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Sample records for ferritic-martensitic steel t91

  1. Formation of alumina-aluminide coatings on ferritic-martensitic T91 steel

    Directory of Open Access Journals (Sweden)

    Choudhary R.K.

    2014-01-01

    Full Text Available In this work, alumina-aluminide coatings were formed on ferritic-martensitic T91 steel substrate. First, coatings of aluminum were deposited electrochemically on T91 steel in a room temperature AlCl3-1-ethyl-3-methyl imidazolium chloride ionic liquid, then the obtained coating was subjected to a two stage heat treatment procedure consisting of prolonged heat treatment of the sample in vacuum at 300 ○C followed by oxidative heat treatment in air at 650 ○C for 16 hours. X-ray diffraction measurement of the oxidatively heat treated samples indicated formation of Fe-Al and Cr-Al intermetallics and presence of amorphous alumina. Energy dispersive X-ray spectroscopy measurement confirmed 50 wt- % O in the oxidized coating. Microscratch adhesion test conducted on alumina-aluminide coating formed on T91 steel substrate showed no major adhesive detachment up to 20 N loads. However, adhesive failure was observed at a few discrete points on the coating along the scratch track.

  2. A roadmap for tailoring the strength and ductility of ferritic/martensitic T91 steel via thermo-mechanical treatment

    International Nuclear Information System (INIS)

    Song, M.; Sun, C.; Fan, Z.; Chen, Y.; Zhu, R.; Yu, K.Y.; Hartwig, K.T.; Wang, H.; Zhang, X.

    2016-01-01

    Ferritic/martensitic (F/M) steels with high strength and excellent ductility are important candidate materials for the life extension of the current nuclear reactors and the design of next generation nuclear reactors. Recent studies show that equal channel angular extrusion (ECAE) was able to improve mechanical strength of ferritic T91 steels moderately. Here, we examine several strategies to further enhance the mechanical strength of T91 while maintaining its ductility. Certain thermo-mechanical treatment (TMT) processes enabled by combinations of ECAE, water quench, and tempering may lead to “ductile martensite” with exceptionally high strength in T91 steel. The evolution of microstructures and mechanical properties of T91 steel were investigated in detail, and transition carbides were identified in water quenched T91 steel. This study provides guidelines for tailoring the microstructure and mechanical properties of T91 steel via ECAE enabled TMT for an improved combination of strength and ductility.

  3. Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang; Miao, Yinbin; Li, Meimei; Kirk, Marquis A.; Maloy, Stuart A.; Stubbins, James F.

    2017-07-01

    In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 x 10(15) ions/cm(2) at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 x 1015 ions/cm(2), about 51% of the loops were a(0)/2 < 111 > type for the 673 K irradiation, and the dominant loop type was a(0)< 100 > for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100-150 MPa. (C) 2017 Elsevier B.V. All rights reserved.

  4. Ferritic/martensitic steels: Promises and problems

    International Nuclear Information System (INIS)

    Klueh, R.L.; Ehrlich, K.; Abe, F.

    1992-01-01

    Ferritic/martensitic steels are candidate structural materials for fusion reactors because of their higher swelling resistance, higher thermal conductivity, lower thermal expansion, and better liquid-metal compatibility than austenitic steels. Irradiation effects will ultimately determine the applicability of these steels, and the effects of irradiation on microstructure and swelling, and on the tensile, fatigue, and impact properties of the ferritic/martensitic steels are discussed. Most irradiation studies have been carried out in fast reactors, where little transmutation helium forms. Helium has been shown to enhance swelling and affect tensile and fracture behavior, making helium a critical issue, since high helium concentrations will be generated in conjunction with displacement damage in a fusion reactor. These issues are reviewed to evaluate the status of ferritic/martensitic steels and to assess the research required to insure that such steels are viable candidates for fusion applications

  5. Influence of microstructural development during annealing at 780oC on creep resistance of ferritic-martensitic T91 (9%Cr-1%Mo-V-Nb) steel

    International Nuclear Information System (INIS)

    De Cicco, H; Zavaleta Gutierrez, N; Marrero, J; Luppo, M.I; Danon, C.A

    2006-01-01

    Due to its good properties of creep resistance, toughness and rust resistance, martensitic-ferritic 9%Cr-1%Mo steels are widely used for the production of heating plant components, boilers, heat exchangers, piping and tubing, etc. The effectiveness in steels of MX carbonitrides such as (Nb,V) (C,N) on improving creep resistance at high temperatures is well known. Controlling the behavior of the MX phases to precipitation, during annealing, is essential for obtaining a stable microstructure that can resist high temperatures. This study investigates the relationship between creep resistance and the microstructural changes that occur at different annealing times at a temperature of 780 o C -used industrially during the production and post-welding- in T91 steel. Creep trials were carried out at 600 o C and 190 MPa, and the samples were characterized using optic microscopy (OM), inductively coupled plasma-atomic emission spectrometry (ICP-AES) and transmission electron microscopy (TEM), the latter including a facility for energy dispersive spectroscopy (EDS). Based on its morphological characteristics, the MX precipitates are classified into three types, types I, II and III. Tempering time at 780 o C has been found to be one of the factors that determines which MX is dominant in the annealed steel. The presence of type MX-III, formed by the secondary precipitation of a VN particle adhering to a NbX, commonly called 'wing', seems to favor creep resistance in these steels. This type of of precipitate, then, fills an effective role in the anchoring of dislocations during creep (cw)

  6. Future directions for ferritic/martensitic steels for nuclear applications

    International Nuclear Information System (INIS)

    Klueh, R.L.; Swindeman, R.W.

    2000-01-01

    High-chromium (7-12% Cr) ferritic/martensitic steels are being considered for nuclear applications for both fission and fusion reactors. Conventional 9-12Cr Cr-Mo steels were the first candidates for these applications. For fusion reactors, reduced-activation steels were developed that were patterned on the conventional steels but with molybdenum replaced by tungsten and niobium replaced by tantalum. Both the conventional and reduced-activation steels are considered to have an upper operating temperature limit of about 550degC. For improved reactor efficiency, higher operating temperatures are required. For ferritic/martensitic steels that could meet such requirements, oxide dispersion-strengthened (ODS) steels are being considered. In this paper, the ferritic/martensitic steels that are candidate steels for nuclear applications will be reviewed, the prospect for ODS steel development and the development of steels produced by conventional processes will be discussed. (author)

  7. Radiation-induced segregation and phase stability in ferritic-martensitic alloy T 91

    Energy Technology Data Exchange (ETDEWEB)

    Wharry, Janelle P.; Jiao Zhijie; Shankar, Vani [University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109-2104 (United States); Busby, Jeremy T. [Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37831 (United States); Was, Gary S., E-mail: gsw@umich.edu [University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109-2104 (United States)

    2011-10-01

    Radiation-induced segregation in ferritic-martensitic alloy T 91 was studied to understand the behavior of solutes as a function of dose and temperature. Irradiations were conducted using 2 MeV protons to doses of 1, 3, 7 and 10 dpa at 400 deg. C. Radiation-induced segregation at prior austenite grain boundaries was measured, and various features of the irradiated microstructure were characterized, including grain boundary carbide coverage, the dislocation microstructure, radiation-induced precipitation and irradiation hardening. Results showed that Cr, Ni and Si segregate to prior austenite grain boundaries at low dose, but segregation ceases and redistribution occurs above 3 dpa. Grain boundary carbide coverage mirrors radiation-induced segregation. Irradiation induces formation of Ni-Si-Mn and Cu-rich precipitates that account for the majority of irradiation hardening. Radiation-induced segregation behavior is likely linked to the evolution of the precipitate and dislocation microstructures.

  8. Radiation induced microstructural evolution in ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Kohno, Y.; Kohyama, A.; Asakura, K.; Gelles, D.S.

    1993-01-01

    R and D of ferritic/martensitic steels as structural materials for fusion reactor is one of the most important issues of fusion technology. The efforts to characterize microstructural evolution under irradiation in the conventional Fe-Cr-Mo steels as well as newly developed Fe-Cr-Mn or Fe-Cr-W low activation ferritic/ martensitic steels have been continued. This paper provides some of the recent results of heavy irradiation effects on the microstructural evolution of ferritic/martensitic steels neutron irradiated in the FFTF/MOTA (Fast Flux Test Facility/Materials Open Test Assembly). Materials examined are Fe-10Cr-2Mo dual phase steel (JFMS: Japanese Ferritic/Martensitic Steel), Fe-12Cr-XMn-1Mo manganese stabilized martensitic steels and Fe-8Cr-2W Tungsten stabilized low activation martensitic steel (F82H). JFMS showed excellent void swelling resistance similar to 12Cr martensitic steel such as HT-9, while the manganese stabilized steels and F82H showed less void swelling resistance with small amount of void swelling at 640-700 K (F82H: 0.14% at 678 K). As for irradiation response of precipitate behavior, significant formation of intermetallic χ phase was observed in the manganese stabilized steels along grain boundaries which is though to cause mechanical property degradation. On the other hand, precipitates identified were the same type as those in unirradiated condition in F82H with no recognition of irradiation induced precipitates, which suggested satisfactory mechanical properties of F82H after the irradiation. (author)

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

    Science.gov (United States)

    2012-03-01

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

  10. Water corrosion resistance of ODS ferritic-martensitic steel tubes

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasuji

    2008-01-01

    Oxide dispersion strengthened (ODS) ferritic-martensitic steels have superior radiation resistance; it is possible to achieve a service temperature of up to around 973 K because of their superior creep strength. These advantages of ODS steels facilities their application to long-life cladding tubes in advanced fast reactor fuel elements. In addition to neutron radiation resistance, sufficient general corrosion resistance to maintain the strength of the cladding, and the stress corrosion cracking (SCC) resistance for spent-fuel-pool cooling systems and high-temperature oxidation for the fuel-clad chemical interaction (FCCI) of ODS ferritic steel are required. Although the addition of Cr to ODS is effective in preventing water corrosion and high-temperature oxidation, an excessively high amount of Cr leads to embrittlement due to the formation of a Cr-rich α' precipitate. The Cr content in 9Cr-ODS martensite and 12Cr-ODS ferrite, the ODS steels developed by the Japan Atomic Energy Agency (JAEA), is controlled. In a previous paper, it has been demonstrated that the resistances of 9Cr- and 12Cr-ODS ferritic-martensitic steels for high-temperature oxidation are superior to those of conventional 12Cr ferritic steel. However, the water corrosion data of ODS ferritic-martensitic steels are very limited. In this study, a water corrosion test was conducted on ODS steels in consideration of the spent-fuel-pool cooling condition, and the results were compared with those of conventional austenitic stainless steel and ferritic-martensitic stainless steel. (author)

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

    Science.gov (United States)

    2013-06-01

    report of FSW on a ferritic- martensitic stainless steel is the work of Chung, which applied this approach to a dissimilar weld between F82H (ferritic... martensitic ) and SUS304 (austenitic stainless ) [43]. 7 D. CORROSION OF FERRITIC/ MARTENSITIC STEELS IN HIGH TEMPERATURE MOLTEN SALT COOLANTS In...Philadelphia, PA, 1992, pp. 1267–1286, March 1990. [15] S. Rosenwasser, ―The application of martensitic stainless steels in a lifelong fusion first wall

  12. Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 deg. C

    International Nuclear Information System (INIS)

    Matijasevic, M.; Lucon, E.; Almazouzi, A.

    2008-01-01

    High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 deg. C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 deg. C and 300 deg. C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 deg. C (up to 2.6 dpa), and tested between -170 deg. C and 300 deg. C. Irradiation effects at lower irradiation temperatures are more significant

  13. Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

    Science.gov (United States)

    Matijasevic, M.; Lucon, E.; Almazouzi, A.

    2008-06-01

    High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between -160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between -170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.

  14. Influence of tempering on mechanical properties of ferritic martensitic steels

    International Nuclear Information System (INIS)

    Chun, Y. B.; Han, C. H.; Choi, B. K.; Lee, D. W.; Kim, T. K.; Jeong, Y. H.; Cho, S.

    2012-01-01

    In the mid-1980s research programs for development of low activation materials began. This is based on the US Nuclear Regulatory Commission Guidelines (10CFR part 61) that were developed to reduce long-lived radioactive isotopes, which allows nuclear reactor waste to be disposed of by shallow land burial when removed from service. Development of low activation materials is also key issue in nuclear fusion systems, as the structural components can became radioactive due to nuclear transmutation caused by exposure to high dose neutron irradiation. Reduced-activation ferritic martensitic (RAFM) steels have been developed in the leading countries in nuclear fusion technology, and are now being considered as primary candidate material for the test blanket module (TBM) in the international thermonuclear experiment reactor (ITER). RAFM steels developed so far (e.g., EUROFER 97 and F82H) meet the requirement for structural application in the ITER. However, if such alloys are used in the DEMO or commercial fusion reactor is still unclear, as the reactors are designed to operate under much severe conditions (i.e., higher outlet coolant temperature and neutron fluences). Such harsh operating conditions lead to development of RAFM steels with better creep and irradiation resistances. Mechanical properties of RAFM steels are strongly affected by microstructural features including the distribution, size and type of precipitates, dislocation density and grain size. For a given composition, such microstructural characteristics are determined mainly by thermo-mechanical process employed to fabricate the final product, and accordingly a final heat treatment, i.e., tempering is the key step to control the microstructure and mechanical properties. In the present work, we investigated mechanical properties of the RAFM steels with a particular attention being paid to effects of tempering on impact and creep properties

  15. Deformation twinning in irradiated ferritic/martensitic steels

    Science.gov (United States)

    Wang, K.; Dai, Y.; Spätig, P.

