electron microscopy (TEM), following fracture surface examination by scanning electron microscopy (SEM). After irradiation, the fracture surfaces of F82H BM and TIG weldment showed a martensitic mixed quasi-cleavage and ductile-dimple fracture. The microstructure of the deformed region of irradiated F82H BM contained dislocation channels. This suggests that dislocation channeling could be the dominant deformation mechanism in this steel, resulting in the loss of strain-hardening capacity. While, the necked region of the irradiated F82H TIG, where showed less hardening than F82H BM, showed deformation bands only. From these results, it is suggested that the pre-irradiation microstructure, especially the dislocation density, could affect the post-irradiation deformation mode.

  12. Fatigue behavior of irradiated helium-containing ferritic steels for fusion reactor applications*1

    NASA Astrophysics Data System (ADS)

    Grossbeck, M. L.; Vitek, J. M.; Liu, K. C.

    1986-11-01

    The martensitic alloys 12Cr-1MoVW and 9Cr-1MoVNb have been irradiated in the High Flux Isotope Reactor (HFIR) and subsequently tested in fatigue. In order to achieve helium levels characteristic of fusion reactors, the 12Cr-1MoVW was doped with 1 and 2% Ni, resulting in helium levels of 210 and 410 appm at damage levels of 25 dpa. The 9Cr-1MoVNb was irradiated to a damage level of 3 dpa and contained less than 5 appm He. Irradiations were carried out at 55°C and testing at 22°C. No significant changes were found in 9Cr-1MoVNb upon irradiation at this damage level, but effects that could possibly be attributed to helium were found in 12Cr-1MoVW. Levels of 210 and 410 appm He produced cyclic strengthening of 29 and 34% over unirradiated nickel-doped materials, respectively. This cyclic hardening, attributable largely to helium, resulted in degradation of the cyclic life. However, the fatigue life remained comparable to or better than unirradiated 20%-cold-worked 316 stainless steel.

  13. Mechanical properties and microstructure of three Russian ferritic/martensitic steels irradiated in BN-350 reactor to 50 dpa at 490C

    SciTech Connect

    Dvoriashin, Alexander M.; Porollo, S. I.; Konobeev, Yu V.; Budylkin, N. I.; Mironova, E. G.; Ioltukhovsky, A. G.; Leonteva-Smirnova, M. V.; Garner, Francis A.

    2007-08-01

    Ferritic/martensitic (F/M) steels are being considered for application in fusion reactors, intense neutron sources, and accelerator-driven systems. While EP-450 is traditionally used with sodium coolants in Russia, EP-823 and EI-852 steels with higher silicon levels have been developed for reactor facilities using lead-bismuth coolant. To determine the influence of silicon additions on short-term mechanical properties and microstructure, ring specimens cut from cladding tubes of these three steels were irradiated in sodium at 490С in the BN-350 reactor to 50 dpa. Post-irradiation tensile testing and microstructural examination show that EI-852 steel (1.9 wt% Si) undergoes severe irradiation embrittlement. Microstructural investigation showed that the formation of near-continuous -phase precipitates on grain boundaries is the main cause of the embrittlement.

  14. 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.

  15. Defect and void evolution in oxide dispersion strengthened ferritic steels under 3.2 MeV Fe + ion irradiation with simultaneous helium injection

    NASA Astrophysics Data System (ADS)

    Kim, I.-S.; Hunn, J. D.; Hashimoto, N.; Larson^1, D. L.; Maziasz, P. J.; Miyahara, K.; Lee, E. H.

    2000-08-01

    In an attempt to explore the potential of oxide dispersion strengthened (ODS) ferritic steels for fission and fusion structural materials applications, a set of ODS steels with varying oxide particle dispersion were irradiated at 650°C, using 3.2 MeV Fe + and 330 keV He + ions simultaneously. The void formation mechanisms in these ODS steels were studied by juxtaposing the response of a 9Cr-2WVTa ferritic/martensitic steel and solution annealed AISI 316LN austenitic stainless steel under the same irradiation conditions. The results showed that void formation was suppressed progressively by introducing and retaining a higher dislocation density and finer precipitate particles. Theoretical analyses suggest that the delayed onset of void formation in ODS steels stems from the enhanced point defect recombination in the high density dislocation microstructure, lower dislocation bias due to oxide particle pinning, and a very fine dispersion of helium bubbles caused by trapping helium atoms at the particle-matrix interfaces.

  16. Relationship Between Grain Boundary Structure and Radiation Induced Segregation in a Neutron Irradiated 9 wt. % Cr Model Ferritic/Martensitic Steel

    SciTech Connect

    Field, Kevin G; Miller, Brandon; Chichester, Heather J.M.; Sridharan, K.; Allen, Todd R.

    2014-01-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs but has only been demonstrated in ion irradiated specimens. A 9 wt. % Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of neutron radiation environment on the RIS-GB structure dependence. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  17. Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9 wt.% Cr model ferritic/martensitic steel

    SciTech Connect

    Field, Kevin G.; Miller, Brandon D.; Chichester, Heather J. M.; Sridharan, Kumar; Allen, Todd R.

    2014-02-01

    Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

  18. In situ proton irradiation creep of ferritic-martensitic steel T91

    NASA Astrophysics Data System (ADS)

    Xu, Cheng; Was, Gary S.

    2013-10-01

    An irradiation creep apparatus was developed for in situ straining of T91 strip samples while exposed to 2-3 MeV proton irradiation at 300-600 °C. Thermal creep experiments were conducted at 600 °C, 47 MPa, and 500 °C, 160 MPa. The thermal creep strains were in reasonable agreement with literature data on bulk samples of T91. An irradiation creep experiment was conducted at 500 °C and 160 MPa with a damage rate range from 3.1 × 10-6 dpa/s to 4.9 × 10-6 dpa/s. The creep rate of T91 was found to increase linearly with dose rate. A TEM investigation of the irradiated microstructure showed signs of dislocation pileup, subgrain formation, and small dislocation loops. The results illustrate the utility of accelerator-creep experiments to obtain creep rates at low dose and the capability to observe transient changes in real time, thus providing the tools for isolating the effects of individual variables on creep rate of T91.

  19. Kinetic of solute clustering in neutron irradiated ferritic model alloys and a French pressure vessel steel investigated by atom probe tomography

    NASA Astrophysics Data System (ADS)

    Meslin, E.; Radiguet, B.; Pareige, P.; Barbu, A.

    2010-04-01

    The embrittlement of reactor pressure vessel steels under neutron irradiation is partly due to the formation of solute clusters. To gain more insight into their formation mechanisms, ferritic model alloys (low copper Fe-0.08 at.% Cu, Fe-0.09 Cu-1.1 Mn-0.7 Ni (at.%), and a copper free Fe-1.1 Mn-0.7 Ni (at.%)) and a French 16MND5 reactor pressure vessel steel, were irradiated in a test reactor at two fluxes of 0.15 and 9 × 10 17 n( E> 1 MeV) m -2 s -1 and at increasing doses from 0.18 to 1.3 × 10 24 n( E> 1 MeV) m -2. Atom probe tomography analyses revealed that nanometer-size solute clusters were formed during irradiation in all the materials, even in the copper free Fe-1.1 Mn-0.7 Ni (at.%) alloy. It should be noted that solute segregation in a low-Ni ferritic material was never reported before in absence of the highly insoluble copper impurity. The manganese and nickel segregation is suggested to result from a radiation-induced mechanism.

  20. 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.

  1. 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.

  2. 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.

  3. Ferritic/martensitic steels - overview of recent results

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Gelles, D. S.; Jitsukawa, S.; Kimura, A.; Odette, G. R.; van der Schaaf, B.; Victoria, M.

    2002-12-01

    Considerable research work has been conducted on the ferritic/martensitic steels since the last International Conference on Fusion Reactor Materials in 1999. Since only a limited amount of that work can be reviewed in this paper, four areas will be emphasized: (1) the international collaboration under the auspices of the International Energy Agency (IEA) to address potential problems with ferritic/martensitic steels and to prove their feasibility for fusion, (2) the major uncertainty that remains concerning the effect of transmutation helium on mechanical properties of the steels when irradiated in a fusion neutron environment, (3) development of new reduced-activation steels beyond the F82H and JLF-1 steels studied in the IEA collaboration, and (4) work directed at developing oxide dispersion-strengthened steels for operation above 650 °C.

  4. 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.

  5. 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.

  6. Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

    NASA Astrophysics Data System (ADS)

    T, Hino; K, Yamaguchi; Y, Yamauchi; Y, Hirohata; K, Tsuzuki; Y, Kusama

    2005-04-01

    Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study, deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus. After a ferritic steel sample was irradiated by 1.7 keV D+ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fluence dependence of retained amount of deuterium was measured by changing the ion fluence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention. Ferritic steel sample was

  7. 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.

  8. A dual ion irradiation study of helium-dpa interactions on cavity evolution in tempered martensitic steels and nanostructured ferritic alloys

    NASA Astrophysics Data System (ADS)

    Yamamoto, Takuya; Wu, Yuan; Robert Odette, G.; Yabuuchi, Kiyohiro; Kondo, Sosuke; Kimura, Akihiko

    2014-06-01

    Cavity evolutions in a normalized and tempered martensitic steel (TMS) and two nanostructured ferritic alloys (NFA) under Fe3+ and He+ dual ion beam irradiations (DII) at 500 °C and 650 °C were characterized. The irradiation conditions encompass a wide range of displacement per atom damage (dpa), He and He/dpa. The 500 °C DII produced damage and He levels of ≈10-47 dpa and ≈400-2000 appm, respectively. Transmission electron microscopy (TEM) showed that DII of a 8Cr TMS, at 500 °C to up to 60 dpa and 2100 appm He, produced a moderate density of non-uniformly distributed cavities with bimodal sizes ranging from ≈1 nm He bubbles to ≈20 nm faceted voids, and swelling ≈0.44%. In contrast, the same irradiation conditions produced only small ≈1.3 nm diameter bubbles and swelling of ≈0.05% in the NFA MA957. Similar bubble distributions were observed in MA957 and a developmental NFA DII at 650 °C up to ≈80 dpa and ≈3900 appm He. These results demonstrate the outstanding He management capability of the oxide nano-features in the NFA. The various data trends are shown as a function of dpa, He, He/dpa and He*dpa.

  9. Development of ferritic steels for fusion reactor applications

    SciTech Connect

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

    1988-08-01

    Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs.

  10. MECHANICAL PROPERTIES AND MICROSTRUCTURE OF THREE RUSSIAN Mechanical Properties And Microstructure Of Three Russian Ferritic/Martensitic Steels Irradiated In BN-350 Reactor To 50 dpa at 490C

    SciTech Connect

    Dvoriashin, Alexander M; Porollo, S I; Konobeev, Yu V; Budylkin, N I; Minonova, E G; Loltukhovsky, A G; Leonteva-Smirnova, M V; Bochvar, A A; Garner, Francis A

    2007-03-01

    Ferritic/martensitic (F/M) steels are being considered for application in fusion reactors, intense neutron sources, and accelerator-driven systems. While EP-450 is traditionally used with sodium coolants in Russia, EP-823 and EI-852 steels with higher silicon levels have been developed for reactor facilities using lead-bismuth coolant. To determine the influence of silicon additions on short-term mechanical properties and microstructure, ring specimens cut from cladding tubes of these three steels were irradiated in sodium at 490°С in the BN-350 reactor to 50 dpa. Post-irradiation tensile testing and microstructural examination show that EI-852 steel (1.9 wt% Si) undergoes severe irradiation embrittlement. Microstructural investigation showed that the formation of near-continuous phase precipitates on grain boundaries is the main cause of the embrittlement.

  11. Nanoindentation on ion irradiated steels

    NASA Astrophysics Data System (ADS)

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

    2009-06-01

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

  12. Abnormal ferrite in hyper-eutectoid steels

    SciTech Connect

    Chairuangsri, T.; Edmonds, D.V.

    2000-04-19

    The microstructural characteristics of ultra-high carbon hyper-eutectoid Fe-C and Fe-C-Cu experimental steels have been examined after isothermal transformation in a range just beneath the eutectoid temperature. Particular attention was paid to the formation of so-called abnormal ferrite, which refers to coarse ferrite grains which can form, in hyper-eutectoid compositions, on the pro-eutectoid cementite before the pearlite reaction occurs. Thus it is confirmed that the abnormal ferrite is not a result of pearlite coarsening, but of austenite decomposition before the conditions for coupled growth of pearlite are established. The abnormal ferrite formed on both allotriomorphic and Widmanstaetten forms of pro-eutectoid cementite, and significantly, it was observed that the pro-eutectoid cementite continued to grow, despite being enclosed by the abnormal ferrite. Under certain conditions this could lead to the eventual formation of substantially reduced amounts of pearlite. Thus, a model for carbon redistribution that allows the proeutectoid cementite to thicken concurrently with the abnormal ferrite is presented. The orientation relationships between the abnormal ferrite and pro-eutectoid cementite were also determined and found to be close to those which have been reported between pearlitic ferrite and pearlitic cementite.

  13. 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.

  14. Ferrite morphology and variations in ferrite content in austenitic stainless steel welds

    SciTech Connect

    David, S.A.; Hanzelka, S.E.; Haltom, C.P.

    1981-07-01

    Four distinct ferrite morphologies have been identified in type 308 stainless steel multipass welds: vermicular, lacy, acicular, and globular. The first three ferrite types are related to transformations following solidification and the fourth is related to the shape instability of the residual ferrite. An earlier study showed that most of the ferrite observed in austenitic stainless steel welds contaning a duplex structure may be identified as residual primary ferrite resulting from incomplete delta ..-->.. ..gamma.. transformation during solidification and/or residual ferrite after Widmanstaetten austenite precipitation in primary ferrite. These modes of ferrite formation can be used to explain observed ferrite morphologies in austenitic stainless steel welds. Variations in ferrite content within the weld were related to weld metal composition, ferrite morphology, and dissolution of ferrite resulting from thermal cycles during subsequent weld passes. An investigation of the type 308 stainless steel filler metal solidified over cooling rates ranging from 7 to 1600/sup 0/C/s showed that the cooling rate of the weld metal within the freezing range of the alloy affects the amount of ferrite in the microstructure very litte. However, the scale of the solidification substructure associated with various solidification rates may influence the ferrite dissolution kinetics.

  15. Ferrite morphology and variations in ferrite content in austenitic stainless steel welds

    SciTech Connect

    David, S.A.