    2018-04-01

    Two different ferritic/martensitic steels were tensile tested to gain insight into the mechanisms of embrittlement induced by the combined effects of displacement damage and helium after proton/neutron irradiation in SINQ, the Swiss spallation neutron source. The irradiation conditions were in the range: 15.8-19.8 dpa (displacement per atom) with 1370-1750 appm He at 245-300 °C. All the samples fractured in brittle mode with intergranular or cleavage fracture surfaces when tested at room temperature (RT) or 300 °C. After tensile test, transmission electron microscopy (TEM) was employed to investigate the deformation microstructures. TEM-lamella samples were extracted directly below the intergranular fracture surfaces or cleavage surfaces by using the focused ion beam technique. Deformation twinning was observed in irradiated specimens at high irradiation dose. Only twins with {112} plane were observed in all of the samples. The average thickness of twins is about 40 nm. Twins initiated at the fracture surface, became gradually thinner with distance away from the fracture surface and finally stopped in the matrix. Novel features such as twin-precipitate interactions, twin-grain boundary and/or twin-lath boundary interactions were observed. Twinning bands were seen to be arrested by grain boundaries or large precipitates, but could penetrate martensitic lath boundaries. Unlike the case of defect free channels, small defect-clusters, dislocation loops and dense small helium bubbles were observed inside twins.

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

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1996-01-01

    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

  17. Corrosion-resistant coating technique for oxide-dispersion-strengthened ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami

    2014-01-01

    Oxide-dispersion-strengthened (ODS) steels are attractive materials for application as fuel cladding in fast reactors and first-wall material of fusion blanket. Recent studies have focused more on high-chromium ferritic (12-18 wt% Cr) ODS steels with attractive corrosion resistance properties. However, they have poor material workability, require complicated heat treatments for recrystallization, and possess anisotropic microstructures and mechanical properties. On the other hand, low-chromium ferritic/martensitic (8-9 wt% Cr) ODS steels have no such limitations; nonetheless, they have poor corrosion resistance properties. In our work, we developed a corrosion-resistant coating technique for a low-chromium ferritic/martensitic ODS steel. The ODS steel was coated with the 304 or 430 stainless steel, which has better corrosion resistances than the low-chromium ferritic/martensitic ODS steels. The 304 or 430 stainless steel was coated by changing the canning material from mild steel to stainless steel in the conventional material processing procedure for ODS steels. Microstructural observations and micro-hardness tests proved that the stainless steels were successfully coated without causing a deterioration in the mechanical property of the low-chromium ferritic/martensitic ODS steel. (author)

  18. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

  19. Numerical simulation of transformation-induced microscopic residual stress in ferrite-martensite lamellar steel

    International Nuclear Information System (INIS)

    Mikami, Y; Inao, A; Mochizuki, M; Toyoda, M

    2009-01-01

    The effect of transformation-induced microscopic residual stress on fatigue crack propagation behavior of ferrite-martensite lamellar steel was discussed. Fatigue tests of prestrained and non-prestrained specimens were performed. Inflections and branches at ferrite-martensite boundaries were observed in the non-prestrained specimens. On the other hand, less inflections and branches were found in the prestrained specimens. The experimental results showed that the transformation induced microscopic residual stress has influence on the fatigue crack propagation behavior. To estimate the microscopic residual, a numerical simulation method for the calculation of microscopic residual stress stress induced by martensitic transformation was performed. The simulation showed that compressive residual stress was generated in martensite layer, and the result agree with the experimental result that inflections and branches were observed at ferrite-martensite boundaries.

  20. Estimation of Oxidation Kinetics and Oxide Scale Void Position of Ferritic-Martensitic Steels in Supercritical Water

    Directory of Open Access Journals (Sweden)

    Li Sun

    2017-01-01

    Full Text Available Exfoliation of oxide scales from high-temperature heating surfaces of power boilers threatened the safety of supercritical power generating units. According to available space model, the oxidation kinetics of two ferritic-martensitic steels are developed to predict in supercritical water at 400°C, 500°C, and 600°C. The iron diffusion coefficients in magnetite and Fe-Cr spinel are extrapolated from studies of Backhaus and Töpfer. According to Fe-Cr-O ternary phase diagram, oxygen partial pressure at the steel/Fe-Cr spinel oxide interface is determined. The oxygen partial pressure at the magnetite/supercritical water interface meets the equivalent oxygen partial pressure when system equilibrium has been attained. The relative error between calculated values and experimental values is analyzed and the reasons of error are suggested. The research results show that the results of simulation at 600°C are approximately close to experimental results. The iron diffusion coefficient is discontinuous in the duplex scale of two ferritic-martensitic steels. The simulation results of thicknesses of the oxide scale on tubes (T91 of final superheater of a 600 MW supercritical boiler are compared with field measurement data and calculation results by Adrian’s method. The calculated void positions of oxide scales are in good agreement with a cross-sectional SEM image of the oxide layers.

  1. Effect of microstructure on the fracture toughness of ferrite-martensite-bainite steels

    International Nuclear Information System (INIS)

    Byun, Thak Sang; Kim, In Sup

    1988-01-01

    The effect of microstructure on the fracture toughness of ferrite-martensite -bainite steels was investigated with Fe-0.11C-1.64Mn-0.78Si composition. One inch compact tension specimens (1T-CTSs) were machined from hot rolled plates. The microstructure of ferrite-martensite-bainite was introduced to the specimens by the heat treatment of intercritical annealing at 800deg C and isothermal holding at 350deg C. Holding at 350deg C increased volume fraction of bainite, while decreased that of martensite, and refined martensite particles. Single specimen unloading compliance method was used in fracture test to obtain J-resistance (J-R) curve and to determine the fracture toughness(J IC ). Introduction of bainite to the ferrite-martensite steel improved the fracture toughness due to the deformation of bainite which relaxed the stress concentration on the interface of ferrite and martensite. Observation of fracto-graphs through the scanning electron microscope(SEM) identified the fracture mechanism of ferrite-martensite-bainite steels as dimple nucleation and crack growth by decohesion of ferrite matrix and second phase particles and by microvoid coales cence. (Author)

  2. Microstructure and tensile properties of high strength duplex ferrite-martensite (DFM) steels

    International Nuclear Information System (INIS)

    Chakraborti, P.C.; Mitra, M.K.

    2007-01-01

    Duplex ferrite-martensite (DFM) steels containing 38-80% martensite of varying morphologies were developed by batch intercritical annealing of a commercial variety vanadium bearing 0.2% C-Mn steel at different temperatures. Microstructures before intercritical annealing were found to control the morphological distribution of the phase constituents of the developed DFM steels. Tensile test results revealed best strength-ductility combination for finely distributed lamellar ferrite-martensite phase aggregate containing ∼60% martensite developed from a prior martensitic structure. Taking consideration of the modified law of mechanical mixture the experimental tensile strength data of the developed DFM steels has been formulated with some success and very good estimation for tensile strengths of pure ferrite and low carbon martensite has been made from tensile strength data of DFM steels

  3. Development of oxide dispersion strengthened 9Cr ferritic-martensitic steel clad tube for fast reactor

    International Nuclear Information System (INIS)

    Laha, K.; Saroja, S.; Mathew, M.D.; Jayakumar, T.; Vijay, R.; Venugopal Reddy, A.; Lakshminarayana, B.; Kapoor, Komal; Jha, S.K.; Tonpe, S.S.

    2012-01-01

    One of the key issues in the economical operation of FBR is to achieve high burn-up of fuel (200-250 GWd/t) which considerably reduces the fuel cycle cost. This imposes stringent requirements of void swelling resistance upto 200 dpa for the core structural materials. Presently used alloy 09 (a modified austenitic stainless steel, 15Cr-15Ni-Ti) for PFBR has void swelling limit less than 150 dpa. Because of the inherent void swelling resistance, 9-12Cr steels ferritic/martensitic steels are qualified for irradiation upto 200 dpa but their low creep strength at temperatures above 600 deg C restricts their application as a clad material. Oxide dispersion strengthening is found to be promising means of extending the creep resistance of ferritic/martensitic steels beyond 650 deg C without sacrificing the inherent advantages of high thermal conductivity and low swelling of ferritic steels

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Lab., TN (United States)

    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.

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

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1997-01-01

    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

  7. Irradiation damage of ferritic/martensitic steels: Fusion program data applied to a spallation neutron source

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1997-01-01

    Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs

  8. Chemical compatibility study of lithium titanate with Indian reduced activation ferritic martensitic steel

    International Nuclear Information System (INIS)

    Sonak, Sagar; Jain, Uttam; Haldar, Rumu; Kumar, Sanjay

    2015-01-01

    Highlights: • Chemical compatibility between Li_2TiO_3 and Indian RAFM steel has been studied at ITER operating temperature. • The lithium titanate chemically reacted with ferritic martensitic steel to form a brittle and non-adherent oxide layer. • The layer grew in a parabolic manner as a function of heating time. • Diffusion of oxygen (from Li_2TiO_3) appears to be controlling the oxide layer. - Abstract: Chemical compatibility between lithium titanate and Indian reduced activation ferritic-martensitic steel (In-RAFMS) was studied for the first time under ITER operating temperature. Lithium titanate required for the study was synthesized in-house. Coupons of In-RAFMS were packed inside lithium titanate powder and heated at 550 °C up to 900 h under inert argon atmosphere. The lithium titanate chemically reacted with ferritic martensitic steel to form a brittle and non-adherent oxide layer. The layer grew in a parabolic manner as a function of heating time. Microstructural and phase evolution of this oxide layer was studied using XRD, SEM and EPMA. Iron and chromium enriched zones were found within the oxide layer. Diffusion of oxygen (from Li_2TiO_3) appears to be controlling the oxide layer.

  9. Corrosion of Ferritic-Martensitic steels in high temperature water: A literature Review

    International Nuclear Information System (INIS)

    Fernandez, P.; Lapena, J.; Blazquez, F.

    2001-01-01

    Available literature concerning corrosion of high-chromium ferritic/martensitic steel in high temperature water as reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, environmentally assisted cracking (EAC) including stress corrosion cracking (SCC), corrosion fatigue and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS). Are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. (Author)

  10. Material physical properties of 11Cr-ferritic/martensitic steel (PNC-FMS) wrapper tube materials

    International Nuclear Information System (INIS)

    Yano, Yasuhide; Kaito, Takeji; Ohtsuka, Satoshi; Tanno, Takashi; Uwaba, Tomoyuki; Koyama, Shinichi

    2012-09-01

    It is necessary to develop core materials for fast reactors in order to achieve high-burnup. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, various physical properties of PNC-FMS wrapper materials were measured and equations and future standard measurement technique of physical properties for the design and evaluation were conducted. (author)

  11. Recent progress of R and D activities on reduced activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Q., E-mail: qunying.huang@fds.org.cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, P.O. Box 1135, Hefei, Anhui 230031 (China); Baluc, N. [CRPP-EPFL, ODGA C110 5232 Villigen PSI (Switzerland); Dai, Y. [LNM, PSI, 5232 Villigen PSI (Switzerland); Jitsukawa, S. [JAEA, 2-4 Shirakata, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan); Kimura, A. [IAE, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Konys, J. [KIT, P.O. Box 3640, 76021 Karlsruhe (Germany); Kurtz, R.J. [PNNL, Richland, WA 99352 (United States); Lindau, R. [KIT, P.O. Box 3640, 76021 Karlsruhe (Germany); Muroga, T. [NIFS, Oroshi, Toki, Gifu 509-5292 (Japan); Odette, G.R. [UCSB, Santa Barbara, CA (United States); Raj, B. [IGCAR, Kalpakkam 603 102 (India); Stoller, R.E.; Tan, L. [ORNL, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Tanigawa, H. [JAEA, Naka, Ibaraki 311-0193 (Japan); Tavassoli, A.-A.F. [DMN/Dir, DEN, CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Yamamoto, T. [UCSB, Santa Barbara, CA (United States); Wan, F. [DMPC, USTB, Beijing 100083 (China); Wu, Y. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, P.O. Box 1135, Hefei, Anhui 230031 (China)

    2013-11-15

    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 test blanket modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties 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 to different TBM and DEMO options.

  12. Influence of liquid lead and lead-bismuth eutectic on tensile, fatigue and creep properties of ferritic/martensitic and austenitic steels for transmutation systems

    Energy Technology Data Exchange (ETDEWEB)

    Gorse, D., E-mail: dominique.gorse-pomonti@polytechnique.edu [CNRS-LSI, Ecole Polytechnique, route de Saclay, 91128, Palaiseau Cedex (France); Auger, T. [CNRS-MSSMAT, Ecole Centrale Paris, Grande Voie des Vignes, 92290, Chatenay-Malabry Cedex (France); Vogt, J.-B.; Serre, I. [CNRS-LMPGM, 59655, Villeneuve d' Ascq Cedex (France); Weisenburger, A. [ForschungszentrumKarlsruheGmbH, P.O. Box 3640, 76021 Karlsruhe (Germany); Gessi, A.; Agostini, P. [ENEA, CR Brasimone, 40032 Camugnano, Bologna (Italy); Fazio, C. [ForschungszentrumKarlsruheGmbH, P.O. Box 3640, 76021 Karlsruhe (Germany); Hojna, A.; Di Gabriele, F. [Ustav jaderneho vyzkumu Rez a.s., Husinec 130, Rez 25068 (Czech Republic); Van Den Bosch, J.; Coen, G.; Almazouzi, A. [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, 2400 Mol (Belgium); Serrano, M. [CIEMAT, Avda. Complutense, 22, 28040 Madrid (Spain)

    2011-08-31

    In this paper, the tensile, fatigue and creep properties of the Ferritic/Martensitic (F/M) steel T91 and of the Austenitic Stainless (AS) Steel 316L in lead-bismuth eutectic (LBE) or lead, obtained in the different organizations participating to the EUROTRANS-DEMETRA project are reviewed. The results show a remarkable consistency, referring to the variety of metallurgical and surface state conditions studied. Liquid Metal Embrittlement (LME) effects are shown, remarkable on heat-treated hardened T91 and also on corroded T91 after long-term exposure to low oxygen containing Liquid Metal (LM), but hardly visible on passive or oxidized smooth T91 specimens. For T91, the ductility trough was estimated, starting just above the melting point of the embrittler (T{sub M,E} = 123.5 deg. C for LBE, 327 deg. C for lead) with the ductility recovery found at 425 deg. C. LME effects are weaker on 316L AS steel. Liquid Metal Assisted Creep (LMAC) effects are reported for the T91/LBE system at 550 deg. C, and for the T91/lead system at 525 deg. C. Today, if the study of the LME effects on T91 and 316L in LBE or lead can be considered well documented, in contrast, complementary investigations are necessary in order to quantify the LMAC effects in these systems, and determine rigorously the threshold creep conditions.

  13. Corrosion of austenitic and ferritic-martensitic steels exposed to supercritical carbon dioxide

    International Nuclear Information System (INIS)

    Tan, L.; Anderson, M.; Taylor, D.; Allen, T.R.

    2011-01-01

    Highlights: → Oxidation is the primary corrosion phenomenon for the steels exposed to S-CO 2 . → The austenitic steels showed significantly better corrosion resistance than the ferritic-martensitic steels. → Alloying elements (e.g., Mo and Al) showed distinct effects on oxidation behavior. - Abstract: Supercritical carbon dioxide (S-CO 2 ) is a potential coolant for advanced nuclear reactors. The corrosion behavior of austenitic steels (alloys 800H and AL-6XN) and ferritic-martensitic (FM) steels (F91 and HCM12A) exposed to S-CO 2 at 650 deg. C and 20.7 MPa is presented in this work. Oxidation was identified as the primary corrosion phenomenon. Alloy 800H had oxidation resistance superior to AL-6XN. The FM steels were less corrosion resistant than the austenitic steels, which developed thick oxide scales that tended to exfoliate. Detailed microstructure characterization suggests the effect of alloying elements such as Al, Mo, Cr, and Ni on the oxidation of the steels.

  14. Report of IEA workshop on reduced activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    IEA Workshop on Reduced Activation Ferritic/Martensitic Steels under implementing agreement for program of research and development on fusion materials was held at Tokyo Yayoi Kaikan and JAERI headquarter on November 2-3, 2000. The objective of this workshop was a review of the fusion material development programs, the progress of the collaboration and the irradiation effects studies on RAF/M steels in the collaborating parties (Europe, Russia the United States, and Japan). Moreover, the development of plans for future collaboration was discussed. The present report contains viewgraphs presented at the workshop. (author)

  15. Mechanical properties of friction stir welded 11Cr-ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Yano, Y.; Sato, Y.S.; Sekio, Y.; Ohtsuka, S.; Kaito, T.; Ogawa, R.; Kokawa, H.

    2013-01-01

    Friction stir welding was applied to the wrapper tube materials, 11Cr-ferritic/martensitic steel, designed for fast reactors and defect-free welds were successfully produced. The mechanical and microstructural properties of the friction stir welded steel were subsequently investigated. The hardness values of the stir zone were approximately 550 Hv (5.4 GPa) with minimal dependence on the rotational speed, even though they were much higher than those of the base material. However, tensile strengths and elongations of the stir zones were high at 298 K, compared to those of the base material. The excellent tensile properties are attributable to the fine grain formation during friction stir welding

  16. Diffusion bonding of 9Cr ODS ferritic/martensitic steel with a phase transformation

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon, E-mail: shnoh@kaeri.re.kr [Nuclear Materials Division, Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Kimura, Akihiko [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto (Japan); Kim, Tae Kyu [Nuclear Materials Division, Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon (Korea, Republic of)

    2014-10-15

    Highlights: • Diffusion bonding was employed to join 9Cr oxide dispersion strengthened ferritic/martensitic steel under uniaxial hydrostatic pressure, and the microstructure and tensile properties of the joints were investigated. • ODS steel was successfully diffusion bonded at an austenization temperature to migrate a residual diffusion bonding interface. • The tensile properties of the joint region were comparable with that of the base metal with a ductile fracture occurred far from the bonding interface. • It is considered that diffusion bonding with a phase transformation can be a very useful joining method for fabricating components in next-generation nuclear systems using 9Cr ODS ferritic/martensitic steel. - Abstract: Diffusion bonding was employed to join oxide-dispersion-strengthened ferritic/martensitic steel under uniaxial hydrostatic pressure using a high vacuum hot press, and the microstructure and tensile properties of the joints were investigated. 9Cr oxide dispersion strengthened (ODS) steel was successfully diffusion bonded at 1150 °C for 1 h to migrate a residual bonding interface. Following heat treatment, including normalising at 1050 °C and tempering at 800 °C for 1 h, comparable results without inclusions or micro-voids at the bonding interface, or degradation in the base metal were achieved. Transmission electron microscopy (TEM) observation revealed that the nano-oxide particles in the bonding region were uniformly distributed in the matrix. At room temperature, the joint had nearly the same tensile properties with that of the base metal. The tensile strength of the joint region at elevated temperatures was comparable with that of the base metal. The total elongation of the joint region decreased slightly, but reached 80% of the base metal at 700 °C, and a ductile fracture occurred far from the bonding interface. Therefore, it is considered that diffusion bonding with a phase transformation can be a very useful joining method for

  17. Sigma phases in an 11%Cr ferritic/martensitic steel with the normalized and tempered condition

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Yinzhong, E-mail: shenyz@sjtu.edu.cn [School of Mechanical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Zhou, Xiaoling; Shi, Tiantian; Huang, Xi [School of Mechanical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Shang, Zhongxia [School of Materials Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Liu, Wenwen; Ji, Bo; Xu, Zhiqiang [School of Mechanical Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China)

    2016-12-15

    At the present time 9–12% Cr ferritic/martensitic (F/M) steels with target operating temperatures up to 650 °C and higher are being developed in order to further increase thermal efficiency so as to reduce coal consumption and air pollution. An 11% Cr F/M steel was prepared by reference to the nominal chemical composition of SAVE12 steel with an expected maximum use temperature of 650 °C. The precipitate phases of the 11% Cr F/M steel normalized at 1050 °C for 0.5 h and tempered at 780 °C for 1.5 h were investigated by transmission electron microscopy. Except for Cr-/Cr-Fe-Co-rich M{sub 23}C{sub 6}, Nb-/V-/Ta-Nb-/Nd-rich MX, Fe-rich M{sub 5}C{sub 2}, Co-rich M{sub 3}C and Fe-Co-rich M{sub 6}C phases previously identified in the steel, two types of sigma phases consisting of σ-FeCr and σ-FeCrW were found to be also present in the normalized and tempered steel. Identified σ-FeCr and σ-FeCrW phases have a simple tetragonal crystal structure with estimated lattice parameters a/c = 0.8713/0.4986 and 0.9119/0.5053 nm, respectively. The compositions in atomic pct of the observed sigma phases were determined to be approximately 50Fe-50Cr for the σ-FeCr, and 30Fe-55Cr-10W in addition to a small amount of Ta, Co and Mn for the σ-FeCrW. The sigma phases in the steel exhibit various blocky morphologies, and appear to have a smaller amount compared with the dominant phases Cr-rich M{sub 23}C{sub 6} and Nb-/V-/Ta-Nb-rich MX of the steel. The σ-FeCr phase in the steel was found to precipitate at δ-ferrite/martensite boundaries, suggesting that δ-ferrite may rapidly induce the formation of sigma phase at δ-ferrite/martensite boundaries in high Cr F/M steels containing δ-ferrite. The formation mechanism of sigma phases in the steel is also discussed in terms of the presence of δ-ferrite, M{sub 23}C{sub 6} precipitation, precipitation/dissolution of M{sub 2}X, and steel composition. - Highlights: •Precipitate phases in normalized and tempered 11%Cr F/M steel are

  18. New stainless steels of ferrite-martensite grade and perspectives of their application in thermonuclear facilities and fast reactors

    International Nuclear Information System (INIS)

    Ajtkhozhin, Eh.S.; Maksimkin, O.P.

    2007-01-01

    Review of scientific literature for last 5 years in which results on study of radiation effect on ferrite-martensite steels - construction materials of fast reactors and most probable candidates for first wall and blanket of the thermonuclear facilities ITER and Demo - are presented. Alongside with this a prior experimental data on study of microstructure changing and physical- mechanical properties of ferrite-martensite steel EhP-450 - the material of hexahedral case of spent assembly of BN-350 fast reactor- are cited. Principal attention was paid to considering of radiation effects of structural components content changing and ferrite-martensite steel swelling irradiated at comparatively low values of radiation damage climb rate

  19. Behavior of implanted hydrogen in ferritic/martensitic steels under irradiation

    Science.gov (United States)

    Wan, F.; Takahashi, H.; Ohnuki, S.; Nagasaki, R.

    1988-07-01

    The aim of this study was to clarify the behavior of hydrogen under irradiation in ferritic/martensitic stainless steel Fe-10Cr-2Mo-1Ni. Hydrogen was implanted into the specimens by ion accelerator or chemical cathodic charging method, followed by electron irradiation in a HVEM at temperatures from room temperature to 773 K. Streaks in the electron diffraction patterns were observed only during electron irradiation at 623-723 K. From these results it is suggested that the occurrence of the streak pattern is due to the formation of radiation-induced complexes of Ni or Cr with hydrogen along directions.

  20. Compression behavior of a ferritic-martensitic Cr-Mo steel

    DEFF Research Database (Denmark)

    Zhang, Zhenbo; Mishin, Oleg; Pantleon, Wolfgang

    2012-01-01

    The compression behavior of a ferritic-martensitic Cr-Mo steel is characterized for strain rates ranging from 10-4 s-1 to 10-1 s-1 and engineering strains up to 40%. Adiabatic heating causes a reduction in flow stress during continuous compression at a strain rate of 10-1 s-1. No reduction...... in the flow stress is observed if interrupted compression tests are performed with loading and holding steps. Two work-hardening stages with work-hardening rates decreasing linearly with the flow stress are identified and interpreted in terms of the KocksMecking model. The microstructural evolution...

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

    Science.gov (United States)

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

    2008-12-01

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

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

    International Nuclear Information System (INIS)

    Li Rutao; Zuo Xiurong; Hu Yueyue; Wang Zhenwei; Hu, Dingxu

    2011-01-01

    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: →The pipe with ferrite/martensite microstructure shows high deformability. →The base metal of the pipe consists of ferrite and martensite. →Heat affected zone shows excellent low temperature toughness. →Weld metal mainly consists of intragranularly nucleated acicular ferrites. →Weld metal shows excellent low temperature toughness and high strength.

  3. Microstructural control and high temperature mechanical property of ferritic/martensitic steels for nuclear reactor application

    International Nuclear Information System (INIS)

    Adetunji, G.J.