    1981-04-01

    Four distinct ferrite morphologies have been identified in Type 308 stainless steel multipass welds: vermicular, lacy, acicular, and globular. The first three ferrite types are related to transformations following solidfication and the fourth is related to the shape instability of the residual ferrite. An earlier study showed that most of the ferrite observed in austenitic stainless steel welds containing a duplex structure may be identified as residual primary ferrite resulting from incomplete delta ..-->.. ..gamma.. transformation during solidification and/or residual ferrite after Widmanstatten austenite precipitation in primary ferrite. These modes of ferrite formation can be used to explain observed ferrite morphologies in austenitic stainless steel welds. Variations in ferrite content within the weld were also related to weld metal composition, ferrite morphology, and dissolution of ferrite resulting from thermal cycles during subsequent weld passes. An investigation of the Type 308 stainless steel filler metal solidified over cooling rates ranging from 7 to 1600/sup 0/C/s (44.6 to 2912/sup 0/F/s) showed that the cooling rate of the weld metal within the freezing range of the alloy affects the amount of ferrite in the microstructure very little. However, the scale of the solidification substructure associated with various solidification rates may influence the ferrite dissolution kinetics.

  16. 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.

  17. A preliminary ferritic-martensitic stainless steel constitution diagram

    SciTech Connect

    Balmforth, M.C.; Lippold, J.C.

    1998-01-01

    This paper describes preliminary research to develop a constitution diagram that will more accurately predict the microstructure of ferritic and martensitic stainless steel weld deposits. A button melting technique was used to produce a wide range of compositions using mixtures of conventional ferritic and martensitic stainless steels, including types 403, 409, 410, 430, 439 and 444. These samples were prepared metallographically, and the vol-% ferrite and martensite was determined quantitatively. In addition, the hardness and ferrite number (FN) were measured. Using this data, a preliminary constitution diagram is proposed that provides a more accurate method for predicting the microstructures of arc welds in ferritic and martensitic stainless steels.

  18. Tensile response of low activation ferritic steels irradiated in ORR at temperatures in the range 60-400 °C

    NASA Astrophysics Data System (ADS)

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

    2002-12-01

    Post-irradiation tensile test results are reported for a series of low activation steels containing manganese following irradiation in the Oak Ridge Reactor at 60, 200, 330 and 400 °C to ˜10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5% and W to 1.0%. Strengths are higher in all alloys for irradiation conditions below 400 °C, with peak hardening occurring following irradiation at 200 °C. The 9Cr alloy class exhibited the smallest increases in hardening. Test results were consistent with previous results obtained on fast flux test facility-irradiated specimens. Manganese does not appear to play a role in the hardening observed at these low irradiation temperatures.

  19. Development of low activation Ferritic steels

    NASA Astrophysics Data System (ADS)

    Noda, T.; Abe, F.; Araki, H.; Okada, M.

    1986-11-01

    Fe-(2-15)%Cr-(0-4)%W-0.1%C and Fe-9%Cr-(0-l)%V-0.1%C steels were prepared on the basis of reduced activation of ferritic steels. Tempering characteristics of these alloys were studied as a preliminary evaluation of mechanical properties. Alloys except for 12-15%Cr, 9%Cr-4%W, and 9%Cr-1%V showed a single phase of martensite. Carbides which precipitated in as-tempered steels are M 23C 6, M 6C, and W 2C for Cr-W steels and M 23C 6 and V 4C 3 for Cr-V steels. The toughness of the alloys was examined with Charpy impact test. The minimum DBTT (ductile-brittle transition temperature) was observed at around 0.25 at% of W or V concentration for 9%Cr steels. 9%Cr-V steels were superior to commercial 9%Cr-2%Mo steel in the point of toughness. The order of alloying element with a low DBTT was V > Mo > W.

  20. 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.

  1. 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.

  2. 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.

  3. Microstructural examination of low activation ferritic steels following irradiation in ORR at 330 and 400 °C to ˜10 dpa

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    2004-08-01

    Microstructural examinations are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 °C to ˜10 dpa. Alloy compositions included 2% Cr, 9% Cr and 12% Cr steels with V to 1.5% and W to 1.0%. 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 °C higher than anticipated.

  4. 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.

  5. 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.

  6. 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.

  7. Neutron irradiation creep in stainless steel alloys

    NASA Astrophysics Data System (ADS)

    Schüle, Wolfgang; Hausen, Hermann

    1994-09-01

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

  8. Design of tough ferritic steels for cryogenic use

    SciTech Connect

    Morris, J.W. Jr.

    1985-10-01

    This paper describes the design of ferritic steels and weldments that combine strength and toughness at cryogenic temperatures. The alloy must have a ductile-brittle transition temperature below the intended service temperature and a high fracture toughness in the ductile mode. Its systematic design uses the microstructure-property relations that govern the transition temperature and fracture toughness to identify a suitable microstructure, and then employs the microstructure-processing relations that govern its thermal response to manipulate the microstructure into the appropriate form. The procedure is illustrated by describing the heat treatments, microstructures and properties of a variety of laboratory and commercial alloys, including conventional ''9Ni'' steel, the low-Ni and Fe-Mn ferritic steels that have been developed as an alternative to 9Ni, the 12Ni steels that are promising for use at 4K, and the welding procedures and ferritic filler metals that are useful for ferritic cryogenic steels.

  9. Delta ferrite in the weld metal of reduced activation ferritic martensitic steel

    NASA Astrophysics Data System (ADS)

    Sam, Shiju; Das, C. R.; Ramasubbu, V.; Albert, S. K.; Bhaduri, A. K.; Jayakumar, T.; Rajendra Kumar, E.

    2014-12-01

    Formation of delta(δ)-ferrite in the weld metal, during autogenous bead-on-plate welding of Reduced Activation Ferritic Martensitic (RAFM) steel using Gas Tungsten Arc Welding (GTAW) process, has been studied. Composition of the alloy is such that delta-ferrite is not expected in the alloy; but examination of the weld metal revealed presence of delta-ferrite in the weld metal. Volume fraction of delta-ferrite is found to be higher in the weld interface than in the rest of the fusion zone. Decrease in the volume fraction of delta-ferrite, with an increase in preheat temperature or with an increase in heat input, is observed. Results indicate that the cooling rate experienced during welding affects the volume fraction of delta-ferrite retained in the weld metal and variation in the delta-ferrite content with cooling rate is explained with variation in the time that the weld metal spends in various temperature regimes in which delta-ferrite is stable for the alloy during its cooling from the liquid metal to the ambient temperature. This manuscript will discuss the effect of welding parameters on formation of delta-ferrite and its retention in the weld metal of RAFM steel.

  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. 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)

  13. 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.

  14. ON QUANTIFICATION OF HELIUM EMBRITTLEMENT IN FERRITIC/MARTENSITIC STEELS

    SciTech Connect

    Gelles, David S.

    2000-12-01

    Helium accumulation due to transmutation has long been considered a potential cause for embrittlement in ferritic/martensitic steels. Three Charpy impact databases involving nickel- and boron-doped alloys are quantified with respect to helium accumulation, and it is shown that all predict a very large effect of helium production on embrittlement. If these predictions are valid, use of Ferritic/Martensitic steels for Fusion first wall applications is highly unlikely. It is therefore necessary to reorient efforts regarding development of these steels for fusion applications to concentrate on the issue of helium embrittlement.

  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. Ferritic weldment of grain-refined ferritic steels for cryogenic use

    SciTech Connect

    Kim, H.J.; Syn, C.K.; Morris, J.W. Jr.

    1981-08-01

    The problem of welding grain-refined Fe-12Ni-0.25Ti for 4K service was first approached in this laboratory by using high nickel filler metals such as are often specified for ferritic steel weldments at 77K. This approach led to an undesirable brittleness in the fusion zone and a low yield strength in the weld metal. A more promising approach was developed in joint research between the Japanese steel companies, who showed that quench-and-tempered 9Ni steel may be welded for 77K service with a matching ferritic filler if a multipass GTAW technique is employed. The present paper reports the initial resultsof similar studies on ferritic GTA weldments in grain-refined 9Ni steel. Information is included on the preparation of the 9Ni steel and the weld filler metal, on the welding procedure, the microstructure of both the weld metal and the heat affected zone, and on impact toughness and fracture toughness testing at 77/sup 0/K and 4.2/sup 0/K. The results show that it is possible to weld grain-refined 9Ni steel with ferritic weld filler metal so as to retain good toughness at cryogenic temperatures. The results of this work may permit the utilization of retreated commercial grade 9Ni steel in structural applications within helium-cooled cryogenic devices where high strength and good toughness are required. (LCL)

  17. Recent Progress of R&D Activities on Reduced Activation Ferritic/Martensitic Steels

    SciTech Connect

    Huang, Q.; Baluc, N.; Dai, Y.; Jitsukawa, S.; Kimura, A.; Konys, J.; Kurtz, Richard J.; Lindau, R.; Muroga, T.; Odette, George R.; Raj, B.; Stoller, Roger E.; Tan, L.; Tanigawa, Hiroyasu; Tavassoli, A,-A.F.; Yamamoto, Takuya; Wan, F.; Wu, Y.

    2013-01-03

    Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30 years in China, Europe, India, Japan, Russia and the USA for application in ITER TBM and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical porperties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation with different TBM and DEMO options

  18. 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...

  19. 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.

  20. Microstructure and texture of Nb + Ti stabilized ferritic stainless steel

    SciTech Connect

    Yan Haitao Bi Hongyun; Li Xin; Xu Zhou

    2008-12-15

    The microstructure, texture and grain boundary character distribution of Nb + Ti stabilized ferritic stainless steel were analyzed using scanning electron microscopy, transmission electron microscopy and X-ray diffraction. The addition of alloying elements such as Ti and Nb to ferritic stainless steel causes the formation of TiN, NbC and Fe{sub 2}Nb. The textures of cold rolled samples were dominated by the {alpha}-fiber, while the textures of annealed samples exhibit a very strong {gamma}-fiber. The changes in texture are closely related to the grain boundary characteristics.

  1. The filler powders laser welding of ODS ferritic steels

    NASA Astrophysics Data System (ADS)

    Liang, Shenyong; Lei, Yucheng; Zhu, Qiang

    2015-01-01

    Laser welding was performed on Oxide Dispersion Strengthened (ODS) ferritic steel with the self-designed filler powders. The filler powders were added to weld metal to produce nano-particles (Y-M-O and TiC), submicron particles (Y-M-O) and dislocation rings. The generated particles were evenly distributed in the weld metal and their forming mechanism and behavior were analyzed. The results of the tests showed that the nano-particles, submicron particles and dislocation rings were able to improve the micro-hardness and tensile strength of welded joint, and the filler powders laser welding was an effective welding method of ODS ferritic steel.

  2. 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.

  3. 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.

  4. Microstructural probing of ferritic/martensitic steels using internal transmutation-based positron source

    NASA Astrophysics Data System (ADS)

    Krsjak, Vladimir; Dai, Yong

    2015-10-01

    This paper presents the use of an internal 44Ti/44Sc radioisotope source for a direct microstructural characterization of ferritic/martensitic (f/m) steels after irradiation in targets of spallation neutron sources. Gamma spectroscopy measurements show a production of ∼1MBq of 44Ti per 1 g of f/m steels irradiated at 1 dpa (displaced per atom) in the mixed proton-neutron spectrum at the Swiss spallation neutron source (SINQ). In the decay chain 44Ti → 44Sc → 44Ca, positrons are produced together with prompt gamma rays which enable the application of different positron annihilation spectroscopy (PAS) analyses, including lifetime and Doppler broadening spectroscopy. Due to the high production yield, long half-life and relatively high energy of positrons of 44Ti, this methodology opens up new potential for simple, effective and inexpensive characterization of radiation induced defects in f/m steels irradiated in a spallation target.

  5. Dynamic recrystallization of ferrite in interstitial free steel

    SciTech Connect

    Tsuji, N.; Matsubara, Y.; Saito, Y.

    1997-08-15

    The present study using IF steel confirmed that dynamic recrystallization can occur also in ferrite where it has been generally considered that recovery is an only restoration process during hot deformation. Although the occurrence of DRX has been clarified by microstructural observations and crystallographic determinations, stress-strain curves do not show obvious drop of stress which has been typically reported in the case of DRX of austenite. This result indicates that it is quite difficult to distinguish whether DRX occurs in ferrite only by stress-strain behavior. The noticeable feature of DRX of ferrite is inhomogeneity of recrystallization, i.e., some of the initial grains are hard to recrystallize. This is presumably due to orientation dependence of recrystallization, which is the essential feature of ferrite.

  6. Irradiation Assisted Grain Boundary Segregation in Steels

    SciTech Connect

    Lu, Zheng; Faulkner, Roy G.

    2008-07-01

    The understanding of radiation-induced grain boundary segregation (RIS) has considerably improved over the past decade. New models have been introduced and much effort has been devoted to obtaining comprehensive information on segregation from the literature. Analytical techniques have also improved so that chemical analysis of layers 1 nm thick is almost routine. This invited paper will review the major methods used currently for RIS prediction: namely, Rate Theory, Inverse Kirkendall, and Solute Drag approaches. A summary is made of the available data on phosphorus RIS in reactor pressure vessel (RPV) steels. This will be discussed in the light of the predictions of the various models in an effort to show which models are the most reliable and easy to use for forecasting P segregation behaviour in steels. A consequence of RIS in RPV steels is a radiation induced shift in the ductile to brittle transition temperature (DBTT). It will be shown how it is possible to relate radiation-induced P segregation levels to DBTT shift. Examples of this exercise will be given for RPV steels and for ferritic steels being considered for first wall fusion applications. Cr RIS in high alloy stainless steels and associated irradiation-assisted stress corrosion cracking (IASCC) will be briefly discussed. (authors)

  7. 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.

  8. Anomalous ultrasonic attenuation in ferritic steels at elevated temperatures.

    PubMed

    Hutchinson, Bevis; Lundin, Peter; Lindh-Ulmgren, Eva; Lévesque, Daniel

    2016-07-01

    An unexpected peak in attenuation has been observed at ∼800°C when heating low carbon steels in a laser-ultrasonic instrument. An explanation is given in terms of enhanced crystalline anisotropy with increasing temperature in the bcc ferrite range combined with subsequent transformation to austenite at still higher temperatures. An analysis based on theoretical models of attenuation in the Rayleigh regime is in good agreement with the experimental observations. PMID:27015796

  9. 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%).

  10. 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.

  11. 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)

  12. Irradiation creep and swelling from 400 to 600 °C of the oxide dispersion strengthened ferritic alloy MA957

    NASA Astrophysics Data System (ADS)

    Toloczko, M. B.; Gelles, D. S.; Garner, F. A.; Kurtz, R. J.; Abe, K.

    2004-08-01

    An irradiation creep and swelling study was performed on the Y 2O 3-strengthened ODS ferritic steel MA957. Pressurized tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 to 110 dpa at temperatures ranging from 400 to 600 °C. None of the stress-free tubes exhibited any evidence of swelling as determined by diameter change measurements. With a few exceptions, the irradiation creep behavior is similar to that of conventional ferritic-martensitic steels. Calculated creep compliance values are equal to those of HT9 irradiated within the same temperature range, except at 600 °C where the creep rate of MA957 is about one-half the value for HT9. The magnitude of the creep transient for MA957 is comparable to HT9, again except at 600 °C where the transient is much lower for MA957.

  13. UNDERSTANDING DAMAGE MECHANISMS IN FERRITIC/MARTENSITIC STEELS

    SciTech Connect

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

    2003-04-22

    Advanced ferritic/martensitic steels are being used extensively in fossil energy applications. New steels such as 2 1/4Cr-W-V (T23, T24), 3Cr-W-V, 9Cr-Mo-V (T91), 7Cr-W-V, 9Cr-W-V (T92 and T911), and 12Cr-W-V (T122, SAVE 12, and NF12) are examples of tubing being used in boilers and heat recovery steam generators (1). Other products for these new steels include piping, plates, and forgings. There is concern about the high-temperature performance of the advanced steels for several reasons. First, they exhibit a higher sensitivity to temperature than the 300 series stainless steels that they often replace. Second, they tend to be metallurgically unstable and undergo significant degradation at service temperatures in the creep range. Third, the experience base is limited in regard to duration. Fourth, they will be used for thick-section, high-pressure components that require high levels of integrity. To better understand the potential limitations of these steels, damage models are being developed that consider metallurgical factors as well as mechanical performance factors. Grade 91 steel was chosen as representative of these steels for evaluation of cumulative damage models since laboratory and service exposures of grade 91 exceed 100,000 hours.

  14. A micro-alloyed ferritic steel strengthened by nanoscale precipitates

    SciTech Connect

    Shen, Yongfeng; Wang, Chong M.; Sun, Xin

    2011-10-25

    A ferritic steel with finely dispersive precipitates was investigated to reveal the fundamental strengthening mechanisms. The steel has a yield strength of 760 MPa, approximately three times higher than that of conventional Ti-bearing high strength hot-rolled sheet steels, and its ultimate tensile strength reaches 850 MPa with an elongation-to-failure value of 18%. Using energy dispersive X-ray spectroscopy (EDXS) and transmission electron microscope (TEM), fine carbides TiC with an average diameter of 10 nm were observed in the ferrite matrix of the 0.08%Ti steel, and some cubic M23C6 precipitates were also observed at the grain boundaries and the interior of the grains. The finely dispersive TiC precipitates in the matrix provide matrix strengthening. The estimated magnitude of precipitation strengthening is around 458 MPa, depending on the average size of the nanoscale precipitates. Dislocation densities increased from 3.42×1013 m-2 to 1.69 × 1014 m-2, respectively, with increasing tensile strain from 5.5% to 22%. The measured work-hardening behavior can be related to the observed dislocation accumulations resulting from the dispersive nano-scale precipitates.

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

    SciTech Connect

    Leitnaker, J.M.

    1981-05-05

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

  16. 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

  17. 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.

  18. Development of ferritic weldments for grain-refined ferritic steels for 4. 2K service

    SciTech Connect

    Kim, H.J.

    1982-11-01

    The weldability of grain-refined ferritic nickel steels designed for structural use in liquid helium was investigated. Plates of interstitial-free Fe-12Ni-0.25Ti alloy and carbon-containing 9 Ni steel were welded with 14 Ni ferritic fillers using a gas tungsten arc welding (GTAW) process with pure argon gas shielding. The ferritic weldments made have a strength closely matching those of the base plates without a significant loss in base metal toughness at temperatures as low as 4.2 K. The comparable toughness obtained in the welded region is attributed to three factors; the defect-free weldment, the chemical cleanliness of the GTAW weld deposit, and the in-process formation of an appropriate microstructure in the welded region. Special emphasis in this study was placed on changes in microstructures with respect to the characteristic of the weld thermal cycles and the effect of the resultant microstructures on low temperature toughness. In the heat-affected zone (HAZ) of multipass welded 9Ni steel, the retained (or precipitated) austenite is removed by the weld heat cycles but the sequential rapid heat cycles to successively lower peak temperatures associated with succeeding weld passes re-establish high toughness by sequentially refining the grain size and gettering carbon in the form of cementite precipitates. On the other hand, the high toughness in the HAZ of the 12Ni alloy and in the weld deposit is a direct consequence of repeated grain refinement through the overlapped austenitizing cycles and is not affected by the tempering cycles because of the carbon-free nature of these materials. 46 figures.

  19. Development of ferritic weldments for grain-refined ferritic steels for 4. 2K service

    SciTech Connect

    Kim, H.J.

    1982-01-01

    The weldability of grain-refined ferritic nickel steels designed for structural use in liquid helium was investigated. Plates of interstitial-free Fe-12Ni-0.25Ti alloy and carbon-containing 9Ni steel were welded with 14Ni ferritic fillers using a gas tungsten arc welding (GTAW) process with pure argon gas shielding. The ferritic weldments made have a strength closely matching those of the base plates without a significant loss in base metal toughness at temperatures as low as 4.2K. The comparable toughness obtained in the welded region is attributed to three factors; the defect-free weldment, the chemical cleanliness of the GTAW weld deposit, and the in-process formation of an appropriate microstructure in the welded region. Special emphasis in this study was placed on changes in microstructures with respect to the characteristic of the weld thermal cycles and the effect of the resultant microstructures on low temperature toughness. In the heat-affected zone (HAZ) of multipass welded 9Ni steel, the retained (or precipitated) austenite is removed by the weld heat cycles but the sequential rapid heat cycles to successively lower peak temperatures associated with succeeding weld passes re-establish high toughness by sequentially refining the grain size and gettering carbon in the form of cementite precipitates. On the other hand, the high toughness in the HAZ of the 12Ni alloy and in the weld deposit is a direct consequence of repeated grain refinement through the overlapped austenitizing cycles and is not affected by the tempering cycles because of the carbon-free nature of these materials.

  20. 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.

  1. 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.''...

  2. Pros and cons of nickel- and boron-doping to study helium effects in ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Hashimoto, N.; Klueh, R. L.; Shiba, K.

    2002-12-01

    In the absence of a 14 MeV neutron source, the effect of helium on structural materials for fusion must be simulated using fission reactors. Helium effects in ferritic/martensitic steels have been studied by adding nickel and boron and irradiating in a mixed-spectrum reactor. Although the nickel- and boron-doping techniques have limitations and difficulties to estimate helium effects on the ferritic/martensitic steels, past irradiation experiments using these techniques have demonstrated similar effects on the swelling and Charpy impact properties that are indicative of a helium effect. Although both techniques have disadvantages, it should be possible to plan experiments using the nickel- and boron-doping techniques to develop an understanding of the effects of helium on mechanical properties.

  3. Carbides in a High-Chromium Ferritic/Martensitic Steel

    NASA Astrophysics Data System (ADS)

    Shen, Yin Zhong; Ji, Bo; Zhou, Xiao Ling; Zhu, Jun

    2014-06-01

    The precipitate phases in an 11 pct Cr ferritic/martensitic steel normalized at 1323 K (1050 °C) for 0.5 hour and tempered at 1053 K (780 °C) for 1.5 hours have been investigated. Except for dominant phases, Cr-rich M23C6 carbide and Nb-rich, Ta-Nb-rich, and V-rich MC carbides, needle-like precipitates with a typical size of 70 to 310 and 10 to 30 nm for the length of the long and short axis of the needles, respectively, were also observed on the extraction carbon replica of the steel. The typical metallic element composition of the needle-like precipitates is about 53-82Fe, 14-26Cr, 0.5-18Ta, 1-6W, and 2-5Co in atomic pct. Through energy dispersive X-ray analysis and electron diffraction along with calculations regarding lattice parameter and interplanar spacing, the needle-like precipitates were identified as a Fe-rich M5C2 carbide, which is not known to have been reported previously in high-chromium steels. The M5C2 carbide has a base-centered monoclinic crystal structure with the approximate lattice parameters a/ b/ c = 1.142/0.5186/0.5383 nm and β = 104.68 deg. The formation of the Fe-rich M5C2 carbides in the steel has been discussed. The effect of chromium content in matrix and boron addition on the precipitate phases in ferritic/martensitic steels has also been discussed.

  4. 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.

  5. Microstructural development in reduced activation ferritic alloys irradiated to 200 dpa at 420$deg;C

    NASA Astrophysics Data System (ADS)

    Gelles, D. S.

    1994-09-01

    Density change and microstructural development are reported for nine reduced activation ferritic steels covering the range 2.3 to 12Cr with varying additions of V and/or W for hardening and up to 6.5 Mn for austenite stability. Specimens were examined following irradiation in FFTF/MOTA at 420°C to a dose exceeding 200 dpa. Void swelling was found, but the swelling remained at 5% or below, with the worst case in an alloy of 9Cr-2Mn-1WV. The carbide structure pinning martensite lath boundaries remained in place.

  6. 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.

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

    PubMed

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

    2016-02-01

    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. PMID:26932112

  8. 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.

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

    SciTech Connect

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

    1997-10-15

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

  10. Stress-relief cracking of a new ferritic steel

    NASA Astrophysics Data System (ADS)

    Nawrocki, Jesse Gerald

    The mechanism of stress-relief cracking in the coarse-grained heat-affected zone (CGHAZ) of low-alloy ferritic steels was studied through a tempering study, stress-relaxation testing, and detailed microstructural characterization. A new ferritic alloy steel, HCM2S, was used as the model system. Common 2.25Cr-1 Mo steel, which is susceptible to stress-relief cracking, was used for comparison to HCM2S. The CGHAZ was simulated using Gleeble techniques. A dense distribution of small tungsten-rich carbides within the prior austenite grains induced secondary hardening in the CGHAZ of HCM2S. The CGHAZ of 2.25Cr-1 Mo steel exhibited secondary hardening due to the intragranular precipitation of many Fe-rich M3C carbides. The hardness of HCM2S was more stable at longer times and high temperatures than 2.25Cr-1 Mo steel due to the intragranular precipitation of small W and V-rich carbides. The CGHAZs of HCM2S and 2.25Cr-1 Mo steel were susceptible to stress-relief cracking between 575 and 725°C. HCM2S exhibited C-curve behavior with respect to the time to failure as a function of post-weld heat treatment (PWHT) temperature. No segregation of tramp elements to prior austenite grain boundaries was detected in HCM2S. Both intergranular and intragranular carbide precipitation controlled the stress-relief cracking behavior. The amount of intergranular failure increased with test temperature due to the increasing amounts of Fe-rich M3C carbides at the prior austenite grain boundaries. These carbides acted as cavity nucleation sites. The cavities coalesced to form microcracks along prior austenite grain boundaries. Eventually, the remaining uncracked areas could not support the load and failed by ductile rupture. The balance of intergranular and intragranular carbide precipitation resulted in the C-curve behavior. The nose of the C-curve occurred at 675°C. The intragranular regions were strong because of a dense distribution of W/Fe-rich carbides, but the prior austenite grain

  11. Unraveling the Effect of Thermomechanical Treatment on the Dissolution of Delta Ferrite in Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Rezayat, Mohammad; Mirzadeh, Hamed; Namdar, Masih; Parsa, Mohammad Habibi

    2016-02-01

    Considering the detrimental effects of delta ferrite stringers in austenitic stainless steels and the industrial considerations regarding energy consumption, investigating, and optimizing the kinetics of delta ferrite removal is of vital importance. In the current study, a model alloy prone to the formation of austenite/delta ferrite dual phase microstructure was subjected to thermomechanical treatment using the wedge rolling test aiming to dissolve delta ferrite. The effect of introducing lattice defects and occurrence of dynamic recrystallization (DRX) were investigated. It was revealed that pipe diffusion is responsible for delta ferrite removal during thermomechanical process, whereas when the DRX is dominant, the kinetics of delta ferrite dissolution tends toward that of the static homogenization treatment for delta ferrite removal that is based on the lattice diffusion of Cr and Ni in austenite. It was concluded that the optimum condition for dissolution of delta ferrite can be defined by the highest rolling temperature and strain in which DRX is not pronounced.

  12. Radiation damage in ferritic/martensitic steels for fusion reactors: a simulation point of view

    NASA Astrophysics Data System (ADS)

    Schäublin, R.; Baluc, N.

    2007-12-01

    Low activation ferritic/martensitic steels are good candidates for the future fusion reactors, for, relative to austenitic steels, their lower damage accumulation and moderate swelling under irradiation by the 14 MeV neutrons produced by the fusion reaction. Irradiation of these steels, e.g. EUROFER97, is known to produce hardening, loss of ductility, shift in ductile to brittle transition temperature and a reduction of fracture toughness and creep resistance starting at the lowest doses. Helium, produced by transmutation by the 14 MeV neutrons, is known to impact mechanical properties, but its effect at the microstructure level is still unclear. The mechanisms underlying the degradation of mechanical properties are not well understood, despite numerous studies on the evolution of the microstructure under irradiation. This impedes our ability to predict materials' behaviour at higher doses for use in the future fusion reactors. Simulations of these effects are now essential. An overview is presented on molecular dynamics simulations of the primary state of damage in iron and of the mobility of a dislocation, vector of plasticity, in the presence of a defect.

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

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  14. 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.