    1991-04-01

    The materials under study are 9-12% Cr ferritic/martensitic steels, alternative candidate materials for application in core components of nuclear power reactors. This work involves (1) Investigation of high temperature fracture mechanism during slow tensile and limited creep testing at 600 o C (2) Extensive study of solute element segregation both theoretically and experimentally (3) Investigation of effects by thermal ageing and irradiation on microstructural developments in relation to high temperature mechanical behaviour. From (1) the results obtained indicate that the important microstructural characteristics controlling the fracture of 9-12% Cr ferritic/martensitic steels at high temperature are (a) solute segregation to inclusion-matrix interfaces (b) hardness of the martensitic matrix and (c) carbide particle size distribution. From (2) the results indicate a strong concentration gradient of silicon and molybdenum near lath packet boundaries for certain quenching rates from the austenitizing temperature. From (3) high temperature tensile data were obtained for irradiated samples with thermally aged ones as control. (author)

  4. TIG of Reduced Activation Ferrite/Martensitic Steel for the Korean ITER-TBM

    International Nuclear Information System (INIS)

    Ku, Duck Young; Ahn, Mu Young; Yu, In Keun; Cho, Seun Gyon; Oh, Seung Jin

    2010-01-01

    Test Blanket Modules (TBM) will be tested in ITER to verify the capability of tritium breeding and recovery and the extraction of thermal energy suitable for the production of electricity. A Helium Cooled Solid Breeder (HCSB) TBM has been developed in Korea to accomplish these goals. Reduced Activation Ferritic/Martensitic (RAFM) steel has been chosen as the primary candidate structural material for Korean TBM. Due to the complexity of the First wall (FW) and Side wall (SW), it is necessary to develop various joining technologies, such as Hot Isostatic Pressing (HIP), Electron Beam Welding (EBW) and Tungsten Inert Gas (TIG) welding, for the successful fabrication of TBM. In this study, the mechanical properties of TIG welded RAFM steel were investigated. Various mechanical tests of TIG-welded RAFM steel were performed to obtain the optimized TIG welding process for RAFM steel

  5. TIG of Reduced Activation Ferrite/Martensitic Steel for the Korean ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Duck Young; Ahn, Mu Young; Yu, In Keun; Cho, Seun Gyon [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); Oh, Seung Jin [KHNP, Daejeon (Korea, Republic of)

    2010-10-15

    Test Blanket Modules (TBM) will be tested in ITER to verify the capability of tritium breeding and recovery and the extraction of thermal energy suitable for the production of electricity. A Helium Cooled Solid Breeder (HCSB) TBM has been developed in Korea to accomplish these goals. Reduced Activation Ferritic/Martensitic (RAFM) steel has been chosen as the primary candidate structural material for Korean TBM. Due to the complexity of the First wall (FW) and Side wall (SW), it is necessary to develop various joining technologies, such as Hot Isostatic Pressing (HIP), Electron Beam Welding (EBW) and Tungsten Inert Gas (TIG) welding, for the successful fabrication of TBM. In this study, the mechanical properties of TIG welded RAFM steel were investigated. Various mechanical tests of TIG-welded RAFM steel were performed to obtain the optimized TIG welding process for RAFM steel

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

    Science.gov (United States)

    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.

  7. Evaluation of temper embrittlement of martensitic and ferritic-martensitic steels by acoustic emission

    International Nuclear Information System (INIS)

    Lu, Yusho; Takahashi, Hideaki; Shoji, Tetsuo

    1987-01-01

    Martensitic (HT-9) and ferritic-martensitic steels (9Cr-2Mo) are considered as fusion first wall materials. In this investigation in order to understand the sensitivity of temper embrittlement in these steels under actual service condition, fracture toughness testing was made by use of acoustic emission technique. The temper embrittlement was characterized in terms of fracture toughness. The fracture toughness of these steels under 500 deg C, 100 hrs, and 1000 hrs heat treatment was decreased and their changes in micro-fracture process have been observed. The fracture toughness changes by temper embrittlement was discussed by the characteristic of AE, AE spectrum analysis and fractographic investigation. The relation between micro-fracture processes and AE has been clarified. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    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.

  9. Development of Reduced Activation Ferritic-Martensitic Steels and fabrication technologies for Indian test blanket module

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T., E-mail: tjk@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-10-01

    For the development of Reduced Activation Ferritic-Martensitic Steel (RAFMS), for the Indian Test Blanket Module for ITER, a 3-phase programme has been adopted. The first phase consists of melting and detailed characterization of a laboratory scale heat conforming to Eurofer 97 composition, to demonstrate the capability of the Indian industry for producing fusion grade steel. In the second phase which is currently in progress, the chemical composition will be optimized with respect to tungsten and tantalum for better combination of mechanical properties. Characterization of the optimized commercial scale India-specific RAFM steel will be carried out in the third phase. The first phase of the programme has been successfully completed and the tensile, impact and creep properties are comparable with Eurofer 97. Laser and electron beam welding parameters have been optimized and welding consumables were developed for Narrow Gap - Gas Tungsten Arc welding and for laser-hybrid welding.

  10. Creep deformation of high Cr-Mo ferritic/martensitic steels by material softening

    International Nuclear Information System (INIS)

    Kim, Sung Ho; Song, B. J.; Ryu, Woo Seog

    2005-01-01

    High Cr (9-12%Cr) ferritic/martensitic steels represent a valuable alternative to austenitic stainless steel for high temperature applications up to 600 .deg. C both in power and petrochemical plant, as well as good resistance to oxidation and corrosion. Material softening is the main physical phenomenon observed in the crept material. Thermally-induced change (such as particle coarsening or matrix solute depletion) and strain-induced change (such as dynamic subgrain growth) of microstructure degraded the alloy strength. These microstructural changes during a creep test cause the material softening, so the strength of the materials decreased. Many researches have been performed for the microstructural changes during a creep test, but the strength of crept materials has not been measured. In the present work, we measured the yield and tensile strength of crept materials using Indentationtyped Tensile Test System (AIS 2000). Material softening was quantitatively evaluated with a creep test condition, such as temperature and applied stress

  11. Deformation mechanisms in ferritic/martensitic steels and the impact on mechanical design

    International Nuclear Information System (INIS)

    Ghoniem, Nasr M.; Po, Giacomo; Sharafat, Shahram

    2013-01-01

    Structural steels for nuclear applications have undergone rapid development during the past few decades, thanks to a combination of trial-and-error, mechanism-based optimization, and multiscale modeling approaches. Deformation mechanisms are shown to be intimately related to mechanical design via dominant plastic deformation modes. Because mechanical design rules are mostly based on failure modes associated with plastic strain damage accumulation, we present here the fundamental deformation mechanisms for Ferritic/Martensitic (F/M) steels, and delineate their operational range of temperature and stress. The connection between deformation mechanisms, failure modes, and mechanical design is shown through application of design rules. A specific example is given for the alloy F82H utilized in the design of a Test Blanket Module (TBM) in the International Thermonuclear Experimental Reactor (ITER), where several constitutive equations are developed for design-related mechanical properties

  12. Deformation mechanisms in ferritic/martensitic steels and the impact on mechanical design

    Energy Technology Data Exchange (ETDEWEB)

    Ghoniem, Nasr M., E-mail: ghoniem@seas.ucla.edu; Po, Giacomo; Sharafat, Shahram

    2013-10-15

    Structural steels for nuclear applications have undergone rapid development during the past few decades, thanks to a combination of trial-and-error, mechanism-based optimization, and multiscale modeling approaches. Deformation mechanisms are shown to be intimately related to mechanical design via dominant plastic deformation modes. Because mechanical design rules are mostly based on failure modes associated with plastic strain damage accumulation, we present here the fundamental deformation mechanisms for Ferritic/Martensitic (F/M) steels, and delineate their operational range of temperature and stress. The connection between deformation mechanisms, failure modes, and mechanical design is shown through application of design rules. A specific example is given for the alloy F82H utilized in the design of a Test Blanket Module (TBM) in the International Thermonuclear Experimental Reactor (ITER), where several constitutive equations are developed for design-related mechanical properties.

  13. Development of an extensive database of mechanical properties for Reduced Activation Ferritic/Martensitic Steels

    International Nuclear Information System (INIS)

    Tanigawa, H.; Shiba, K.; Ando, M.; Wakai, E.; Jitsukawa, S.; Hirose, T.; Kasada, R.; Kimura, A.; Kohyama, A.; Kohno, Y.; Klueh, R.L.; Sokolov, M.; Stoller, R.; Zinklek, S.; Yamamoto, T.; Odette, G.; Kurtz, R.J.

    2007-01-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H (8Cr-2W-0.2V-0.04Ta-0.1C) and JLF-1 (9Cr-2W-0.2V-0.08Ta-0.1C) are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldablility, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The Japan/US collaboration program also has been conducted with the emphasis on heavy irradiation effects of F82H, JLF-1 and ORNL9Cr2WVTa over the past two decades using Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, and the irradiation condition of the irradiation capsules of those reactors were precisely controlled by the well matured capsule designing and instrumentation. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the database status of RAFMs, mainly on F82H, to identify the key issues for the future development of database. Tensile, fracture toughness, creep and fatigue properties and microstructural studies before and after irradiation are summarized. (authors)

  14. Development of an extensive database of mechanical properties for Reduced Activation Ferritic/Martensitic Steels

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H.; Shiba, K.; Ando, M.; Wakai, E.; Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kasada, R.; Kimura, A.; Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan); Kohno, Y. [Muroran Institute of Technology, Muroran, Hokkaido (Japan); Klueh, R.L. [0ak Ridge Noational Laboratory, TN (United States); Sokolov, M.; Stoller, R.; Zinklek, S. [0ak Ridge Noational Laboratory, Materials Science and Technology Div., TN (United States); Yamamoto, T.; Odette, G. [UCSB, Dept. of Chemical Engineering UCSB, Santa-Barbara (United States); Kurtz, R.J. [Pacifie Northwest National Laboratory, Richland WA (United States)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H (8Cr-2W-0.2V-0.04Ta-0.1C) and JLF-1 (9Cr-2W-0.2V-0.08Ta-0.1C) are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldablility, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The Japan/US collaboration program also has been conducted with the emphasis on heavy irradiation effects of F82H, JLF-1 and ORNL9Cr2WVTa over the past two decades using Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, and the irradiation condition of the irradiation capsules of those reactors were precisely controlled by the well matured capsule designing and instrumentation. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the database status of RAFMs, mainly on F82H, to identify the key issues for the future development of database. Tensile, fracture toughness, creep and fatigue properties and microstructural studies before and after irradiation are summarized. (authors)

  15. Resistance spot weldability of 11Cr–ferritic/martensitic steel sheets

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Yano, Yasuhide; Ito, Masahiro

    2012-01-01

    Resistance spot welding of 11Cr–0.4Mo–2W, V, Nb ferritic/martensitic steel sheets with different thicknesses was examined to develop a manufacturing technology for a fast reactor fuel subassembly with an inner duct structure. In the spot welding, welding current, electrode force, welding time and holding time were varied as welding parameters to investigate the appropriate welding conditions. Welding conditions under which spot weld joints did not have either crack or void defects in the nugget could be found when the electrode force was increased to 9.8 kN. It was also found that the electrode cap with a longer tip end length was effective for preventing weld defect formations. Strength of the spot welded joint was characterized from micro hardness and shear tension tests. In addition, the ductile-to-brittle transition temperature of the spot welded joint was measured by Charpy impact tests with specimens that had notches in the welded zone.

  16. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    International Nuclear Information System (INIS)

    Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

    2008-01-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed

  17. Thermo-mechanical fatigue behavior of reduced activation ferrite/martensite stainless steels

    International Nuclear Information System (INIS)

    Petersen, C.; Rodrian, D.

    2002-01-01

    The thermo-mechanical cycling fatigue (TMCF) behavior of reduced activation ferrite/martensite stainless steels is examined. The test rig consists of a stiff load frame, which is directly heated by the digitally controlled ohmic heating device. Cylindrical specimens are used with a wall thickness of 0.4 mm. Variable strain rates are applied at TMCF test mode, due to the constant heating rate of 5.8 K/s and variable temperature changes. TMCF results of as received EUROFER 97 in the temperature range between 100 and 500-600 deg. C show a reduction in life time (a factor of 2) compared to F82H mod. and OPTIFER IV. TMCF-experiments with hold times of 100 and 1000 s show dramatic reduction in life time for all three materials

  18. Radiation effects on low cycle fatigue properties of reduced activation ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Hirose, T.; Tanigawa, H.; Ando, M.; Kohyama, A.; Katoh, Y.; Narui, M.

    2002-01-01

    The reduced activation ferritic/martensitic steel, RAFs F82H IEA heat has been fatigue-tested at ambient temperature under diametral strain controlled conditions. In order to evaluate the effects of radiation damage and transmutation damage on fatigue characteristics, post-neutron irradiation and post-helium ion implantation fatigue tests were carried out. Fracture surfaces and fatigue crack initiation on the specimen surface were observed by SEM. Low-temperature irradiation caused an increase in stress amplitude and a reduction in fatigue lifetime corresponding to radiation hardening and loss of ductility. Neutron irradiated samples showed brittle fracture surface, and it was significant for large strain tests. On the other hand, helium implantation caused delay of cyclic softening. However, brittle crack initiation and propagation did not depend on the helium concentration profiles

  19. Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, P.; Lapena, J.; Blazquez, F. [Ciemat, Madrid (Spain)

    2000-07-01

    Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigree. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs.

  20. Corrosion of High Chromium Ferritic/Martensitic Steels in High Temperature Water. a Literature Review

    International Nuclear Information System (INIS)

    Fernandez, P.; Lapena, J.; Blazquez, F.