  15. Effect of V and Ta on the precipitation behavior of 12%Cr reduced activation ferrite/martensite steel

    SciTech Connect

    Xiao, Xiang; Liu, Guoquan; Hu, Benfu; Wang, Jinsan; Ullah, Asad

    2013-08-15

    12%Cr reduced activation ferrite/martensite steels are promising candidate materials for good corrosion and irradiation resistance used for supercritical water-cooled reactor cladding and in-core components. V and Ta are considered to have improved the creep strength of high Cr steels by precipitating as MX phase. In this paper, a series of trial products microalloyed with V and V–Ta are produced, and the microstructure is characterized after quenching at 1050 °C and tempering at 780 °C by using TEM method to investigate the effect of these elements on the precipitation behavior of 12%Cr reduced activation ferrite/martensite steel. The results from both the experimental observations and thermodynamic and kinetic calculations reveal that V and V–Ta can promote the stable MX precipitation instead of M{sub 2}X, thus increasing the volume fraction of M{sub 23}C{sub 6}. Two-phase separation behavior of the (Ta, V)(C, N) carbonitride into a Ta(V)C(N) phase and a V(Ta)N(C) phase in 12Cr3WVTa steel is observed and further discussed. - Highlights: • Microalloyed with V and V-Ta can promote the precipitation of MX instead of M{sub 2}X. • The presence of delta-ferrite in microstructure affects the morphology of MX. • Two-phase separation of MX carbonitride was observed in 12Cr3WVTa steel.

  16. Fractal characteristics of fracture morphology of steels irradiated with high-energy ions

    NASA Astrophysics Data System (ADS)

    Xian, Yongqiang; Liu, Juan; Zhang, Chonghong; Chen, Jiachao; Yang, Yitao; Zhang, Liqing; Song, Yin

    2015-06-01

    A fractal analysis of fracture surfaces of steels (a ferritic/martensitic steel and an oxide-dispersion-strengthened ferritic steel) before and after the irradiation with high-energy ions is presented. Fracture surfaces were acquired from a tensile test and a small-ball punch test (SP). Digital images of the fracture surfaces obtained from scanning electron microscopy (SEM) were used to calculate the fractal dimension (FD) by using the pixel covering method. Boundary of binary image and fractal dimension were determined with a MATLAB program. The results indicate that fractal dimension can be an effective parameter to describe the characteristics of fracture surfaces before and after irradiation. The rougher the fracture surface, the larger the fractal dimension. Correlation of the change of fractal dimension with the embrittlement of the irradiated steels is discussed.

  17. 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.

  18. 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.

  19. A Micro-Alloyed Ferritic Steel Strengthened by Nanoscale Precipitates

    SciTech Connect

    Shen, Yongfeng; Wang, Chong M.; Sun, Xin

    2011-08-04

    A high strength ferritic steel with finely dispersive precipitates was investigated to reveal the fundamental strengthening mechanisms. Using energy dispersive X-ray spectroscopy (EDXS) and transmission electron microscope (TEM), fine carbides with an average diameter of 10 nm were observed in the ferrite matrix of the 0.08%Ti steel, and some cubic M23C6 precipitates were also observed at the grain boundaries and the interior of grains. The dual precipitate structure of finely dispersive TiC precipitates in the matrix and coarse M23C6 at grain boundaries provides combined matrix and grain boundary strengthening. The calculated amount of precipitation strengthening by the carbides was approximately 450 ~ 630 MPa, depending on the average size of nanoscale precipitates. This value is two or three times higher than that of conventional Ti-bearing high strength hot-rolled sheet steels. Dislocation densities increased from 3.42×1013 m-2 to 1.69 × 1014 m-2, espectively, with increasing tensile strain from 5.5% to 22%. The effect of the particle size, particle distribution and intrinsic particle strength have been investigated through dislocation dynamics (DD) simulations and the relationship for resolved shear stress for single crystal under this condition has been presented using simulation data. The results show that the finely dispersive precipitates can strengthen the material by pinning the dislocations up to a certain shear stress and retarding the recovery as well as annihilation of dislocations. The DD results also show that strengthening is not only a function of the density of the nano-scale precipitates but also of their size.

  20. 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.

  1. Fatigue Crack Growth Behavior of Gas Metal Arc Welded AISI 409 Grade Ferritic Stainless Steel Joints

    NASA Astrophysics Data System (ADS)

    Lakshminarayanan, A. K.; Shanmugam, K.; Balasubramanian, V.

    2009-10-01

    The effect of filler metals such as austenitic stainless steel, ferritic stainless steel, and duplex stainless steel on fatigue crack growth behavior of the gas metal arc welded ferritic stainless steel joints was investigated. Rolled plates of 4 mm thickness were used as the base material for preparing single ‘V’ butt welded joints. Center cracked tensile specimens were prepared to evaluate fatigue crack growth behavior. Servo hydraulic controlled fatigue testing machine with a capacity of 100 kN was used to evaluate the fatigue crack growth behavior of the welded joints. From this investigation, it was found that the joints fabricated by duplex stainless steel filler metal showed superior fatigue crack growth resistance compared to the joints fabricated by austenitic and ferritic stainless steel filler metals. Higher yield strength and relatively higher toughness may be the reasons for superior fatigue performance of the joints fabricated by duplex stainless steel filler metal.

  2. Hot-rolling of reduced activation 8CrODS ferritic steel

    NASA Astrophysics Data System (ADS)

    Wu, Xiaochao; Ukai, Shigeharu; Leng, Bin; Oono, Naoko; Hayashi, Shigenari; Sakasegawa, Hideo; Tanigawa, Hiroyasu

    2013-11-01

    The 8CrODS ferritic steel is based on J1-lot developed for the advanced fusion blanket material to increase the coolant outlet temperature. A hot-rolling was conducted at the temperature above Ar3 of 716 °C, and its effect on the microstructure and tensile strength in 8CrODS ferritic steel was evaluated, comparing together with normalized and tempered specimen. It was confirmed that hot-rolling leads to slightly increased fraction of the ferrite and highly improved tensile strength. This ferrite was formed by transformation from the hot-rolled austenite during cooling due to fine austenite grains induced by hot-rolling. The coarsening of the transformed ferrite in hot-rolled specimen can be attributed to the crystalline rotation and coalescence of the similar oriented grains. The improved strength of hot-rolled specimen was ascribed to the high dislocation density and replacement of easily deformed martensite with the transformed coarse ferrite.

  3. Investigation of microstructure and thermal stability of pulsed plasma processed chromium ferritic-martensitic steels

    NASA Astrophysics Data System (ADS)

    Emelyanova, O.; Dzhumaev, P.; Yakushin, V.; Polsky, V.

    2016-04-01

    This paper presents results of the microstructural evolution and thermal stability of the promising Russian ferritic-martensitic steels (EP 823, EP 900, EK 181 and ChS 139) for the nuclear and fusion application after surface modification by high temperature pulsed plasma flows (HTPPF) treatment. Investigations of microstructure, topography and elemental content changes associated with irradiation by nitrogen plasma with energy density 19-28 J/ cm2 and pulse duration 20 μs were carried out. Changes in microstructure and elemental content occurring in the modified surface layer were characterized by means of scanning electron microscopy (SEM) and X-ray microanalysis (EDS and WDS). It was shown that independently of initial microstructure and phase composition, HTPPF treatment of ferritic- martensitic steels leads to formation of ultrafine homogeneous structure in the near surface layers with typical grain size ∼100 nm. Results of microstructure investigations after annealing during 1 hour demonstrates significant thermal stability of nanostructure formed by HTPPF treatment.

  4. 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.

  5. 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.

  6. 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.

  7. The Irradiation Performance and Microstructural Evolution in 9Cr-2W Steel Under Ion Irradiation

    NASA Astrophysics Data System (ADS)

    Alsagabi, Sultan; Charit, Indrajit; Pasebani, Somayeh

    2016-02-01

    Grade 92 steel (9Cr-2W) is a ferritic-martensitic steel with good mechanical and thermal properties. It is being considered for structural applications in Generation IV reactors. Still, the irradiation performance of this alloy needs more investigation as a result of the limited available data. The irradiation performance investigation of Grade 92 steel would contribute to the understanding of engineering aspects including feasibility of application, economy, and maintenance. In this study, Grade 92 steel was irradiated by iron ion beam to 10, 50, and 100 dpa at 30 and 500 °C. In general, the samples exhibited good radiation damage resistance at these testing parameters. The radiation-induced hardening was higher at 30 °C with higher dislocation density; however, the dislocation density was less pronounced at higher temperature. Moreover, the irradiated samples at 30 °C had defect clusters and their density increased at higher doses. On the other hand, dislocation loops were found in the irradiated sample at 50 dpa and 500 °C. Further, the irradiated samples did not show any bubble or void.

  8. Analysis of delta-ferrite data from production stainless steel pipe welds

    SciTech Connect

    Hebble, T.L.; Canonico, D.A.; Edmonds, D.P.; Goodwin, G.M.; Nanstad, R.K.

    1984-01-01

    An American Society of Mechanical Engineers task group on stainless steel weld materials was organized to determine the need for ferrite measurements of production welds required by the US Nuclear Regulatory Commission Regulatory Guide 1.31 (Rev. 1). The task group studied paired ferrite measurements (i.e., calculated and measured ferrite numbers (FNs) for the material qualifications versus measured ferrite numbers for corresponding production welds (PWs)). Our purpose was to compare delta-ferrite content as measured in the filler metal weld qualification pad with that in the resultant PW. Welds made predominantly by three common processes (submerged arc, shielded metal arc, and gas tungsten arc) were included in the study. Weld metals investigated included types 308, 308L, 316, and 316L stainless steel. An initial evaluation of the paired ferrite measurements was made by the task group, and specific conclusions and recommendations were made. We describe the analysis of the data and the conclusions drawn. The data base consisted of a heterogeneous collection of 1449 paired ferrite measurements for several forms and combinations of types 304 and 316 stainless steel pipe qualification pad and production welds. Qualification pad values ranged from 5 to 15 FN, and corresponding values for the PWs ranged from 2.3 to 17.5 FN. Only two PW ferrite numbers were less than 3. For qualification weld ferrite numbers less than 14, the median PW ferrite number was in reasonable agreement. However, the results show a wide scatter. As a result of this analysis and the task group evaluation, we concluded that the requirements of Regulatory Guide 1.31 on the measurement of ferrite in PWs are not necessary and that a ferrite number of 5 in the qualification welds will, in most cases, result in PW ferrite contents greater than 3 FN.

  9. Mechanical behaviour of ferritic ODS steels - Temperature dependancy and anisotropy

    NASA Astrophysics Data System (ADS)

    Fournier, B.; Steckmeyer, A.; Rouffie, A.-L.; Malaplate, J.; Garnier, J.; Ratti, M.; Wident, P.; Ziolek, L.; Tournie, I.; Rabeau, V.; Gentzbittel, J. M.; Kruml, T.; Kubena, I.

    2012-11-01

    Ferritic 14%Cr and 18%Cr ODS steels produced at CEA in round bars or plates were tested mechanically. The present paper reports results obtained in tension, impact, fatigue, creep and toughness tests. These tests were carried out at various temperatures and in different directions. These materials show a pronounced anisotropy at all tested temperatures. No matter the loading, the transversal direction is always found to be far less resistant than the longitudinal one. This anisotropy is mainly observed in terms of damage mechanisms, with intergranular fracture preferentially occurring along the extrusion direction. This intergranular fracture mode leads to very low and anisotropic toughness values and to the absence of tertiairy creep stage, pointing out the unstable nature of fracture, even at high temperature. The unrealistically high values of the Norton exponent measured in creep suggests the existence of a threshold stress, which is consistent with the mainly kinematic nature of the stress as revealed by fatigue tests.

  10. Calculation of the ferrite volume in some dual phase steels

    SciTech Connect

    Qixun, D.; Ruzeng, Y.

    1997-04-01

    The relation between the {gamma}/{gamma} + {alpha} boundary temperature, T{sub {alpha}}, and the equivalent values of [Cr] and [Ni], as well as the variation of the ferrite volume, V{sub f}, with the temperature in {alpha} + {gamma} dual-phase steels have been studied. With the aid of a computer, the regressive expressions derived from the experimental results are: T{sub {alpha}} (C) = T{sub 3} + 21.2 [Cr] {minus} 15.8 [Ni] + 223; V{sub f} (%) = 0.715 {l_brace}exp [0.015(T-T{sub {delta}})] {minus} exp[0.015(T{sub c}-T{sub {delta}})]{r_brace} + 1.85 exp [0.0083(T{sub {alpha}}-T{sub c})].

  11. The effect of a tin barrier layer on the permeability of hydrogen through mild steel and ferritic stainless steel

    SciTech Connect

    Bowker, J.; Piercy, G.R.

    1984-11-01

    Experiments were performed to measure the effectiveness of a commercially electroplated tin layer as a barrier to hydrogen, and to see how this altered when the tin layer was converted to FeSn. The authors measured the permeability of hydrogen through AISI 410 ferritic stainless steel and determined the effectiveness of tin as a surface barrier on it. The measured values for the permeability of hydrogen in iron and ferritic stainless steel are shown.

  12. Effect of Nb on high-temperature properties for ferritic stainless steel

    SciTech Connect

    Fujita, N.; Kikuchi, M.; Ohmura, K.; Suzuki, T.; Funaki, S.; Hiroshige, I.

    1996-09-15

    In order to improve the efficiency of automobile engines and to reduce their weight, there is a move toward the use of conventional stainless steel sheets and pipes for exhaust manifolds to replace cast iron, the traditional material for this application. The exhaust manifold is used in an environment that includes engine vibrations as well as heating and cooling cycles caused by the travel pattern. Therefore, among high-temperature characteristics, thermal fatigue resistance is an important one that affects the life span of an exhaust manifold. Generally, austenitic steels have higher strength at high temperature than ferritic steels. However, type 304, a typical austenitic stainless steel, has less thermal fatigue resistance than type 430, a typical ferritic stainless steel. This is because austenitic steels have higher coefficient of thermal expansion than ferritic steels. Therefore, to obtain a material with excellent thermal fatigue resistance, it would conceivably be best to attempt to increase the high temperature strength of ferritic stainless steels. The present study centered on improvement of the high-temperature proof strength of ferritic stainless steels. The mechanism of high temperature strengthening by Nb addition, which was shown to be one of the most effective methods to improve proof strength at high temperature, was discussed.

  13. Ferritic weldment of grain-refined ferritic steels for cryogenic service

    SciTech Connect

    Kim, H.J.; Morris, J.W. Jr.; Syn, C.K.