    2000-01-01

    Available literature concerning corrosion of high-chromium ferritic/martensitic steels in high temperature water has been reviewed. The subjects considered are general corrosion, effect of irradiation on corrosion, stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC). In addition some investigations about radiation induced segregation (RIS) are shown in order to know the compositional changes at grain boundaries of these alloys and their influence on corrosion properties. The data on general corrosion indicate moderate corrosion rates in high temperature water up to 350 degree centigrade. Considerably larger corrosion rates were observed under neutron irradiation. The works concerning to the behaviour of these alloys to stress corrosion cracking seem to conclude that in these materials is necessary to optimize the temper temperature and to carry out the post-weld heat treatments properly in order to avoid stress corrosion cracking. (Author) 40 refs

  1. Investigation on the Enhanced Oxidation of Ferritic/Martensitic Steel P92 in Pure Steam

    Directory of Open Access Journals (Sweden)

    Juntao Yuan

    2014-04-01

    Full Text Available Oxidation of ferritic/martensitic steel P92 was investigated in pure oxygen and in pure steam at 600–800 °C by thermogravimetric analysis (TGA, optical microscopy (OM, scanning electron microscopy (SEM, and X-ray diffraction (XRD. The results showed that the oxidation of P92 was significantly enhanced and multilayer scale with an outer iron oxides layer formed in pure steam. At 700 °C, the gas switch markedly influenced the scaling kinetics and scale microstructure. It was supposed that the higher affinity of iron to steam would be attributed to the enhanced oxidation of P92 in pure steam, and the much easier transport of hydroxyl would account for the significant difference induced by gas switch.

  2. Mechanical characterization of a reduced activation 9 Cr ferritic/martensitic steel of spanish production

    International Nuclear Information System (INIS)

    Rodriguez, D.; Serrano, M.

    2012-01-01

    This paper shows the first results concerning the characterization of two heats of a reduced activation 9 Cr ferritic/martensitic steel (RAFM) made in Spain, called AF1B and AF2A. The results of this characterization are compared with their European counterparts, EUROFER97-2, which was chosen as reference material. All activities described were performed in the Structural Materials Unit of CIEMAT, within the national project TECNO-FUS CONSOLIDER INGENIO.The two Spanish heats have the same production process and heat treatment. Both heats have a similar tensile behaviour similar to EUROFER97-2, but on the other hand impact properties are lower. The microstructure of AF1B reveals large biphasic inclusions that affecting its mechanical properties, especially the impact properties. AF2A casting was free of these inclusions. (Author) 24 refs.

  3. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Hirose, T.; Ogiwara, H.; Enoeda, M.; Akiba, M.

    2007-01-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 μm and 10 μm were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

  4. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ogiwara, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Naka-gun, Ibaraki-ken (Japan); Akiba, M. [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki (Japan)

    2007-07-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 {mu}m and 10 {mu}m were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F., E-mail: farhad.tavassoli@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA/DEN/DANS/DMN, F-91191 Gif-sur-Yvette (France); Diegele, E., E-mail: eberhard.diegele@kit.edu [Karlsruhe Institut of Technology (KIT), Karlsruhe (Germany); Lindau, R., E-mail: rainer.lindau@kit.edu [Karlsruhe Institut of Technology (KIT), Karlsruhe (Germany); Luzginova, N., E-mail: Natalia.Luzginova@gmail.com [NRG-Petten, 1755 ZG Petten (Netherlands); Tanigawa, H., E-mail: tanigawa.hiroyasu@jaea.go.jp [Japan Atomic Energy Authority (JAEA), Tokai, Ibaraki, 319-1195 (Japan)

    2014-12-15

    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.

  7. Thermal and mechanical behaviour of the reduced-activation-ferritic-martensitic steel EUROFER

    International Nuclear Information System (INIS)

    Lindau, R.; Moeslang, A.; Schirra, M.

    2002-01-01

    Reduced activation ferritic/martensitic (RAFM) steels are being considered for structural application in potential fusion energy systems. Based on the substantial experience with RAFM developmental steels of OPTIFER type, an industrial 3.5 tons batch of a 9CrWVTa-RAFM steel, called EUROFER 97 had been specified and ordered. A characterisation programme has been launched to determine the relevant mechanical and physical-metallurgical properties in order to qualify the steel for fusion application. The hardening, tempering and transformation behaviour of EUROFER is in good agreement with that of other RAFM-steels like OPTIFER and the Japanese industrial scale heat F82H mod. Tensile tests, performed between RT and 750 deg. C, show comparable strength and ductility values that are not strongly affected by different heat treatments and ageing at 580 and 600 deg. C up to 3300 h. Impact bending tests indicate a superior ductile to brittle transition temperature (DBTT) of EUROFER in the as-received condition compared with that of F82H mod. Creep tests between 450 and 650 deg. C up to test times of 15000 h reveal a creep strength similar to other RAFM steels like OPTIFER and F82H mod. EUROFER shows a good low-cycle fatigue behaviour with longer lifetimes than F82H mod. The deformation and softening behaviour is similar

  8. Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

    Science.gov (United States)

    Obraztsov, S. M.; Konobeev, Yu. V.; Birzhevoy, G. A.; Rachkov, V. I.

    2006-12-01

    The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.

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

    Science.gov (United States)

    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 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.

  10. Development of Reduced Activation Ferritic-Martensitic Steels in South Korea

    International Nuclear Information System (INIS)

    Chun, Y. B.; Choi, B. K.; Han, C. H.; Lee, D. W.; Cho, S.; Kim, T. K.; Jeong, Y. H.

    2012-01-01

    In the mid-1980s research programs for development of low activation materials began. This is based on the US Nuclear Regulatory Commission Guidelines (10CFR part 61) that were developed to reduce longlived radioactive isotopes, which allows nuclear reactor waste to be disposed of by shallow land burial when removed from service. Development of low activation materials is also key issue in nuclear fusion systems, as the structural components can became radioactive due to nuclear transmutation caused by exposure to high dose neutron irradiation. Reduced-activation ferritic martensitic (RAFM) steels have been developed in the leading countries in nuclear fusion technology, and are now being considered as candidate structural material for the test blanket module (TBM) in the international thermonuclear experiment reactor (ITER). South Korea joined the ITER program in 2003 and since then extensive effort has been made for developing the helium-cooled solid-breeder (HCSB) TBM which is scheduled to be tested in the ITER program. However, there has been no research activity to develop RAFM steels in South Korea, while all the participants in the ITER program have developed their own RAFM steels. It is recently that the Korea Atomic Energy Research Institute (KAERI) started the Korean RAFM steel research program, aiming at an application for the HCSB-type TBM structure in ITER. In what follows, the current status of RAFM steels and the R and D program led by KAERI to develop Korean RAFM steels are summarized

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

    Science.gov (United States)

    Vasantharaja, P.; Vasudevan, M.

    2012-02-01

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

  12. European development of ferritic-martensitic steels for fast reactor wrapper applications

    International Nuclear Information System (INIS)

    Bagley, K.; Little, E.A.; Levy, V.; Alamo, A.

    1987-01-01

    9-12%Cr ferritic-martensitic stainless steels are under development in Europe for fast reactor sub-assembly wrapper applications. Within this class of alloys, attention is focussed on three key specifications, viz. FV448 and DIN 1.4914 (both 10-12%CrMoVNb steels) and EM10 (an 8-10%Cr-0.15%C steel), which can be optimized to give acceptably low ductile-brittle transition characteristics. The results of studies on these steels, and earlier choices, covering heat treatment and compositional optimization, evolution of wrapper fabrication routes, pre and post-irradiation mechanical property and fracture toughness behaviour, microstructural stability, void swelling and in-reactor creep characteristics are reviewed. The retention of high void swelling to displacement doses in excess of 100 dpa in reactor irradiations reaffirms the selection of 9-12%Cr steels for on-going wrapper development. Moreover, irradiation-induced changes in mechanical properties (e.g. in-reactor creep and impact behaviour), measured to intermediate doses, do not give cause for concern; however, additional data to higher doses and at the lower irradiation temperatures of 370 0 -400 0 C are needed in order to fully endorse these alloys for high burnup applications in advanced reactor systems

  13. Influence of Prior Fatigue Cycling on Creep Behavior of Reduced Activation Ferritic-Martensitic Steel

    Science.gov (United States)

    Sarkar, Aritra; Vijayanand, V. D.; Parameswaran, P.; Shankar, Vani; Sandhya, R.; Laha, K.; Mathew, M. D.; Jayakumar, T.; Rajendra Kumar, E.

    2014-06-01

    Creep tests were carried out at 823 K (550 °C) and 210 MPa on Reduced Activation Ferritic-Martensitic (RAFM) steel which was subjected to different extents of prior fatigue exposure at 823 K at a strain amplitude of ±0.6 pct to assess the effect of prior fatigue exposure on creep behavior. Extensive cyclic softening that characterized the fatigue damage was found to be immensely deleterious for creep strength of the tempered martensitic steel. Creep rupture life was reduced to 60 pct of that of the virgin steel when the steel was exposed to as low as 1 pct of fatigue life. However, creep life saturated after fatigue exposure of 40 pct. Increase in minimum creep rate and decrease in creep rupture ductility with a saturating trend were observed with prior fatigue exposures. To substantiate these findings, detailed transmission electron microscopy studies were carried out on the steel. With fatigue exposures, extensive recovery of martensitic-lath structure was distinctly observed which supported the cyclic softening behavior that was introduced due to prior fatigue. Consequently, prior fatigue exposures were considered responsible for decrease in creep ductility and associated reduction in the creep rupture strength.

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

    Science.gov (United States)

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

    2017-02-01

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

  15. Current status of reduced-activation ferritic/martensitic steels R and D for fusion energy

    International Nuclear Information System (INIS)

    Kimura, Akihiko

    2005-01-01

    Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM). Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement. A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Krsjak, Vladimir, E-mail: vladimir.krsjak@psi.ch; Dai, Yong

    2015-10-15

    This paper presents the use of an internal {sup 44}Ti/{sup 44}Sc 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 {sup 44}Ti 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 {sup 44}Ti → {sup 44}Sc → {sup 44}Ca, 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 {sup 44}Ti, 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.

  17. Development of ferritic-martensitic P9 steel for wrapper application in future SFRs

    International Nuclear Information System (INIS)

    Choudhary, B.K.; Mathew, M.D.; Isaac Samuel, E.; Moitra, A.

    2011-01-01

    The paper deals with the outcome of the research and development efforts directed towards the development of ferritic-martensitic P9 steel for wrapper application in future sodium cooled fast reactors with an objective to achieve high fuel burnup and more economical nuclear energy. The important and critical issues involved for the development of P9 wrappers such as optimisation of chemical composition in terms of trace elements like sulphur and phosphorous and appropriate thermo-mechanical treatments along with thermal ageing and irradiation effects on fracture properties have been discussed. Tensile properties evaluated at temperatures ranging from 300 to 873 K on the experimental three heats of P9 steel with different silicon contents and made using primary vacuum induction melting followed by secondary electro slag refining route, have been presented. Fracture behaviour examined mainly in terms of ductile to brittle transition temperature and upper shelf energy provided encouraging results. Based on these investigations, a roadmap has been drawn to make experimental P9 steel wrappers for tests in fast breeder test reactor and prototype fast breeder reactor. (author)

  18. Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

    International Nuclear Information System (INIS)

    Little, E.A.; Stoter, L.P.

    1982-01-01

    The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

  19. Multiscale simulation of yield strength in reduced-activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chen Chong; Zhang, Chi; Yang, Zhigang [Key Laboratory of Advanced Materials of Ministry of Education, School of Materials Science and Engineering, Tsinghua University, Beijing (China); Zhao, Ji Jun [State Key Laboratory of Materials Modification by Laser, Electron, and Ion Beams, School of Physics and Optoelectronic Technology and College of Advanced Science and Technology, Dalian University of Technology, Dalian (China)

    2017-04-15

    One of the important requirements for the application of reduced-activation ferritic/martensitic (RAFM) steel is to retain proper mechanical properties under irradiation and high-temperature conditions. To simulate the yield strength and stress-strain curve of steels during high-temperature and irradiation conditions, a multiscale simulation method consisting of both microstructure and strengthening simulations was established. The simulation results of microstructure parameters were added to a superposition strengthening model, which consisted of constitutive models of different strengthening methods. Based on the simulation results, the strength contribution for different strengthening methods at both room temperature and high-temperature conditions was analyzed. The simulation results of the yield strength in irradiation and high-temperature conditions were mainly consistent with the experimental results. The optimal application field of this multiscale model was 9Cr series (7–9 wt.%Cr) RAFM steels in a condition characterized by 0.1–5 dpa (or 0 dpa) and a temperature range of 25–500°C.

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

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, S.Bulent

    2015-01-01

    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.

  1. High Temperature Elastic Properties of Reduced Activation Ferritic-Martensitic (RAFM) Steel Using Impulse Excitation Technique

    Science.gov (United States)

    Tripathy, Haraprasanna; Raju, Subramanian; Hajra, Raj Narayan; Saibaba, Saroja

    2018-03-01

    The polycrystalline elastic constants of an indigenous variant of 9Cr-1W-based reduced activation ferritic-martensitic (RAFM) steel have been determined as a function of temperature from 298 K to 1323 K (25 °C to 1000 °C), using impulse excitation technique (IET). The three elastic constants namely, Young's modulus E, shear modulus G, and bulk modulus B, exhibited significant softening with increasing temperature, in a pronounced non-linear fashion. In addition, clearly marked discontinuities in their temperature variations are noticed in the region, where ferrite + carbides → austenite phase transformation occurred upon heating. Further, the incidence of austenite → martensite transformation upon cooling has also been marked by a step-like jump in both elastic E and shear moduli G. The martensite start M s and M f finish temperatures estimated from this study are, M s = 652 K (379 °C) and M f =580 K (307 °C). Similarly, the measured ferrite + carbide → austenite transformation onset ( Ac 1) and completion ( Ac 3) temperatures are found to be 1126 K and 1143 K (853 °C and 870 °C), respectively. The Poisson ratio μ exhibited distinct discontinuities at phase transformation temperatures; but however, is found to vary in the range 0.27 to 0.29. The room temperature estimates of E, G, and μ for normalized and tempered microstructure are found to be 219 GPa, 86.65 GPa, and 0.27, respectively. For the metastable austenite phase, the corresponding values are: 197 GPa, 76.5 GPa, and 0.29, respectively. The measured elastic properties as well as their temperature dependencies are found to be in good accord with reported estimates for other 9Cr-based ferritic-martensitic steel grades. Estimates of θ D el , the elastic Debye temperature and γ G, the thermal Grüneisen parameter obtained from measured bulk elastic properties are found to be θ D el = 465 K (192 °C) and γ G = 1.57.

  2. Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

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

    Energy Technology Data Exchange (ETDEWEB)

    Sam, Shiju, E-mail: shiju@ipr.res.in [Institute for Plasma Research, Gandhinagar, Gujarat 382 428 (India); Das, C.R.; Ramasubbu, V.; Albert, S.K.; Bhaduri, A.K.; Jayakumar, T. [Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Rajendra Kumar, E. [Institute for Plasma Research, Gandhinagar, Gujarat 382 428 (India)

    2014-12-15

    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.

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

    Science.gov (United States)

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

    2012-12-01

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

  5. Technical issues of fabrication technologies of reduced activation ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Tanigawa, Hiroyasu; Sakasegawa, Hideo; Hirose, Takanori

    2013-01-01

    Highlights: • The key technical issues of RAFM steel fabrication are the control of Ta, and deoxidation of the steel with a limited amount of Al addition. • Addition of Ta with poor deoxidation might results in the agglomeration of inclusions at 1/2t position. • ESR was proved to be effective removing Ta oxide inclusions and avoiding agglomeration of inclusions at 1/2t position, and achieving low oxygen concentration. -- Abstract: The key issue for DEMO application is that Reduced activation ferritic/martensitic (RAFM) steels fabrication technologies has to be highly assured, especially with respect to high availability, reliability and reduced activation capability on the DEMO level fabrication, which requires not a few tons but thousand tons RAFM fabrication. One of the key technical issues of RAFM fabrication is the control of Ta, and deoxidation of the steel with a limited amount of Al addition. The series of F82H (Fe–8Cr–2W–V, Ta) melting revealed that Ta have tendency to form oxide on melting process, and this will have large impact on reliability of the steels. Al is also the key elements, as it is commonly used for deoxidation of steels, and achieving lower oxygen level is essential to obtain good mechanical properties, but the maximum concentration of Al is limited in view of reduced activation capability. These tendency and limitation resulted in the Ta oxide agglomeration in the middle of plate, but the remelting process, ESR (electro slag remelting), was found to be successful on removing those Ta oxides

  6. Corrosion Behavior and Oxide Film Formation of T91 Steel under Different Water Chemistry Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, D. Q.; Shi, C.; Li, J.; Gao, L. X. [Shanghai University of Electric Power, Shanghai (China); Lee, K. Y. [Dalian University of Technology, Dalian (China)

    2017-02-15

    The corrosion behavior of a ferritic/martensitic steel T91 exposed to an aqueous solution containing chloride and sulfate ions is investigated depending on the stimulated all-volatile treatment (AVT) and under oxygenated treatment (OT) conditions. The corrosion of T91 steel under OT condition is severe, while the corrosion under AVT condition is not. The co-existence of chloride and sulfate ions has antagonistic effect on the corrosion of T91 steel in both AVT and OT conditions. Unlike to corrosion resistance in the aqueous solution, OT pretreatment provides T91 steel lower oxidation-resistance than VAT pretreatment. From scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction (XRD) analysis, the lower corrosion resistance in the aqueous solution by VAT conditions possibly is due to the formation of pits. In addition, the lower oxidation resistance of T91 steel pretreated by OT conditions is explained as follows: the cracks formed during the immersion under OT conditions accelerated peeling-off rate of the oxide film.

  7. Progress of reduced activation ferritic/martensitic steel development in Japan

    International Nuclear Information System (INIS)

    Jitsukawa, S.; Kimura, A.; Kohyama, A.; Ukai, S.; Sawai, T.; Wakai, E.; Shiba, K.; Miwa, Y.; Furuya, K.; Tanigawa, H.; Ando, M.

    2005-01-01

    Recent accomplishment by the Japanese activity for the reduced activation ferritic/martensitic steel (RAF/M) development has been reviewed. Some of the results obtained in EU and US by international collaborative activities are also introduced. Effect of irradiation on the shift of ductile-to-brittle transition temperature (DBTT) has been evaluated to a dose of 20dpa. Results suggest that RAF/M appears to satisfy the requirement on DBTT-shift for the blanket application in the dose range up to several tens of dpa. Also, enhancement effect of DBTT-shift by transmutation produced helium (He) atoms was revealed to be smaller than has been suggested previously. Preliminary studies about the effect of irradiation on fatigue mechanism, the susceptibility to environmentally assisted cracking in water and flow stress-strain relation have been conducted for the specimens irradiated to several dpa, including the post irradiation tensile property examination of the joints by Hot-isostatic press (HIP) bonding method. The results also indicate that RAF/Ms exhibit suitable properties for ITER test blanket module. (author)

  8. Evaluation of creep rupture property of high strength ferritic/martensitic steel (PNC-FMS)

    International Nuclear Information System (INIS)

    Uehira, Akihiro; Mizuno, Tomoyasu; Ukai, Shigeharu; Yoshida, Eiichi

    1999-04-01

    High Strength Ferritic/Martensitic Steel (PNC-FMS : 11Cr-0.5Mo-2W,Nb,V), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. The material design base standard (tentative) of PNC-FMS was established and the creep rupture strength reduction factor in the standard was determined in 1992. This factor was based on only evaluation of decarburization effect on tensile strength after sodium exposure. In this study, creep rupture properties of PNC-FMS under out of pile sodium exposure and in pile were evaluated, using recent test results as well as previous ones. The evaluation results are summarized as follows : a. Decarburization rate constant of pressurized tubes under sodium exposure is identical with stress free specimens. b. In case of the same decarburization content under out of pile sodium exposure, creep strength tends to decrease more significantly than tensile strength. c. Creep strength under out of pile sodium exposure showed significant decrease in high temperature and long exposure time, but in pile (MOTA) creep strength showed little decrease. A new creep rupture strength reduction factor, which is the ratio of creep rupture strength under sodium exposure or in pile to in air, was made by correlating the creep rupture strength. This new method directly using the ratio of creep rupture strength was evaluated and discussed from the viewpoint of design applicability, compared with the conventional method based on decarburization effect on tensile strength. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  10. Compatibility of reduced activation ferritic/martensitic steels with liquid breeders

    International Nuclear Information System (INIS)

    Muroga, T.; Nagasaka, T.; Kondo, M.; Sagara, A.; Noda, N.; Suzuki, A.; Terai, T.

    2008-10-01

    The compatibility of Reduced Activation Ferritic/Martensitic Steel (RAFM) with liquid Li and molten-salt Flibe have been characterized and accessed. Static compatibility tests were carried out in which the specimens were immersed into liquid Li or Flibe in isothermal autoclaves. Also carried out were compatibility tests in flowing liquid Li by thermal convection loops. In the case of liquid Li, the corrosion rate increased with temperature significantly. The corrosion was almost one order larger for the loop tests than for the static tests. Chemical analysis showed that the corrosion was enhanced when the level of N in Li is increased. Transformation from martensitic to ferritic phase and the resulting softening were observed in near-surface area of Li-exposed specimens, which were shown to be induced by decarburization. In the case of Flibe, the corrosion loss was much larger in a Ni crucible than in a RAFM crucible. Both fluorides and oxides were observed on the surfaces. Thus, the key corrosion process of Flibe is the competing process of fluoridation and oxidation. Possible mechanism of the enhanced corrosion in Ni crucible is electrochemical circuit effect. It was suggested that the corrosion loss rate of RAFM by liquid Li and Flibe can be reduced by reducing the level of impurity N in Li and avoiding the use of dissimilar materials in Flibe, respectively. (author)

  11. Fatigue life assessment based on crack growth behavior in reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Nogami, Shuhei; Sato, Yuki; Hasegawa, Akira

    2010-01-01

    Crack growth behavior under low cycle fatigue in reduced activation ferritic/martensitic steel, F82H IEA-heat (Fe-8Cr-2W-0.2V-0.02Ta), was investigated to improve the fatigue life assessment method of fusion reactor structural material. Low cycle fatigue test was carried out at room temperature in air at a total strain range of 0.4-1.5% using an hourglass-type miniature fatigue specimen. The relationship between the surface crack length and life fraction was described using one equation independent of the total strain range. Therefore, the fatigue life and residual life could be estimated using the surface crack length. Moreover, the microcrack initiation life could be estimated using the total strain range if there was a one-to-one correspondence between the total strain range and number of cycles to failure. The crack growth rate could be estimated using the total strain range and surface crack length by introducing the concept of the normalized crack growth rate. (author)

  12. Evaluation of mechanical properties of weldments for reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, T. [Muroran Institute of Technology, Dept. of Materials Science and Engineeering, Muroran, Hokkaido (Japan); Tanigawa, H.; Ando, M. [Japan Atomic Energy Agency, Naga-gun, Ibaraki-ken (Japan); Komazaki, S.; Kohno, Y. [Muroran Institute of Technology, Muroran (Japan)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels are the first candidate material for fusion reactor, and will be used as the structural materials of ITER test blanket modules (TBM). TBM will be assembled by welding various parts, it is important to be clearly mechanical properties of weldments to qualify TBM structure. In this paper, unirradiated mechanical properties of weldments, which is consisted of weld metal, heat affected zone (HAZ) and base metal region, obtained from TIG and EB welded F82H IEA-heat were evaluated by tensile, Charpy impact and creep test. Charpy impact test revealed that impact properties of weld metal does not deteriorate compared with that of base metal. The creep tests were carried out at temperatures of 773-873 K and at stress levels of 130-280 MPa, with the specimens which include weld metal and HAZ region in the gage section. In these conditions, rupture time of weldments yield to about 100-1000 hours. In the high-stress range, creep lives of welded joint decreased about 40% of base metal. However, in the low-stress range, creep lives of welded joint decrease about 60 to 70% of base metal. The failure at fine grain HAZ region (Type IV failure) does not occur in these conditions. The mechanism of these properties deterioration will be discussed based on the detailed analyses on microstructure changes. (authors)

  13. Effect of helium on fatigue crack growth and life of reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Nogami, Shuhei; Takahashi, Manabu; Hasegawa, Akira; Yamazaki, Masanori

    2013-01-01

    The effects of helium on the fatigue life, micro-crack growth behavior up to final fatigue failure, and fracture mode under fatigue in the reduced activation ferritic/martensitic steel, F82H IEA-heat, were investigated by low cycle fatigue tests at room temperature in air at a total strain range of 0.6–1.5%. Significant reduction of the fatigue life due to helium implantation was observed for a total strain range of 1.0–1.5%, which might be attributable to an increase in the micro-crack propagation rate. However, the reduction of fatigue life due to helium implantation was not significant for a total strain range of 0.6–0.8%. A brittle fracture surface (an original point of micro-crack initiation) and a cleavage fracture surface were observed in the helium-implanted region of fracture surface. A striation pattern was observed in the non-implanted region. These fracture modes of the helium-implanted specimen were independent of the strain range

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

  15. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER

    International Nuclear Information System (INIS)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-01-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Schubert, L.E. [Univ. of Missouri, Rolla, MO (United States)

    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. Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications

    Science.gov (United States)

    Zinkle, S. J.; Boutard, J. L.; Hoelzer, D. T.; Kimura, A.; Lindau, R.; Odette, G. R.; Rieth, M.; Tan, L.; Tanigawa, H.

    2017-09-01

    Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniques to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. Material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.

  18. Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

    International Nuclear Information System (INIS)

    Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

    2012-01-01

    The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to its superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.