    1982-01-01

    This paper reports the initial results of the development of a weld filler metal and multipass GTA welding process for grain-refined 9Ni steel. The weld filler metal was cast in 4.5-kg ingots of the nominal composition Fe-14Ni-O.2Ti-0.003B, homogenized, hot-rolled, and then swaged into wire of 1.6-mm diameter. The chemical compositions of the base and filler metal are tabulated and the test welding procedure is described. Among the tests that were conducted on the welds are x-ray diffraction analysis for the residual austenite content; the Charpy impact test at 77 and 4.2 K for the fracture toughness; and scanning electron fractography for the weld metal, the HAZ, and the base metal of 9Ni Charpy specimens broken at 4.2 K. The procedure for grain refinement is defined; the retained austenite volume fraction is calculated using the method proposed by Miller. The results indicate that it is possible to weld grain-refined 9Ni steel with a ferritic weld filler metal and retain good toughness at cryogenic liquid helium temperatures.

  14. A Comparison of Creep Rupture Strength of Ferritic/Austenitic Dissimilar Weld Joints of Different Grades of Cr-Mo Ferritic Steels

    NASA Astrophysics Data System (ADS)

    Laha, K.; Chandravathi, K. S.; Parameswaran, P.; Goyal, Sunil; Mathew, M. D.

    2012-04-01

    Evaluations of creep rupture properties of dissimilar weld joints of 2.25Cr-1Mo, 9Cr-1Mo, and 9Cr-1MoVNb steels with Alloy 800 at 823 K were carried out. The joints were fabricated by a fusion welding process employing an INCONEL 182 weld electrode. All the joints displayed lower creep rupture strength than their respective ferritic steel base metals, and the strength reduction was greater in the 2.25Cr-1Mo steel joint and less in the 9Cr-1Mo steel joint. Failure location in the joints was found to shift from the ferritic steel base metal to the intercritical region of the heat-affected zone (HAZ) of the ferritic steel (type IV cracking) with the decrease in stress. At still lower stresses, the failure in the joints occurred at the ferritic/austenitic weld interface. The stress-life variation of the joints showed two-slope behavior and the slope change coincided with the occurrence of ferritic/austenitic weld interface cracking. Preferential creep cavitation in the soft intercritical HAZ induced type IV failure, whereas creep cavitation at the interfacial particles induced ferritic/austenitic weld interface cracking. Micromechanisms of the type IV failure and the ferritic/austenitic interface cracking in the dissimilar weld joint of the ferritic steels and relative cracking susceptibility of the joints are discussed based on microstructural investigation, mechanical testing, and finite element analysis (FEA) of the stress state across the joint.

  15. 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.

  16. Susceptibility of irradiated steels to hydrogen embrittlement

    NASA Technical Reports Server (NTRS)

    Rossin, A. D.

    1968-01-01

    Investigation determined whether irradiated pressure-vessel steels 4340 and 212-B are susceptible to hydrogen embrittlement and to catastrophic failure. Hydrogen-charging conditions which completely embrittled 4340 steel had negligible effect on 212-B steel in tensile and delayed-failure tests.

  17. 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 PAGESBeta

    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 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

  18. 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.

  19. Irradiation Creep and Swelling from 400 C to 600 C of the Oxide Dispersion Strengthened Ferritic Alloy MA957

    SciTech Connect

    Toloczko, Mychailo B.; Gelles, David S.; Garner, Francis A.; Kurtz, Richard J.; Abe, Katsunori

    2004-04-24

    Recently, there has been a growing interest in the use of oxide dispersion strengthened (ODS) ferritic steels for fusion reactor applications. As part of an extensive study performed at PNNL on the ODS steel MA957 [1], irradiation creep tests were performed on pressurized tubes made from MA957 by two different methods. The tubes were made either by gun drilling alone or by a combination of rod drawing and gun drilling. The different fabrication methods were explored because ODS steels have been difficult to form. The pressurized tubes were irradiated in the Fast Flux Test Facility (FFTF) to doses ranging from 40 dpa to 110 dpa at temperatures ranging from 400 C to 600 C. The effective stresses resulting from the pressurization of the tubes ranged from 0 MPa to 175 MPa.

  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. Field-dependent neutron depolarization study of the ferrite formation in medium-carbon steels

    SciTech Connect

    Te Velthuis, S.G.E.; Van Dijk, N.H.; Rekveldt, M.T.; Sietsma, J.; Van Der Zwaag, S.

    2000-03-14

    Neutron depolarization experiments have been performed on the ferrite and pearlite phase transformations of the medium-carbon C60 and C35 steels. The interaction of the polarized neutron beam with the ferromagnetic ferrite grains gives information on the mean magnetization and the magnetic correlation length. From these parameters the ferrite fraction and the mean ferrite grain size are determined in situ as a function of time and temperature during the phase transformation. The applied magnetic field was varied periodically in order to record a full hysteresis curve of the magnetization, which gives essential information on the microstructure of the ferromagnetic ferrite grains. The field dependence of the mean particle size during the early stages of the pearlite formation is a strong indication of multi-domain behavior, which is absent in the austenite-ferrite transformation and at the end of the pearlite formation.

  2. Effects of transformed ferrite growth on the tensile fracture characteristics of a dual-phase steel

    NASA Astrophysics Data System (ADS)

    Jeong, W. C.; Kim, C. H.

    1988-02-01

    The effects of transformed ferrite growth on the tensile fracture characteristics of a dual-phase steel were investigated by observing crack initiation, propagation, and fracture behaviors. Crack initiation occurred by decohesion between martensite and ferrite. However, cracks propagated along the ferrite-martensite interface in a high temperature quenched specimen, whereas in specimens quenched from lower temperature cracks propagated into the martensite particle. Tensile fracture behaviors were not strongly influenced by the cooling rate. At both cooling rates of 5.6 and 0.1 °C/sec, specimens quenched from high temperature fractured by partially brittle fracture mode, but fracture mode changed to ductile mode as the quenching temperature decreased. The effect of transformed ferrite on the fracture mode was not substantially different from that of retained ferrite. However, the crack initiation and propagation was influenced by the variation in martensite distribution caused by different growth behavior of transformed ferrite.

  3. 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.

  4. Modelling of Nb influence on phase transformation behaviours from austenite to ferrite in low carbon steels

    NASA Astrophysics Data System (ADS)

    Wang, L.; Parker, S. V.; Rose, A. J.; West, G. D.; Thomson, R. C.

    2016-03-01

    In this paper, a new model has been developed to predict the phase transformation behaviours from austenite to ferrite in Nb-containing low carbon steels. The new model is based on some previous work and incorporates the effects of Nb on phase transformation behaviours, in order to make it applicable for Nb-containing steels. Dissolved Nb atoms segregated at prior austenite grain boundaries increase the critical energy for ferrite nucleation, and thus the ferrite nucleation rate is decreased. Dissolved Nb atoms also apply a solute drag effect to the moving transformation interface, and the ferrite grain growth rate is also decreased. The overall transformation kinetics is then calculated according to the classic Johnson-Mehl-Avrami-Kolmogorov (JMAK) theory. The new model predictions are quite consistent with experimental results for various steels during isothermal transformations or continuous cooling.

  5. 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

  6. 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. PMID:26501086

  7. 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.

  8. A complex carbonitride of niobium and vanadium in 9% Cr ferritic steels

    SciTech Connect

    Tokuno, K.; Hamada, K.; Takeda, T. ); Uemori, R. ); Itoh, K. )

    1991-01-01

    It has been considered that small additions of Nb and V have striking effects on the creep strength of high Cr ferritic steels which are used for elevated temperature services such as boilers, steam generators etc. Although Nb and V are thought to form complex precipitates which may act as obstacles for the dislocation glide, the distribution and morphology of the precipitates have not been clarified yet. Several examples of simple precipitates of V in low alloy steels were only reported. In this paper, the morphology of the complex carbonitride of Nb and V in 9% Cr ferritic steels was investigated and the role of the carbonitride on the creep strength was discussed.

  9. 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.

  10. Microstructural characterization of the HAZ in AISI 444 ferritic stainless steel welds

    SciTech Connect

    Silva, Cleiton C. Farias, Jesualdo P.; Miranda, Helio C.; Guimaraes, Rodrigo F.; Menezes, John W.A.; Neto, Moises A.M.

    2008-05-15

    Ferritic stainless steel is used as a coating for equipment in the petroleum refining industry. Welding is the main manufacturing and maintenance process used. However, little information on the metallurgical alterations caused by welding of these steels is found in the literature, prompting this study. In this study the authors evaluated the HAZ microstructure of AISI 444 ferritic stainless steel welded plates, by scanning electron microscopy (SEM) and X-ray diffraction (XRD). The results indicated that a weld thermal cycle caused microphase precipitation in the HAZ of the ferritic stainless steel. Also needle-like Laves phase precipitation occurred in the HAZ, near the partially-melted zone. Other secondary phases such as chi and sigma were observed, as well as nitride, carbide and carbonitride precipitates.

  11. Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels

    SciTech Connect

    Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki

    2007-07-01

    High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future SFR fuels. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and foreign fast reactors. The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to upgrade the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025. (authors)

  12. 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.

  13. Microstructural characterization of irradiated PWR steels using the atom probe field-ion microscope

    SciTech Connect

    Miller, M.K.; Burke, M.G.

    1987-08-01

    Atom probe field-ion microscopy has been used to characterize the microstructure of a neutron-irradiated A533B pressure vessel steel weld. The atomic spatial resolution of this technique permits a complete structural and chemical description of the ultra-fine features that control the mechanical properties to be made. A variety of fine scale features including roughly spherical copper precipitates and clusters, spherical and rod-shaped molybdenum carbide and disc-shaped molybdenum nitride precipitates were observed to be inhomogeneously distributed in the ferrite. The copper content of the ferrite was substantially reduced from the nominal level. A thin film of molybdenum carbides and nitrides was observed on grain boundaries in addition to a coarse copper-manganese precipitate. Substantial enrichment of manganese and nickel were detected at the copper-manganese precipitate-ferrite interface and this enrichment extended into the ferrite. Enrichment of nickel, manganese and phosphorus were also measured at grain boundaries.

  14. 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.

  15. Toughening mechanisms of a high-strength acicular ferrite steel heavy plate

    NASA Astrophysics Data System (ADS)

    Cao, Zhi-Qiang; Bao, Yan-Ping; Xia, Zheng-Hai; Luo, Deng; Guo, Ai-Min; Wu, Kai-Ming

    2010-10-01

    An ultra-low carbon acicular ferrite steel heavy plate was obtained with an advanced thermo-mechanical control process-relaxed precipitation controlled transformation (TMCP-RPC) at Xiangtan Steel, Valin Group. The heavy plate has a tensile strength of approximately 600 MPa with a lower yield ratio. The impact toughness of the heavy plate achieves 280 J at -40°C. The fine-grained mixed microstructures of the heavy plate mainly consist of acicular ferrite, granular bainite, and polygonal ferrite. The high strength and excellent toughness of the heavy plate are attributed to the formation of acicular ferrite microstructure. The prevention of blocks of martensite/retained austenite (M/A) and the higher cleanness are also responsible for the superior toughness.

  16. Mechanical Properties and Microstructure of Dissimilar Friction Stir Welds of 11Cr-Ferritic/Martensitic Steel to 316 Stainless Steel

    NASA Astrophysics Data System (ADS)

    Sato, Yutaka S.; Kokawa, Hiroyuki; Fujii, Hiromichi T.; Yano, Yasuhide; Sekio, Yoshihiro

    2015-12-01

    Dissimilar joints between ferritic and austenitic steels are of interest for selected applications in next generation fast reactors. In this study, dissimilar friction-stir welding of an 11 pct Cr ferritic/martensitic steel to a 316 austenitic stainless steel was attempted and the mechanical properties and microstructure of the resulting welds were examined. Friction-stir welding produces a stir zone without macroscopic weld-defects, but the two dissimilar steels are not intermixed. The two dissimilar steels are interleaved along a sharp zigzagging interface in the stir zone. During small-sized tensile testing of the stir zone, this sharp interface did not act as a fracture site. Furthermore, the microstructure of the stir zone was refined in both the ferritic/martensitic steel and the 316 stainless steel resulting in improved mechanical properties over the adjacent base material regions. This study demonstrates that friction-stir welding can produce welds between dissimilar steels that contain no macroscopic weld-defects and display suitable mechanical properties.

  17. Intragranular ferrite nucleation in medium-carbon vanadium steels

    SciTech Connect

    Ishikawa, Fusao; Takahashi, Toshihiko ); Ochi, Tatsurou . Muroran R D Lab.)

    1994-05-01

    In this study, the mechanism of intragranular ferrite nucleation is investigated. It is found that intragranular ferrite idiomorphs'' nucleate at vanadium nitrides which precipitate at manganese sulfide particles during cooling in the austenite region. It is observed that intragranular ferrite has the Baker-Nutting orientation relationship with vanadium nitride which precipitated at manganese sulfide. According to classical nucleation theory, the proeutectoid ferrite nucleation rate depends on the following factors: (1) the driving free energy for ferrite nucleation, (2) the diffusivity of carbon atoms in austenite, and (3) the increase in the interfacial energy associated with ferrite nucleation. In the Baker-Nutting orientation relationship, the lattice mismatch across the habit planes is likely to be very small. Depleted zones of solute atoms such as vanadium are assumed to be formed in the austenite matrix around precipitates. The effect of the depleted zones on factors (1) and (2) is estimated thermodynamically and it is proved that those effects are negligibly small. Thus, the authors conclude that the most important factor in nucleation kinetics of intragranular ferrite is the formation of precipitates which can develop coherent, low energy interfaces with ferrite.

  18. 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.

  19. 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.

  20. Irradiation creep of various ferritic alloys irradiated at ˜400°C in the PFR and FFTF reactors

    NASA Astrophysics Data System (ADS)

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

    1998-10-01

    Irradiation creep of three ferritic alloys at ˜400 ∘C has been studied. Specimens were in the form of pressurized tubes. In a joint US/UK creep study, two identical sets of creep specimens constructed from one heat of HT9 were irradiated in fast reactors, one in the Prototypic Fast Reactor (PFR) and the other in the Fast Flux Test Facility (FFTF). The specimens in PFR were irradiated to a dose of ˜50 dpa, whereas the specimens in FFTF were irradiated to a dose of 165 dpa. The observed swelling and creep behavior were very different in the two reactors. Creep specimens constructed from D57, a developmental alloy ferritic alloy, were also irradiated in PFR to a dose of ˜50 dpa. Creep behavior typical of previous studies on ferritic alloys was observed. Finally, creep specimens constructed from MA957, a Y 2O 3 dispersion-hardened ferritic alloy, were irradiated in FFTF to a dose of ˜110 dpa. This alloy exhibited a large amount of densification, and the creep behavior was different than observed in more conventional ferritic or ferritic-martensitic alloys.