  19. Hardening and embrittlement mechanisms of reduced activation ferritic/martensitic steels irradiated at 573 K

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Hashimoto, N. [Hokkaido Univ., Materials Science and Engineering Div., Graduate School of Engineering, Sapporo (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

    2007-07-01

    Full text of publication follows: It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1, showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M{sub 23}C{sub 6} at an irradiation temperature of 573 K. The calculation also suggests the presence of Laves phase at 573 K, which is usually not observed at this temperature, but the ion irradiation on aged F82H with Laves phase suggests that Laves phase becomes amorphous and could not be stable under irradiation at 573 K. This observation indicates the possibility that the irradiation-induced nano-size precipitation could be the consequence of the conflict between precipitation and amorphization of Laves phase. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K. (authors)

  20. Corrosion behavior of austenitic and ferritic/martensitic steels in oxygen-saturated liquid Pb-Bi eutectic at 450circC and 550circC

    OpenAIRE

    倉田 有司; 二川 正敏; 斎藤 滋

    2005-01-01

    Static corrosion tests of various austenitic and ferritic/martensitic steels were conducted in oxygen-saturated liquid Pb-Bi at 450circC and 550circC for 3000h to study the effects of temperature and alloying elements on corrosion behavior. Oxidation, grain boundary corrosion, dissolution and penetration were observed. The corrosion depth decreases at 450circC with increasing Cr content in steels regardless of ferritic/martensitic or austenitic steels. Appreciable dissolution of Ni and Cr doe...

  1. Material science and manufacturing of heat-resistant reduced-activation ferritic-martensitic steels for fusion

    International Nuclear Information System (INIS)

    Ioltukhovskiy, A.G.; Blokhin, A.I.; Budylkin, N.I.; Chernov, V.M.; Leont'eva-Smirnova, M.V.; Mironova, E.G.; Medvedeva, E.A.; Solonin, M.I.; Porollo, S.I.; Zavyalsky, L.P.

    2000-01-01

    A number of issues regarding the development and use of 10-12% Cr reduced-activation ferritic-martensitic steels (RAFMS) for fusion are considered. These include: (1) problems of manufacturing and modifying their composition and metallurgical condition; (2) the influence on properties of their composition, purity, δ-ferrite concentration and cooling rates in the final stages of manufacturing; and (3) the effects of neutron irradiation at 320-650 deg. C up to 108 dpa on their mechanical properties. In addition, neutron activation and nuclear accumulation of elements in RAFMS with different initial concentrations of alloying and impurity elements for typical fusion reactor (DEMO) irradiation regimes have been calculated

  2. Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Kim, S.W.; Tanigawa, H.; Hirose, T.; Kohyama, A.

    2007-01-01

    Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

  3. Effects of irradiation on low cycle fatigue properties for reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.W. [Kyoto Univ., Graduate School of Energy Science (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

    2007-07-01

    Full text of publication follows: In materials life decision for a commercial blanket, thermal fatigue property of materials is a particularly important. The loading of structural materials in fusion reactor is, besides the plasma surface interactions, a combined effect of high heat fluxes and neutron irradiation. Depending on the pulse lengths, the operating conditions, and the thermal conductivity, these oscillating temperature gradients will cause elastic and elastic-plastic cyclic deformation giving rise to (creep-) fatigue in structural first wall and blanket components. Especially, investigation of the fatigue property in Reduced Activation Ferritic/Martensitic (RAF/M) steel and establishment of the evaluation technology are demanded in particular immediately for design/manufacturing of ITER-TBM. And also, fatigue testing after irradiation will be carried out in hot cells with remote control system. Considering limited ability of specimen manipulation in the cells, the specimen and the test method need to be simple for operation. The existing data bases of RAF/M steel provide baseline data set including post-irradiation fatigue data. However, to perform the accurate fatigue lifetime assessment for ITER-TBM and beyond utilizing the existing data base, the mechanical understanding of fatigue fracture is mandatory. It has been previously reported by co-authors that dislocation cell structure was developed on low cycle fatigued RAF/M steel, and led the fatigue crack to develop along prior austenitic grain boundary. In this work, the effects of nuclear irradiation on low cycle fatigue properties for RAF/M steels and its fracture mechanisms were examined based on the flow stress analysis and detailed microstructure analysis. Fracture surfaces and crack initiation site were investigated by scanning electron microscope (SEM). Transmission electron microscopy (TEM) was also applied to clarify the microstructural features of fatigue behavior. It is also important to

  4. Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

    2007-07-01

    Full text of publication follows: Neutron irradiation of 9-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects that cause an increase in yield stress and ultimate tensile strength. This irradiation hardening causes embrittlement, which is observed in Charpy impact and toughness tests as an increase in ductile-brittle transition temperature (DBTT). Based on observations that show little change in strength in these steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above this irradiation-hardening temperature regime. In a recent study of F82H steel irradiated at 300, 380, and 500 deg. C, irradiation hardening-an increase in yield stress-was observed in tensile specimens irradiated at the two lower temperatures, but no change was observed for the specimens irradiated at 500 deg. C. As expected, an increase in DBTT occurred for the Charpy specimens irradiated at 300 and 380 deg. C. However, there was an unexpected increase in the DBTT of the specimens irradiated at 500 deg. C. The observed embrittlement was attributed to the irradiation-accelerated precipitation of Laves phase. This conclusion was based on results from a detailed thermal aging study of F82H, in which tensile and Charpy specimens were aged at 500, 550, 600, and 650 deg. C to 30,000 h. These studies indicated that there was a decrease in yield stress at the two highest temperatures and essentially no change at the two lowest temperatures. Despite the strength decrease or no change, the DBTT increased for Charpy specimens irradiated at all four temperatures. Precipitates were extracted from thermally aged specimens, and the amount of precipitate was correlated with the increase in transition temperature. Laves phase was identified in the extracted precipitates by X-ray diffraction. Earlier studies on conventional elevated-temperature steels also showed embrittlement effects above the irradiation-hardening temperature

  5. Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

    International Nuclear Information System (INIS)

    Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk; Jeong, Jong Ryul; Maeng, Cheol Soo; Lee, Myoung Goo

    2016-01-01

    As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10"-"5 dpa.

  6. Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

    Energy Technology Data Exchange (ETDEWEB)

    Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk [KAIST, Daejeon (Korea, Republic of); Jeong, Jong Ryul [Chungnam University, Daejeon (Korea, Republic of); Maeng, Cheol Soo; Lee, Myoung Goo [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10{sup -5} dpa.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-12-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  9. Effects of Mn addition on microstructures and mechanical properties of 10Cr ODS ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Jin, Hyun Ju; Kim, Tae Kyu

    2014-01-01

    Ferritic/martensitic (FM) steels are very attractive for the structural materials of fast fission reactors such as a sodium cooled fast reactor (SFR) owing to their excellent irradiation resistance to a void swelling, but are known to reveal an abrupt loss of their creep and tensile strengths at temperatures above 600 .deg. C. Accordingly, high temperature strength should be considerably improved for an application of the FM steel to the structural materials of SFR. Oxide dispersion strengthened (ODS) FM steels are considered to be promising candidate materials for high- temperature components operating in severe environments such as nuclear fusion and fission systems due to their excellent high temperature strength and radiation resistance stemming from the addition of extremely thermally stable oxide particles dispersed in the ferritic/martensitic matrix.. To develop an advanced ODS steel for core structural materials for next generation nuclear reactor system applications, it is important to optimize its compositions to improve the high temperature strength and radiation resistance. This study investigates effects of Mn addition on microstructures and mechanical properties of 10Cr ODS FM steel. For this, two 10 Cr ODS FM steels were prepared by mechanical alloying (MA), hot isostatic pressing (HIP), and hot rolling process. Tensile tests were carried out at room temperature and 700 .deg. C to evaluate the influences of the Mn element on the mechanical properties. The microstructures were observed using SEM, electron back-scatter diffraction (EBSD) and transmission electron microscopy (TEM) with energy dispersive spectroscopy (EDS). In the present study, the effects of Mn addition on the microstructure and mechanical properties of ODS FM steels were investigated. The ODS FM steels were manufactured by the MA, HIP and hot-rolling processes

  10. Effect of alloying element partitioning on ferrite hardening in a low alloy ferrite-martensite dual phase steel

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimian, A., E-mail: ebrahimiana@yahoo.com; Ghasemi Banadkouki, S.S.

    2016-11-20

    In this paper, the effect of carbon and other alloying elements partitioning on ferrite hardening behavior were studied in details using a low alloy AISI4340 ferrite-martensite dual phase (DP) steel. To do so, various re-austenitised samples at 860 °C for 60 min were isothermally heated at 650 °C from 3 to 60 min and then water–quenched to obtain the final ferrite-martensite DP microstructures containing different ferrite and martensite volume fractions. Light and electron microscopic observations were supplemented with electron dispersive spectroscopy (EDS) and nanoindentation tests to explore the localized compositional and hardening variations within ferrite grains in DP samples. The experimental results showed that the ferrite hardness was varied with progress of austenite to ferrite phase transformation in DP samples. In the case of a particular ferrite grain in a particular DP sample, despite a homogeneous distribution of carbon concentration, the ferrite hardness was significantly increased by increasing distance from the central location toward the interfacial α/γ areas. Beside a considerable influence of martensitic phase transformation on adjacent ferrite hardness, these results were rationalized in part to the significant level of Cr and Mo pile-up at α/γ interfaces leading to higher solid solution hardening effect of these regions. The reduction of potential energy developed by attractive interaction between C-Cr and C-Mo couples toward the carbon enriched prior austenite areas were the dominating driving force for pile-up segregation.

  11. Corrosion behaviour of dissimilar welds between ferritic-martensitic stainless steel and austenitic stainless steel from secondary circuit of CANDU NPP

    International Nuclear Information System (INIS)

    Popa, L.; Fulger, M.; Tunaru, M.; Velciu, L.; Lazar, M.

    2016-01-01

    Corrosion damages of welds occur in spite of the fact that the proper base metal and filler metal have been correctly selected, industry codes and standards have been followed and welds have been realized with full weld penetration and have proper shape and contour. In secondary circuit of a Nuclear Power Station there are some components which have dissimilar welds. The principal criteria for selecting a stainless steel usually is resistance to corrosion, and white most consideration is given to the corrosion resistance of the base metal, additional consideration should be given to the weld metal and to the base metal immediately adjacent to the weld zone. Our experiments were performed in chloride environmental on two types of samples: non-welded (410 or W 1.4006 ferritic-martensitic steel and 304L or W 1.4307 austenitic stainless steel) and dissimilar welds (dissimilar metal welds: joints between 410 ferritic-martensitic and 304L austenitic stainless steel). To evaluate corrosion susceptibility of dissimilar welds was used electrochemical method (potentiodynamic method) and optic microscopy (microstructural analysis). The present paper follows the localized corrosion behaviour of dissimilar welds between austenitic stainless steel and ferritic-martensitic steel in solutions containing chloride ions. It was evaluated the corrosion rates of samples (welded and non-welded) by electrochemical methods. (authors)

  12. Carbon Contamination During Ion Irradiation - Accurate Detection and Characterization of its Effect on Microstructure of Ferritic/Martensitic Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jing; Toloczko, Mychailo B.; Kruska, Karen; Schreiber, Daniel K.; Edwards, Danny J.; Zhu, Zihua; Zhang, Jiandong

    2017-11-17

    Accelerator-based ion beam techniques have been used to study radiation effects in materials for decades. Although carbon contamination induced by ion beam in target materials is a well-known issue, it has not been fully characterized nor quantified for studies in ferritic/martensitic (F/M) steels that are candidate materials for applications such as core structural components in advanced nuclear reactors. It is an especially important issue for this class of material because of the effect of carbon level on precipitate formation. In this paper, the ability to quantify carbon contamination using three common techniques, namely time-of-flight secondary ion mass spectroscopy (ToF-SIMS), atom probe tomography (APT) and transmission electron microscopy (TEM) is compared. Their effectiveness and short-comings in determining carbon contamination will be presented and discussed. The corresponding microstructural changes related to carbon contamination in ion irradiated F/M steels are also presented and briefly discussed.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. Proceedings of the second milestone meeting of European laboratories on the development of ferritic/martensitic steels for fusion technology

    International Nuclear Information System (INIS)

    Daum, E.; Ehrlich, K.; Schirra, M.

    1997-05-01

    In the frame of the European Fusion Technology Program a series of ferritic/martensitic developmental alloys, the composition of which had been optimized towards low long-term activation, was investigated and compared with conventional 9-12%CrMoVNb steels. It could be shown that by these chemical modifications neither the physical metallurgy nor the transformation behavior was changed markedly. Tensile-, creep-rupture- and fatigue properties are somewhat reduced, whereas the fracture toughness and impact data are far superior to conventional materials. This is an important advantage, especially if the expected detrimental effect of neutron irradiation on the latter properties is taken into account. First results of low-fluence irradiations indicate that the new alloys are less prone to irradiation-induced DBTT shifts. (orig./HM) [de

  15. Stress effects in cylindrical tubes of austenitic and ferritic/martensitic steels with oxide scales. Materials selection for a HPLWR

    International Nuclear Information System (INIS)

    Steiner, H.

    2002-11-01

    In the frame of the studies for a high performance concept of a light water reactor (LWR) different materials for the cladding are investigated, among them are austenitic and ferritic/martensitic (f/m) steels of different Cr content. Due to the envisaged very extended life times of the fuel elements in the reactor, corrosion problems may arise. Thus, cracking and/or spalling effects in oxide scales on metallic components may play an important role in the corrosion process as they lead, in general, to a drastic enhancement in the oxidation rates. Analytical models for different fundamental stress problems in the compound oxide scale/metallic substrate have been developed and implemented in the computer code OXSPA. These models concern the growth stresses in the cylindrical tubes, the stresses due to temperature changes and radial temperature gradients and the stresses due to inside and outside pressures. (orig.)