  1. 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. PMID:17059551

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

    DOEpatents

    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.

  3. 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.

  4. 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.

  5. Quantitative metallographic method for determining delta ferrite content in austenitic stainless steels. Final report

    SciTech Connect

    Pressly, G.A.

    1986-01-01

    Delta ferrite is a magnetic form of iron and has a body centered cubic crystal structure. It is often present as a nonequilibrium phase in austenitic stainless steel welds, castings, and wrought materials. The ferrite content of austenitic stainless steel can directly affect its properties, especially weldability and formability. Therefore, it is highly desirable to be able to predict and/or measure the ferrite content accurately. Current magnetic ferrite measuring methods are not applicable when test materials are geometrically small (less than 2.54 mm thick and 6.35 mm wide). Therefore, a standard metallographic test method STM 00107-A was established to determine delta ferrite content in small weldments and base metals of austenitic stainless steel. This standard test method (STM 00107-A) was then performed on several exemplary metallographic specimens to illustrate its capabilities and applications. The results from the exemplary tests were compared and contrasted to metallographic manual point count measurements, Ferritescope measurements, and predicted values calculated from chemical analyses. By utilizing the manual metallographic point count data, an accuracy of +-16% and a precision of +-0.77% were determined for the standard test method. The comparison of Ferritescope data to standard test method revealed that the results obtained by the two methods are close at low (0 to 3%) ferrite contents and Ferritscope results are substantially greater at higher (6 to 10%) ferrite contents. The standard test method data compiled from the exemplary weld specimens was noted to be very similar to the predicted values calculated from chemical analyses. It was also shown that because the standard test method utilizes optics the morphology of the delta ferrite particles can be determined. This type of determination is possible only with metallographic methods.

  6. Deuterium and helium trapping at TiC particles in ferritic steel

    NASA Astrophysics Data System (ADS)

    Spitznagel, J. A.; Brenner, S. S.; Miller, M. K.; Choyke, W. J.

    1984-05-01

    First wall and blanket materials in Tokamak machines must accommodate increasing concentrations of helium and hydrogen isotopes. Alloy design principles point to the efficacy of trapping He and hydrogen at finely dispersed precipitates to minimize their impact on mechanical properties. Titanium carbide particles are known to trap He effectively in austenitic stainless steel. Less is known about TiC as a trap for helium and hydrogen isotopes in ferritic steels. This paper demonstrates the feasibility of directly measuring the trapping of helium and deuterium at TiC-ferrite interfaces using atom probe field ion microscopy.

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

    NASA Astrophysics Data System (ADS)

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

    2013-12-01

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

  8. Use of ferritic steels in breeder reactors worldwide

    SciTech Connect

    Patriarca, P.

    1983-01-01

    The performance of LMFBR reactor steam generator materials is reviewed. Tensile properties of stainless steel-304, stainless steel-316, chromium-molybdenum steels, and Incoloy 800H are presented for elevated temperatures.

  9. 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.

  10. The fatigue crack initiation at the interface between matrix and {delta}-ferrite in 304L stainless steel

    SciTech Connect

    Rho, B.S.; Hong, H.U.; Nam, S.W.

    1998-10-13

    It is well known that austenitic stainless steels have good mechanical properties and good corrosion resistance at high temperatures and are widely used in high temperature application. However, representative 304L stainless steel among austenitic stainless steels has the undesirable {delta}-ferrite in {gamma} matrix unavoidably because of the limitation of the manufacturing process. While large amounts of {delta}-ferrite in the austenitic stainless steels can give rise to a decrease in the hot workability, the absence of {delta}-ferrite in 304L stainless steel can be the cause of longitudinal facial crack and shortness of continuous cast slab. However, there are few reported papers related with the effect of {delta}-ferrite nucleating the initial crack at the interface between matrix and {delta}-ferrite on fatigue properties at high temperature. In the present work, a comparison of fatigue life with the amount of {delta}-ferrite was examined and to find out the mechanism of crack initiation caused by {delta}-ferrite, dislocation behavior near the interface between {delta}-ferrite and matrix during fatigue testing was analyzed. To analyze the dislocation character near the interface between the matrix and {delta}-ferrite during a low cycle fatigue test, trace analysis was applied. Using Burgers vector and dislocation line direction, calculated by trace analysis, it was possible to obtain some characteristic of dislocation behaviors near the interface.

  11. Transformation Characteristics of Ferrite/Carbide Aggregate in Continuously Cooled, Low Carbon-Manganese Steels

    NASA Astrophysics Data System (ADS)

    Di Martino, S. F.; Thewlis, G.

    2014-02-01

    Transformation characteristics and morphological features of ferrite/carbide aggregate (FCA) in low carbon-manganese steels have been investigated. Work shows that FCA has neither the lamellae structure of pearlite nor the lath structure of bainite and martensite. It consists of a fine dispersion of cementite particles in a smooth ferrite matrix. Carbide morphologies range from arrays of globular particles or short fibers to extended, branched, and densely interconnected fibers. Work demonstrates that FCA forms over similar cooling rate ranges to Widmanstätten ferrite. Rapid transformation of both phases occurs at temperatures between 798 K and 973 K (525 °C and 700 °C). FCA reaction is not simultaneous with Widmanstätten ferrite but occurs at temperatures intermediate between Widmanstätten ferrite and bainite. Austenite carbon content calculations verify that cementite precipitation is thermodynamically possible at FCA reaction temperatures without bainite formation. The pattern of precipitation is confirmed to be discontinuous. CCT diagrams have been constructed that incorporate FCA. At low steel manganese content, Widmanstätten ferrite and bainite bay sizes are significantly reduced so that large amounts of FCA are formed over a wide range of cooling rates.

  12. Modeling the Ferrite-Austenite Transformation in the Heat-Affected Zone of Stainless Steel Welds

    SciTech Connect

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

    1997-12-01

    The diffusion-controlled ferrite-austenite transformation in stainless steel welds was modeled. An implicit finite-difference analysis that considers multi-component diffusion was used. The model was applied to the Fe-Cr-Ni system to investigate the ferrite- austenite transformation in the heat-affected zone of stainless steel weld metal. The transformation was followed as a function of time as the heat-affected zone was subjected to thermal cycles comparable to those experienced during gas-tungsten arc welding. The results showed that the transformation behavior and the final microstructural state are very sensitive to the maximum temperature that is experienced by the heat-affected zone. For high maximum exposure temperatures ({approximately} 1300{degree} C), the ferrite formation that occurs at the highest temperatures is not completely offset by the reverse ferrite dissolution at lower temperatures. As a result, for high temperature exposures there is a net increase in the amount of ferrite in the microstructure. It was also found that if compositional gradients are present in the initial ferrite and austenite phases, the extent of the transformation is impacted.

  13. Comparison of the mechanical strength properties of several high-chromium ferritic steels

    SciTech Connect

    Booker, M.K.; Sikka, V.K.; Booker, B.L.P.

    1981-01-01

    A modified 9 Cr-1 Mo ferritic steel has been selected as an alternative material for breeder reactors. Different 9 Cr-1 Mo steels are already being used commercially in UK and USA and a 9 Cr-2 Mo steel (EM12) is being used commercially in France. The 12% Cr steel alloy HT9 is also often recommended for high-temperature service. Creep-rupture data for all six seels were analyzed to yield rupture life as a function of stress, temperature, and lot-to-lot variations. Yield and tensile strength data for the three 9 Cr-1 Mo materials were also examined. All results were compared with Type 304 stainless steel, and the tensile and creep properties of the modified and British 9 Cr-1 Mo materials were used to calculate allowable stress values S/sub 0/ per Section VIII, Division 1 and S/sub m/ per code Case N-47 to section III of the ASME Boiler and Pressure Vessel Code. these values were compared with code listings for American commercial 9 Cr-1 Mo steel, 2 1/4 Cr-1 Mo steel, and Type 304 stainless steel. The conclusion is made that the modified 9 Cr-1 Mo steel displays tensile and creep strengths superior to those of the other ferritic materials examined and is at least comparable to Type 304 stainless steel from room temperature to about 625/sup 0/C. 31 figures.

  14. Development of Bimodal Ferrite-Grain Structures in Low-Carbon Steel Using Rapid Intercritical Annealing

    NASA Astrophysics Data System (ADS)

    Karmakar, A.; Karani, A.; Patra, S.; Chakrabarti, Debalay

    2013-05-01

    Mixed ferrite grain structures, which have fine- and coarse-grain regions and showing "bimodal" grain size distributions, have been produced by rapid intercritical annealing of warm-rolled (or cold-rolled) samples. Microstructural changes have been analyzed using dilatometric studies, size prediction of transformed and recrystallized grains, and microtexture measurements. Fine austenite grains (<5 μm) developed during rapid annealing and transformed into fine-ferrite grains (2 to 4 μm) after cooling. Coarse-ferrite grains (28 to 42 μm) resulted from the recrystallization and growth of deformed ferrite. The effect of heating rate on microstructural morphologies during intercritical annealing has also been studied. A slow rate of heating (30 K/s) developed a uniform distribution of fine-ferrite grains and austenitic islands, while rapid heating (300 K/s) generated coarse blocks of austenite, elongated along the prior-pearlitic regions, in the ferrite matrix. As expected, bimodal ferrite grain structures or fine-scale dual-phase structures showed superior combination of tensile strength and ductility, compared to the ultrafine-grained steels.

  15. 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.

  16. Ni-Cu-Zn Ferrite Powder Prepared from Steel Pickled Liquor and Electroplating Waste Solutions

    NASA Astrophysics Data System (ADS)

    Liu, Chung-Wen; Fu, Yen-Pei; Lin, Cheng-Hsiung

    2007-03-01

    In this study, we propose a new method of synthesizing Ni-Cu-Zn ferrite powder using steel pickled liquor and electroplating waste solutions as starting materials. It was found that the Ni-Cu-Zn ferrite powder prepared by a hydrothermal process from the waste solutions shows the formation of cubic ferrite with a saturation magnetization (Ms) of 31.5 emu/g and an intrinsic coercive force (Hci) of 19.3 Oe. Upon annealing at 750 °C for 2 h, the saturation magnetization increases to 52.6 emu/g and the intrinsic coercive force reaches 42.8 Oe. This useful method can promote the recycling of industrial waste solution and contribute to the preservation of the earth. Moreover, this method decreases the manufacturing cost in the treatment of the industrial waste solution for electroplating and steel industries.

  17. Effect of Mn Addition on Microstructural Modification and Cracking Behavior of Ferritic Light-Weight Steels

    NASA Astrophysics Data System (ADS)

    Sohn, Seok Su; Lee, Byeong-Joo; Lee, Sunghak; Kwak, Jai-Hyun

    2014-11-01

    In the present study, effects of Mn addition on cracking phenomenon occurring during cold rolling of ferritic light-weight steels were clarified in relation to microstructural modification involving κ-carbide, austenite, and martensite. Four steels were fabricated by varying Mn contents of 3 to 12 wt pct, and edge areas of steel sheets containing 6 to 9 wt pct Mn were cracked during the cold rolling. The steels were basically composed of ferrite and austenite in a band shape, but a considerable amount of κ-carbide or martensite existed in the steels containing 3 to 6 wt pct Mn. Microstructural observation of the deformed region of fractured tensile specimens revealed that cracks which were initiated at ferrite/martensite interfacial κ-carbides readily propagated along ferrite/martensite interfaces or into martensite areas in the steel containing 6 wt pct Mn, thereby leading to the center or edge cracking during the cold rolling. In the steel containing 9 wt pct Mn, edge cracks were found in the final stage of cold rolling because of the formation of martensite by the strain-induced austenite to martensite transformation, whereas they were hardly formed in the steel containing 12 wt pct Mn. To prevent or minimize the cracking, it was recommended that the formation of martensite during the cooling from the hot rolling temperature or during the cold rolling should be suppressed, which could be achieved by the enhancement of thermal or mechanical stability of austenite with decreasing austenite grain size or increasing contents of austenite stabilizers.

  18. Interim storage of sodium in ferritic steel tanks at ambient temperature

    SciTech Connect

    Blackburn, L.D.

    1994-09-30

    Sodium tanks originally fabricated for elevated temperature service in the Clinch River Breeder Reactor Plant (CRBRP) will be used to store sodium removed from the Fast Flux Test Facility (FFTF) in the Sodium Storage Facility (SSF) at ambient temperature. This report presents an engineering review to confirm that protection against brittle fracture of the ferritic steel tanks is adequate for the intended service.

  19. TEM characterization of dislocation loops in irradiated bcc Fe-based steels

    SciTech Connect

    Yao, Bo; Edwards, Danny J.; Kurtz, Richard J.

    2012-12-08

    In this study, we describe a methodology to examine dislocation loops in irradiated steels based on a combination of crystallographic information and g*b invisibility criteria. Dislocation loops in transmission electron microscope (TEM) images can be conveniently analyzed using this method. Through this analysis approach, dislocation loops in reduced activation ferritic/martensitic (RAFM) steels irradiated at 400 *C have been examined. The predominant types of loops found in irradiated RAFM steels were h100i{200} and 1/2h111i 111. The size, density, and density anisotropy of these two types of dislocation loops were quantified. It was observed that the h100i{200} loop density is more than twice that of 1/2h111i{111} loops. A large density anisotropy of h100i{200} loops was identified.

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

    NASA Astrophysics Data System (ADS)

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

    2016-04-01

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

  1. Effect of specimen size and nickel content on the impact properties of 12 Cr-1 MoVW ferritic steel

    SciTech Connect

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

    1983-01-01

    The ferritic steel 12 Cr-1 MoVW is a candidate material for use in the first wall of magnetic fusion reactors. One of the primary concerns of materials in this application is service-induced embrittlement from aging and irradiation. Unirradiated Charpy impact data have been developed on three typical heats of 12 Cr-1 MoVW steel and on heats that have been modified with nickel and chromium additions for subsequent simulation of helium effects of irradiation. The ductile-to-brittle transition temperature and the upper-shelf energy were reduced by nickel additions. The addition of nickel while simultaneously maintaining a constant net chromium equivalent caused the transition temperature to increase. The use of a subsize specimen is mandated by the small volume and high gamma heating in high-flux research reactors used for very high-fluence irradiation experiments. Therefore, an understanding of the behavior of the subsize specimen is important in predicting corresponding irradiation-induced transition temperature shifts and upper-shelf drops of full-size Charpy specimens. Data are reported from subsize Charpy impact specimens, and the full- and subsize specimens are compared. The effect of specimen size on the upper-shelf energy of this material can be reduced better by volume normalization than by area normalization.

  2. Connection between micro and macro hardness pearlitic-ferritic steel

    NASA Astrophysics Data System (ADS)

    Duka, Edlira; Oettel, Heinrich; Dilo, Teuta

    2012-09-01

    Many physical and mechanical properties of materials are closely related to their microstructure, technologies to control the microstructure of materials have been well developed to obtain suitable properties. We measured the volume fraction of perlite and ferrite, micro Vickers hardness in pearlite and ferrite and macro hardness using different sample with different carbon content. The volume fraction of pearlite increases by increasing carbon content. By increasing carbon content, micro and macro hardness increase. We can conclude that for those conditional the mixing rule can't be use.

  3. Irradiation hardening of reduced activation martensitic steels

    NASA Astrophysics Data System (ADS)

    Kimura, A.; Morimura, T.; Narui, M.; Matsui, H.

    1996-10-01

    Irradiation response on the tensile properties of 9Cr2W steels has been investigated following FFTF/MOTA irradiations at temperatures between 646 and 873 K up to doses between 10 and 59 dpa. The largest irradiation hardening accompanied by the largest decrease in the elongation is observed for the specimens irradiated at 646 K at doses between 10 and 15 dpa. The irradiation hardening appears to saturate at a dose of around 10 dpa at the irradiation temperature. No hardening but softening was observed in the specimens irradiated at above 703 K to doses of 40 and 59 dpa. Microstructural observation by transmission electron microscope (TEM) revealed that the dislocation loops with the a<100> type Burgers vector and small precipitates which were identified to be M 6C type carbides existed after the irradiation at below 703 K. As for the void formation, the average size of voids increased with increasing irradiation temperature from 646 to 703 K. No voids were observed above 703 K. Irradiation softening was attributed to the enhanced recovery of martensitic structure under the irradiation. Post-irradiation annealing resulted in hardening by the annealing at 673 K and softening by the annealing at 873 K.

  4. 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.

  5. Role of a complex carbonitride of niobium and vanadium in creep strength of 9% Cr ferritic steels

    SciTech Connect

    Tokuno, K.; Hamada, K.; Takeda, T. ); Uemori, R. ); Itoh, K. )

    1991-08-01

    Complex precipitates of Nb and V in ferritic steels have been considered to act as strong obstacles against dislocation glide. In the previous report, the authors investigated a complex precipitate of Nb and V in 9% Cr ferritic steel and showed that plate-like V-nitrides were formed adhering to spherical Nb (C,N) during tempering. These peculiar shaped precipitates are likely to trap dislocations during high temperature deformation such as creep. In this paper, the effect of the sizes of the complex carbonitrides on the creep strength of 9% Cr ferritic steels was investigated. The sizes of the carbonitrides were varied by changing the amount of N.

  6. Evaluation on Fatigue Crack Propagation of Reduced Activation Ferritic Steel (JLF-1) at High Temperature

    NASA Astrophysics Data System (ADS)

    Yoon, Han Ki; Kim, Sa Woong; Lee, Sang Pill; Katoh, Yutai; Kohyama, Akira

    Recently, reduced activation ferritic/martensitic steel, vanadium alloy and SiC/SiC composite are embossed for nuclear fusion reactor in accordance with the coolant. Especially, reduced activation ferritic/martensitic steel is very suitable material for nuclear fusion reactor, because it has low coefficient of thermal expansion and excellent heat conductivity. The objective of this study is to investigate fatigue crack propagation behavior in the Reduced Activation Ferritic Steel (JLF-1). The fatigue crack propagation behavior of the JLF-1 steel was investigated by the constant-amplitude loading test for the stress ratios R = 0.1, 0.3 and 0.5 respectively. The fatigue crack growth tests carried out at room temperature and 400°C for base metal and weld metal. The effects of stress ratio, test temperature, specimen size and TIG welding on the fatigue crack propagation behaviors for JLF-1 steel were discussed within the Paris law. Particularly, the fatigue crack propagation rate of a weld metal was similar to that of base metal at the stress ratio of 0.3. Also, the fatigue crack propagation rate of a half size specimen was similar to that of a full size specimen at the stress ratios of 0.1, 0.3 and 0.5 respectively. From this result, we can recognize that the fatigue crack propagation behavior of this material can be evaluated by using the half size specimens.

  7. 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.

  8. Residual Ferrite and Relationship Between Composition and Microstructure in High-Nitrogen Austenitic Stainless Steels

    NASA Astrophysics Data System (ADS)

    Wang, Qingchuan; Ren, Yibin; Yao, Chunfa; Yang, Ke; Misra, R. D. K.

    2015-12-01

    A series of high-nitrogen stainless steels (HNS) containing δ-ferrite, which often retained in HNS, were studied to establish the relationship between composition and microstructure. Both ferrite and nitrogen depletions were found in the center regions of cast ingots, and the depletion of nitrogen in that area was found to be the main reason for the existence of δ-ferrite. Because of the existence of heterogeneity, the variation of microstructure with nitrogen content was detected. Hence, the critical contents of nitrogen (CCN) for the fully austenitic HNS were obtained. Then the effects of elements such as N, Cr, Mn, and Mo on austenite stability were investigated via thermodynamic calculations. The CCN of HNS alloys were also obtained by calculations. Comparing the CCN obtained from experiment and calculation, it was found that the forged microstructure of the HNS was close to the thermodynamic equilibrium. To elucidate the above relationship, by regression analysis using calculated thermodynamic data, nitrogen equivalent and a new constitution diagram were proposed. The constitution diagram accurately distinguishes the austenitic single-phase region and the austenite + ferrite dual-phase region. The nitrogen equivalent and the new constitution diagram can be used for alloying design and microstructural prediction in HNS. According to the nitrogen equivalent, the ferrite stabilizing ability of Mo is weaker than Cr, and with Mn content increases, Mn behaves as a weak austenite stabilizer first and then as a ferrite stabilizer.

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

    DOE PAGESBeta

    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,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

  10. Nature of anisotropy of impact toughness of structural steels with ferrite-pearlite structure

    NASA Astrophysics Data System (ADS)

    Goritskii, V. M.; Shneyderov, G. R.; Lushkin, M. A.

    2013-10-01

    The anisotropy of the impact toughness of low-alloy steels of various compositions and purities with a ferrite-pearlite structure has been investigated using samples of type 11 according to the Russian Standard GOST 9454-78. It has been established that the anisotropy coefficient of the impact toughness depends on the anisotropy coefficient of the work of crack propagation and is independent of the degree of striation of the ferrite-pearlite structure and the work for nucleation of the ductile crack.

  11. Long-Term SOFC Stability with Coated Ferritic Stainless Steel Interconnect

    SciTech Connect

    Simner, Steve P.; Anderson, Michael D.; Xia, Gordon; Yang, Z Gary; Stevenson, Jeffry W.

    2005-01-25

    This study details long-term performance data for anode-supported thin-film YSZ-based SOFCs utilizing a ferritic stainless steel cathode current collector (Crofer22 APU) coated with a protective (Mn,Co)3O4 spinel to prevent Cr volatilization. Two standard cathode compositions, La(Sr)FeO3 and La(Sr)MnO3, were considered. The coating proved effective in blocking Cr migration, which resulted in long-term stability of the manganite cathode. In contrast the ferrite cathode indicated degradation that could not be attributed to Cr poisoning.

  12. Decomposition Kinetics of Ferrite in Isothermally Aged SAF 2507-Type Duplex Stainless Steel

    NASA Astrophysics Data System (ADS)

    Berecz, Tibor; Fazakas, Éva; Mészáros, István; Sajó, István

    2015-12-01

    Decomposition of the ferritic phase is studied in isothermally aged SAF 2507 superduplex stainless steel (SDSS) by means of different examination methods. The ferritic phase ( δ) undergoes an eutectoid transformation into secondary austenite ( γ 2) and σ-phase between 650 and 1000 °C. Samples were treated at 900 °C because the incubation time of this transformation is the shortest at this temperature. In order to follow the microstructural changes, x-ray diffraction analysis (XRD), automated electron backscatter diffraction (EBSD), applied magnetic investigation [vibrating sample magnetometer (VSM)], micro-hardness tests, and differential thermal analysis (DTA) were used. The results of XRD and EBSD methods for phase quantification showed nearly the same amounts for all three phases. The results of applied magnetic investigation for the fraction of ferritic phase were also in good agreement with the corresponding results of XRD and EBSD methods. Decomposition of ferrite is similarly well-traceable on EBSD phase maps where the coherent ferritic areas gradually broke into pieces with increasing time of heat treatment. According to the EBSD measurements the σ-phase grains appeared and started to grow after 2 min aging time in the ferritic-austenitic matrix, usually on the boundaries of ferritic and austenitic grains. After 15 min treating time, the microstructure consisted of mainly σ- and austenitic (primary and secondary) phases with negligible amount of ferrite. Chemical composition of the σ-phase was measured by energy-dispersive x-ray spectroscopy (EDS) at different aging times. Activation energies of σ-phase precipitation and α'-phase formation were determined by the Kissinger plot, through DTA measurements; they are 243 and 261 kJ/mol, respectively. Using the results of phase quantifications, the Johnson-Mehl-Avrami equation was fitted.

  13. Proton irradiation creep of FM steel T91

    NASA Astrophysics Data System (ADS)

    Xu, Cheng; Was, Gary S.

    2015-04-01

    Ferritic-martensitic (FM) steel T91 was subjected to irradiation with 3 MeV protons while under load at stresses of 100-200 MPa, temperatures between 400 °C and 500 °C, and dose rates between 1.4 × 10-6 dpa/s and 5 × 10-6 dpa/s to a total dose of less than 1 dpa. Creep behavior was analyzed for parametric dependencies. The temperature dependence was found to be negligible between 400 °C and 500 °C, and the dose rate dependence was observed to be linear. Creep rate was proportional to stress at low stress values and varied with stress to the power 14 above 160 MPa. The large stress exponent of the proton irradiation creep experiments under high stress suggested that dislocation glide was driving both thermal and irradiation creep. Microstructure observations of anisotropic dislocation loops also contributed to the total creep strain. After subtracting the power law creep and anisotropic dislocation loop contributions, the remaining creep strain was accounted for by dislocation climb enabled by stress induced preferential absorption (SIPA) and preferential dislocation glide (PAG).

  14. The Role of Alloying Elements in Nanostructured Ferritic Steels

    SciTech Connect

    Miller, Michael K; Parish, Chad M

    2010-01-01

    The roles of the alloying elements in three nanostructured ferritic alloys (14YWT, MA957 and Eurofer 97) have been established through the characterisation of the microstructure by atom probe tomography and spectrum imaging in a transmission electron microscope. Cr, W, Mo, Ti and Y were found in the ferrite matrix and contributed to solid solution hardening. Ti, Y, C, O and N were found in high number densities of precipitates and nanoclusters both in the grain interior and on grain boundaries and thereby contributed to precipitation hardening. Cr, W and Mo were enriched at the intraparticle regions of the grain boundaries. The solute segregation and precipitation pinned the grain boundaries and contributed to the excellent creep properties of the alloys.

  15. 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.

  16. Influence of Chemical Composition and Heat Treatment Condition on Impact Toughness of 15Cr Ferritic Creep Resistant Steel

    NASA Astrophysics Data System (ADS)

    Toda, Yoshiaki; Tohyama, Hideaki; Kushima, Hideaki; Kimura, Kazuhiro; Abe, Fujio

    Influences of chemical compositions, heat treatment and microstructure on impact toughness of 15Cr ferritic steel have been investigated. Charpy impact values of the furnace cooled steels were lower than 15J/cm2 at room temperature independent of chemical compositions. Drastic improvement in impact toughness has been attained by controlling the carbon and nitrogen contents, by the addition of nickel and by the increase in cooling rate after annealing. However, the effect of nickel on impact toughness strongly depends on carbon and nitrogen contents. Improvement in impact toughness of the 15Cr ferritic steel has not been explained by individual microstructural factors of grain size, distribution of precipitates, volume fraction of martensitic phase. It has been supposed that the increase in Charpy impact toughness of the 15Cr ferritic steel was attained by improvement in toughness of ferrite matrix itself.

  17. 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.

  18. (Mn,Co)(3)O-4 Spinel Coatings on Ferritic Stainless Steels for SOFC Interconnect Applications

    SciTech Connect

    Yang, Z Gary; Xia, Gordon; Li, Xiaohong S.; Stevenson, Jeffry W.

    2007-11-01

    (Mn,Co)3O4 spinel with a nominal composition of Mn1.5Co1.5O4 demonstrates excellent electrical conductivity, satisfactory thermal and structural stability, as well as good thermal expansion match to ferritic stainless steel interconnects. A slurry-coating technique was developed for fabricating the spinel coatings onto the steel interconnects. Thermally grown layers of Mn1.5Co1.5O4 not only significantly decreased the contact resistance between a LSF cathode and stainless steel interconnect, but also acted as a mass barrier to inhibit scale growth on the stainless steel and to prevent Cr outward migration through the coating. The level of improvement in electrical performance and oxidation resistance (i.e. the scale growth rate) was dependent on the ferritic substrate composition. For E-brite and Crofer22 APU, with a relatively high Cr concentration (27wt% and 23%, respectively) and negligible Si, the reduction of contact ASR and scale growth on the ferritic substrates was significant. In comparison, limited improvement was achieved by application of the Mn1.5Co1.5O4 spinel coating on AISI430, which contains only 17% Cr and a higher amount of residual Si.

  19. 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.

  20. 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.

  1. 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.

  2. Microstructural Effects on Fracture Behavior of Ferritic and Martensitic Structural Steels

    NASA Astrophysics Data System (ADS)

    Ibrahim, Omyma H.; Elshazly, Ezzat S.

    2013-02-01

    The effect of microstructure on fracture behavior of 1Cr-0.5Mo and 9Cr-1Mo structural steels was evaluated. 1Cr-0.5Mo steel is used in steam pipes and superheater tubes of power stations. Its microstructure is typically comprised of bainite in a pre-eutectoid ferrite matrix with an average grain size of 10 μm. 9Cr-1Mo steel was developed for applications in steam power stations and as a candidate structural material for first-wall and blanket components of future fusion reactors. Its microstructure consisted of a fully martensitic structure with a prior austenite grain size of 25 μm. The fracture properties were measured using instrumented impact testing at temperatures between -196 and 300 °C. The total impact fracture energy, the crack initiation and propagation energy, the dynamic yield strength, the brittleness temperature, and the cleavage fracture stress were measured. The bainitic-ferritic alloy steel exhibited much higher resistance to ductile fracture at high test temperatures, while its resistance to brittle fracture at low test temperatures was reduced compared to that of the fully martensitic alloy steel. The results were discussed in terms of the chemical composition and microstructure of the two steel types.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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. PMID:26942534

  8. Kinetics of isochronal austenization in modified high Cr ferritic heat-resistant steel

    NASA Astrophysics Data System (ADS)

    Liu, Chenxi; Liu, Yongchang; Zhang, Dantian; Yan, Zesheng

    2011-12-01

    Employment of high Cr ferritic steels as a main structural material is considered as a way to achieve economical competitiveness of main steam pipe and nuclear reactors in power plants. Differential dilatometry and microstructure observation were employed to investigate the isochronal austenitic transformation of the modified high Cr ferritic steel. The kinetics of the isochronal austenitic transformation were described by a phase-transformation model involving site saturation (pre-existing nuclei), diffusion-controlled growth, and incorporating an impingement correction. The experimental results and kinetic analysis indicate that an increase of the heating rate promotes the diffusion-controlled austenitic transformation. The dissolving degree of precipitates during the austenization process affects the activation energy for diffusion and the undissolved precipitates lead to an increase of the onset temperature of the subsequent martensite transformation upon cooling.

  9. 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.

  10. 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.

  11. Ripple reduction and surface coating tests with ferritic steel on JFT-2M

    NASA Astrophysics Data System (ADS)

    Tsuzuki, K.; Sato, M.; Kawashima, H.; Miura, Y.; Kimura, H.; Abe, T.; Uehara, K.; Ogawa, T.; Akiyama, T.; Shibata, T.; Yamamoto, M.; Koike, T.

    2000-12-01

    Applicability of the low-activation ferritic steel (F82H), which is one of the candidate materials for next generation fusion devices, has been tested in JFT-2M. Ferritic steel boards (FB) were installed between toroidal field coils (TFC) and the vacuum vessel at all toroidal sections. The experiment and the calculation show that the ripple amplitude decreases from 2.2% to 1.1% by the FB installation at R=1.6 m. The ripple reduction results in the reduction of fast ion loss and in the shift of the ripple loss to the outer region. No undesirable effect to the energy confinement and the plasma control was observed. The FB will be installed inside the vacuum vessel in the near future. In preparation, surface coating tests, mainly concerning with an in situ boron coating, are carried out.

  12. Analysis of Creep Rupture Behavior of Cr-Mo Ferritic Steels in the Presence of Notch

    NASA Astrophysics Data System (ADS)

    Goyal, Sunil; Laha, K.; Das, C. R.; Mathew, M. D.

    2015-01-01

    Effect of notch on creep rupture behavior of 2.25Cr-1Mo, 9Cr-1Mo, and modified 9Cr-1Mo ferritic steels has been assessed. Creep tests were carried out on smooth and notched specimens of the steels in the stress ranging 90 to 300 MPa at 873 K (600 °C). Creep rupture lives of the steels increased in the presence of notch over those of smooth specimens, thus exhibiting notch strengthening. The strengthening was comparable for the 9Cr-1Mo and 2.25Cr-1Mo steels and appreciably more in modified 9Cr-1Mo steel. The strengthening effect was found to decrease with the decrease in applied stress and increase in rupture life for all the steels. The presence of notch decreased the creep rupture ductility of the steels significantly and the 2.25Cr-1Mo steel suffered more reduction than in the other two 9Cr-steels. Finite element analysis of stress distribution across the notch was carried out to understand the notch strengthening and its variation in the steels. The variation in fracture appearance has also been corroborated based on finite element analysis. Reduction in von-Mises stress across the notch throat plane resulted in strengthening in the steels. Higher reduction in von-Mises stress in modified 9Cr-1Mo steel than that in 2.25Cr-1Mo and 9Cr-1Mo steels induced more strengthening in modified 9Cr-1Mo steel under multiaxial state of stress.

  13. Irradiation creep and microstructural changes of ODS steels of different Cr-contents during helium implantation under stress

    NASA Astrophysics Data System (ADS)

    Chen, J.; Jung, P.; Henry, J.; de Carlan, Y.; Sauvage, T.; Duval, F.; Barthe, M. F.; Hoffelner, W.

    2013-06-01

    Irradiation creep and microstructural changes of two ferritic ODS steels with 12% and 14% Cr have been studied by homogeneously implantation with helium under uniaxial tensile stresses from 40 to 300 MPa. The maximum dose was about 1.2 dpa (5000 appm-He) with displacement damage rates of 1 × 10-5 dpa/s at a temperature of 300 °C. Irradiation creep compliances were measured to be 4.0 × 10-6 dpa-1 MPa-1 and 10 × 10-6 dpa-1 MPa-1 for 12 and 14Cr ODS, respectively. Subsequently, microstructural evolution was studied in detail by TEM observations, showing dislocation loops and bubbles distributed homogenously in the matrix. Some bubbles were attached to ODS particles. Finally, the effects of Cr content on irradiation creep and microstructural changes are discussed, including earlier results of a 19Cr ODS and a PM2000 ferritic steel. Irradiation creep rates of both 12Cr and 14Cr-ODS ferritic steels a temperature of 300 °C show linear stress dependence up to 300 MPa at. Irradiation creep rate per dose rate and stress at a temperature of 300 °C amounts to 4.0 × 10-6 dpa-1 MPa-1 and 10 × 10-6 dpa-1 MPa-1 for 12Cr- and 14Cr-ODS, respectively. Irradiation creep properties are remarkably insensitive to Cr content, grain size and dispersoid size. Dislocation loops and helium bubbles are distributed homogenously in the matrix. In the case of high density fine dispersoids, most bubbles are attached to ODS particles. This may suppress loop formation as well as growth of bubbles, thereby increasing the resistance of ODS ferritic steels against helium embrittlement.

  14. Analysis of chemical changes and microstructure characterization during deformation in ferritic stainless steel.

    PubMed

    Núñez, Andrés; Llovet, Xavier; Almagro, Juan F

    2013-08-01

    Uni- and biaxial tension deformation tests, with different degrees of deformation, have been done on AISI 430 (EN 1.4016) ferritic stainless steel samples, which had both different chemical compositions and had undergone different annealing treatments. The initial and deformed materials were characterized by using electron backscatter diffraction and backscatter electron imaging in a scanning electron microscope together with electron probe microanalysis. The correlation observed among the chemical compositions, annealing treatment, and strain level obtained after deformation is discussed. PMID:23628319

  15. 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.

  16. Heavy-Section Steel Irradiation Program

    SciTech Connect

    Rosseel, T.M.

    2000-04-01

    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. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative 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.

  17. Dilution and Ferrite Number Prediction in Pulsed Current Cladding of Super-Duplex Stainless Steel Using RSM

    NASA Astrophysics Data System (ADS)

    Eghlimi, Abbas; Shamanian, Morteza; Raeissi, Keyvan

    2013-12-01

    Super-duplex stainless steels have an excellent combination of mechanical properties and corrosion resistance at relatively low temperatures and can be used as a coating to improve the corrosion and wear resistance of low carbon and low alloy steels. Such coatings can be produced using weld cladding. In this study, pulsed current gas tungsten arc cladding process was utilized to deposit super-duplex stainless steel on high strength low alloy steel substrates. In such claddings, it is essential to understand how the dilution affects the composition and ferrite number of super-duplex stainless steel layer in order to be able to estimate its corrosion resistance and mechanical properties. In the current study, the effect of pulsed current gas tungsten arc cladding process parameters on the dilution and ferrite number of super-duplex stainless steel clad layer was investigated by applying response surface methodology. The validity of the proposed models was investigated by using quadratic regression models and analysis of variance. The results showed an inverse relationship between dilution and ferrite number. They also showed that increasing the heat input decreases the ferrite number. The proposed mathematical models are useful for predicting and controlling the ferrite number within an acceptable range for super-duplex stainless steel cladding.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Effect of Weld Intercooling Temperature on the Structure and Impact Strength of Ferritic-Martensitic Steels

    SciTech Connect

    T.C. Totemeier; J.A. Simpson; H. Tian

    2006-06-01

    The effect of inadequate weld intercooling (cooling prior to post-weld heat treatment) on the structure and impact properties of 9Cr-1MoVNb (ASME Grade 91) and 12Cr-1Mo-WV (Type 422 stainless) steels was studied. A range of weld intercooling conditions were simulated by air cooling the two steels from the standard 1050°C normalization temperature to temperatures ranging from 250 to 450°C for Grade 91 and 100 to 300°C for Type 422, and then immediately tempering at 760°C for two hours. For Grade 91 steel, austenite retained at the intercooling temperature transformed to ferrite during tempering; final microstructures were mixtures of ferrite and tempered martensite. For Type 422 steel, austenite retained at the intercooling temperature was stable in the tempering condition and formed martensite upon cooling to room temperature; final microstructures were mixtures of tempered and untempered martensite. Hardness and impact properties of the two steels reflected the changes in microstructure with intercooling temperature.

  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. 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.

  7. Cleavage-quasi cleavage in ferritic and martensitic steels

    SciTech Connect

    Odette, G.R.; Edsinger, K.V.; Lucas, G.E.

    1997-12-31

    Confocal microscopy-fracture reconstruction and SEM were used to characterize the sequence-of-events leading to cleavage in a low alloy pressure vessel steel and two 8--12 Cr martensitic steels as a function of temperature. While differences between the steels were observed, they shared some common characteristics that differ from the conventional view of cleavage. Most notably cleavage does not occur as a single weakest link event; rather it is the consequence of a critical condition when a previously nucleated dispersion of microcracks suddenly coalesce to form a large, rapidly propagating macroscopic crack. It is argued that the critical event can be treated as a bridging instability. The stabilizing effect of the ductile ligaments separating the cleavage facets increases with increasing temperature. Indeed, even in the ductile tearing regime cleavage facets form a significant fraction of nuclei for larger microvoids.

  8. Dynamic strain aging behavior of modified 9Cr-1Mo and reduced activation ferritic martensitic steels under low cycle fatigue

    NASA Astrophysics Data System (ADS)

    Mariappan, K.; Shankar, Vani; Sandhya, R.; Prasad Reddy, G. V.; Mathew, M. D.

    2013-04-01

    Influence of temperature and strain rate on low cycle fatigue (LCF) behavior of modified 9Cr-1Mo ferritic martensitic steel and 1.4W-0.06Ta reduced activation ferritic martensitic (RAFM) steel in normalized and tempered conditions was studied. Total strain controlled LCF tests between 300 and 873 K on modified 9Cr-1Mo steel and RAFM steel and at various strain rates on modified 9Cr-1Mo steel were performed at total strain amplitude of ±0.6%. Both the steels showed continuous cyclic softening at all temperatures. Whereas manifestations of dynamic strain aging (DSA) were observed in both the steels which decreased fatigue life at intermediate temperatures, at higher temperatures, oxidation played a crucial role in decreasing fatigue life.

  9. Sensitization of 21% Cr Ferritic Stainless Steel Weld Joints Fabricated With/Without Austenitic Steel Foil as Interlayer

    NASA Astrophysics Data System (ADS)

    Wu, Wenyong; Hu, Shengsun; Shen, Junqi; Ma, Li; Han, Jian

    2015-04-01

    The effects of sensitization heat treatment on the microstructure and electrochemical behavior of 21% Cr ferritic stainless steel weld joints with or without 309L austenite stainless steel as an interlayer were investigated. The joints were processed by pulsed gas tungsten arc welding. With the interlayer, grains in weld bead were refined, and almost fully ferrite. When the joints with the interlayer were maintained at 500 °C for 1 and 4 h, no microstructure changes occurred, whereas Widmanstatten austenite and needle-like austenite formed in the weld bead after sensitization at 815 °C for 1 h. In general, sensitization treatment worsens the corrosion resistance of welds, but the resistance of samples with the 4-h treatment at 500 °C recovered in part compared to those subjected to sensitization at 500 °C for 1 h. This could be due to Cr diffusion from the ferrite that heals the chromium-depletion zone along the grain boundary. However, an increase in temperature does not have the same effect. The corrosion morphology of samples in the weld bead is different from those in base metal after heat treatment at 500 °C for 1 h; in base metal, pitting corrosion occurs, whereas grain boundary corrosion occurs in the weld bead. Corrosion morphology is closely associated with precipitation and segregation along the grain boundary.

  10. Microstructure and properties of pipeline steel with a ferrite/martensite dual-phase microstructure

    SciTech Connect

    Li Rutao Zuo Xiurong Hu Yueyue Wang Zhenwei Hu, Dingxu

    2011-08-15

    In order to satisfy the transportation of the crude oil and gas in severe environmental conditions, a ferrite/martensite dual-phase pipeline steel has been developed. After a forming process and double submerged arc welding, the microstructure of the base metal, heat affected zone and weld metal was characterized using scanning electron microscopy and transmission electron microscopy. The pipe showed good deformability and an excellent combination of high strength and toughness, which is suitable for a pipeline subjected to the progressive and abrupt ground movement. The base metal having a ferrite/martensite dual-phase microstructure exhibited excellent mechanical properties in terms of uniform elongation of 7.5%, yield ratio of 0.78, strain hardening exponent of 0.145, an impact energy of 286 J at - 10 deg. C and a shear area of 98% at 0 deg. C in the drop weight tear test. The tensile strength and impact energy of the weld metal didn't significantly reduce, because of the intragranularly nucleated acicular ferrites microstructure, leading to high strength and toughness in weld metal. The heat affected zone contained complete quenching zone and incomplete quenching zone, which exhibited excellent low temperature toughness of 239 J at - 10 deg. C. - Research Highlights: {yields}The pipe with ferrite/martensite microstructure shows high deformability. {yields}The base metal of the pipe consists of ferrite and martensite. {yields}Heat affected zone shows excellent low temperature toughness. {yields}Weld metal mainly consists of intragranularly nucleated acicular ferrites. {yields}Weld metal shows excellent low temperature toughness and high strength.