  16. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-01

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

  17. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  18. Effect of pre-strain on susceptibility of Indian Reduced Activation Ferritic Martensitic Steel to hydrogen embrittlement

    International Nuclear Information System (INIS)

    Sonak, Sagar; Tiwari, Abhishek; Jain, Uttam; Keskar, Nachiket; Kumar, Sanjay; Singh, Ram N.; Dey, Gautam K.

    2015-01-01

    The role of pre-strain on hydrogen embrittlement susceptibility of Indian Reduced Activation Ferritic Martensitic Steel was investigated using constant nominal strain-rate tension test. The samples were pre-strained to different levels of plastic strain and their mechanical behavior and mode of fracture under the influence of hydrogen was studied. The effect of plastic pre-strain in the range of 0.5–2% on the ductility of the samples was prominent. Compared to samples without any pre-straining, effect of hydrogen was more pronounced on pre-strained samples. Prior deformation reduced the material ductility under the influence of hydrogen. Up to 35% reduction in the total strain was observed under the influence of hydrogen in pre-strained samples. Hydrogen charging resulted in increased occurrence of brittle zones on the fracture surface. Hydrogen Enhanced Decohesion (HEDE) was found to be the dominant mechanism of fracture.

  19. Comparison of the corrosion behavior of austenitic and ferritic/martensitic steels exposed to static liquid Pb Bi at 450 and 550 °C

    Science.gov (United States)

    Kurata, Y.; Futakawa, M.; Saito, S.

    2005-08-01

    Static corrosion tests of various steels were conducted in oxygen-saturated liquid Pb-Bi eutectic at 450 °C and 550 °C for 3000 h to study the effects of temperature and alloying elements on corrosion behavior in liquid Pb-Bi. Corrosion depth decreases at 450 °C with increasing Cr content in steels regardless of ferritic/martensitic steels or austenitic steels. Appreciable dissolution of Ni and Cr does not occur in the three austenitic steels at 450 °C. Corrosion depth of ferritic/martensitic steels also decreases at 550 °C with increasing Cr content in steels whereas corrosion depth of austenitic steels, JPCA and 316SS becomes larger due to ferritization caused by dissolution of Ni at 550 °C than that of ferritic/martensitic steels. An austenitic stainless steel containing about 5%Si exhibits fine corrosion resistance at 550 °C because the protective Si oxide film is formed and prevents dissolution of Ni and Cr.

  20. Comparison of the corrosion behavior of austenitic and ferritic/martensitic steels exposed to static liquid Pb-Bi at 450 and 550 deg. C

    International Nuclear Information System (INIS)

    Kurata, Y.; Futakawa, M.; Saito, S.

    2005-01-01

    Static corrosion tests of various steels were conducted in oxygen-saturated liquid Pb-Bi eutectic at 450 deg. C and 550 deg. C for 3000 h to study the effects of temperature and alloying elements on corrosion behavior in liquid Pb-Bi. Corrosion depth decreases at 450 deg. C with increasing Cr content in steels regardless of ferritic/martensitic steels or austenitic steels. Appreciable dissolution of Ni and Cr does not occur in the three austenitic steels at 450 deg. C. Corrosion depth of ferritic/martensitic steels also decreases at 550 deg. C with increasing Cr content in steels whereas corrosion depth of austenitic steels, JPCA and 316SS becomes larger due to ferritization caused by dissolution of Ni at 550 deg. C than that of ferritic/martensitic steels. An austenitic stainless steel containing about 5%Si exhibits fine corrosion resistance at 550 deg. C because the protective Si oxide film is formed and prevents dissolution of Ni and Cr

  1. High Temperature Oxidation Behavior of T91 Steel in Dry and Humid Condition

    Directory of Open Access Journals (Sweden)

    Yonghao Leong

    2016-09-01

    Full Text Available High temperature oxidation behavior of T91 ferritic/martensitic steel was examined over the temperature range of 500 to 700°C in dry and humid environments.  The weight gain result revealed that oxidation occurs at all range of temperatures and its rate is accelerated by increasing the temperature. The weight gain of the oxidized steel at 700°C in steam condition was six times bigger than the dry oxidation.. SEM/EDX of the cross-sectional image showed that under dry condition, a protective and steady growth of the chromium oxide (Cr2O3 layer was formed on the steel with the thickness of 2.39±0.34 µm. Meanwhile for the humid environment, it is found that the iron oxide layer, which consists of the hematite (Fe2O3 and magnetite (Fe3O4 was formed as the outer scale, and spinnel as inner scale. This result indicated that the oxidation behavior of T91 steel was affected by its oxidation environment. The existence of water vapor in steam condition may prevent the formation of chromium oxide as protective layer.

  2. Summary Report of Summer Work: High Purity Single Crystal Growth & Microstructure of Ferritic-Martensitic Steels

    Energy Technology Data Exchange (ETDEWEB)

    Pestovich, Kimberly Shay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-18

    Harnessing the power of the nuclear sciences for national security and to benefit others is one of Los Alamos National Laboratory’s missions. MST-8 focuses on manipulating and studying how the structure, processing, properties, and performance of materials interact at the atomic level under nuclear conditions. Within this group, single crystal scintillators contribute to the safety and reliability of weapons, provide global security safeguards, and build on scientific principles that carry over to medical fields for cancer detection. Improved cladding materials made of ferritic-martensitic alloys support the mission of DOE-NE’s Fuel Cycle Research and Development program to close the nuclear fuel cycle, aiming to solve nuclear waste management challenges and thereby increase the performance and safety of current and future reactors.

  3. Development of new ferritic / martensitic steels for fuel cladding in fast neutron reactors

    International Nuclear Information System (INIS)

    Ratti, M.

    2009-11-01

    Many studies are directed toward the development of ferritic / martensitic ODS materials for applications in Gen IV programs. In this study, the mechanisms of formation of nano-phases (Y, Ti, O) and the influence of titanium on the precipitation refinement have been analyzed by small angle neutron scattering, X-ray diffraction and neutron diffraction. The obtained results allow developing new materials reinforced by nitrides (NDS which stands for Nitride Dispersion Strengthened). A first CEA patent is now being registered on these NDS materials processed by mechanical alloying. However, microstructural and mechanical characterizations are necessary to improve these new alloys. At last, a tensile and creep database has been acquired on an ODS Fe-18Cr material between room temperature and 650 C. These tests allow a qualitative description of the ODS mechanical behaviour. (author)

  4. Influence of strain rate and temperature on tensile properties and flow behaviour of a reduced activation ferritic-martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Vanaja, J., E-mail: jvanaja4@gmail.com [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Laha, K. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Sam, Shiju [Institute for Plasma Research, Gandhinagar, Gujarat (India); Nandagopal, M.; Panneer Selvi, S.; Mathew, M.D.; Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Rajendra Kumar, E. [Institute for Plasma Research, Gandhinagar, Gujarat (India)

    2012-05-15

    Tensile strength and flow behaviour of a Reduced Activation Ferritic-Martensitic (RAFM) steel (9Cr-1W-0.06Ta-0.22V-0.08C) have been investigated over a temperature range of 300-873 K at different strain rates. Tensile strength of the steel decreased with temperature and increased with strain rate except at intermediate temperatures. Negative strain rate sensitivity of flow stress of the steel at intermediate temperatures revealed the occurrence of dynamic strain ageing in the steel, even though no serrated flow was observed. The tensile flow behaviour of the material was well represented by the Voce strain hardening equation for all the test conditions. Temperature and strain rate dependence of the various parameters of Voce equation were interpreted with the possible deformation mechanisms. The equivalence between the saturation stress at a given strain rate in tensile test and steady state deformation rate at a given stress in creep test was found to be satisfied by the RAFM steel.

  5. Effect of tungsten and tantalum on the low cycle fatigue behavior of reduced activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Shankar, Vani, E-mail: vani@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam-603102 (India); Mariappan, K.; Nagesha, A.; Prasad Reddy, G.V.; Sandhya, R.; Mathew, M.D.; Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam-603102 (India)

    2012-05-15

    Highlights: Black-Right-Pointing-Pointer Effect of tungsten and tantalum on low cycle fatigue behavior of RAFM steels. Black-Right-Pointing-Pointer Both alloying elements W and Ta improved fatigue life. Black-Right-Pointing-Pointer Increase in Ta content improved fatigue life more than W. Black-Right-Pointing-Pointer Optimization of W content at 1.4 wt.%. Black-Right-Pointing-Pointer Softening behavior closely related to W and Ta content. - Abstract: Reduced activation ferritic/martensitic (RAFM) steels are candidate materials for the test blanket modules of International Thermonuclear Experimental Reactor (ITER). Several degradation mechanisms such as thermal fatigue, low cycle fatigue, creep fatigue interaction, creep, irradiation hardening, swelling and phase instability associated irradiation embrittlement must be understood in order to estimate the component lifetime and issues concerning the structural integrity of components. The current work focuses on the effect of tungsten and tantalum on the low cycle fatigue (LCF) behavior of RAFM steels. Both alloying elements tungsten and tantalum improved the fatigue life. Influence of Ta on increasing fatigue life was an order of magnitude higher than the influence of W on improving the fatigue life. Based on the present study, the W content was optimized at 1.4 wt.%. Softening behavior of RAFM steels showed a strong dependence on W and Ta content in RAFM steels.

  6. Influence of strain rate and temperature on tensile properties and flow behaviour of a reduced activation ferritic-martensitic steel

    Science.gov (United States)

    Vanaja, J.; Laha, K.; Sam, Shiju; Nandagopal, M.; Panneer Selvi, S.; Mathew, M. D.; Jayakumar, T.; Rajendra Kumar, E.

    2012-05-01

    Tensile strength and flow behaviour of a Reduced Activation Ferritic-Martensitic (RAFM) steel (9Cr-1W-0.06Ta-0.22V-0.08C) have been investigated over a temperature range of 300-873 K at different strain rates. Tensile strength of the steel decreased with temperature and increased with strain rate except at intermediate temperatures. Negative strain rate sensitivity of flow stress of the steel at intermediate temperatures revealed the occurrence of dynamic strain ageing in the steel, even though no serrated flow was observed. The tensile flow behaviour of the material was well represented by the Voce strain hardening equation for all the test conditions. Temperature and strain rate dependence of the various parameters of Voce equation were interpreted with the possible deformation mechanisms. The equivalence between the saturation stress at a given strain rate in tensile test and steady state deformation rate at a given stress in creep test was found to be satisfied by the RAFM steel.

  7. Microstructural characterization of weld joints of 9Cr reduced activation ferritic martensitic steel fabricated by different joining methods

    Energy Technology Data Exchange (ETDEWEB)

    Thomas Paul, V.; Saroja, S.; Albert, S.K.; Jayakumar, T.; Rajendra Kumar, E., E-mail: vtp@igcar.gov.in

    2014-10-15

    This paper presents a detailed electron microscopy study on the microstructure of various regions of weldment fabricated by three welding methods namely tungsten inert gas welding, electron beam welding and laser beam welding in an indigenously developed 9Cr reduced activation ferritic/martensitic steel. Electron back scatter diffraction studies showed a random micro-texture in all the three welds. Microstructural changes during thermal exposures were studied and corroborated with hardness and optimized conditions for the post weld heat treatment have been identified for this steel. Hollomon–Jaffe parameter has been used to estimate the extent of tempering. The activation energy for the tempering process has been evaluated and found to be corresponding to interstitial diffusion of carbon in ferrite matrix. The type and microchemistry of secondary phases in different regions of the weldment have been identified by analytical transmission electron microscopy. - Highlights: • Comparison of microstructural parameters in TIG, electron beam and laser welds of RAFM steel • EBSD studies to illustrate the absence of preferred orientation and identification of prior austenite grain size using phase identification map • Optimization of PWHT conditions for indigenous RAFM steel • Study of kinetics of tempering and estimation of apparent activation energy of the process.

  8. Study of the first stages of oxidation of a ferritic-martensitic steel Fe-12Cr in CO2

    International Nuclear Information System (INIS)

    Bouhieda, S.

    2012-01-01

    In the framework of the development of Sodium Fast Reactors in France, supercritical carbon dioxide integrated in the Brayton cycle is proposed as new cycle energy conversion system to replace current steam generators. Ferritic-Martensitic steels with 9-12 wt% Cr are good candidates for heat exchanger application because they have good mechanical properties up to a temperature of 600 C, a high thermal conductivity, a low coefficient of thermal expansion and a lower cost than that of austenitic steels. However, it has been found that these steels present a high parabolic oxide growth rate and a strong carburization in the temperature and pressure conditions of the SC-CO 2 cycle (550 C, 250 bar). This study aims to investigate the influence of different parameters (impurities present in CO 2 , thermal ramp rate and surface state) on the oxidation mechanism of a Fe-12 Cr steel in CO 2 at 550 C. It has been shown that depending on these parameters, a thin protective oxide scale without any strong carburization can be obtained. A model is proposed to explain the experimental results. (author) [fr

  9. Small punch tensile/fracture test data and 3D specimen surface data on Grade 91 ferritic/martensitic steel from cryogenic to room temperature.

    Science.gov (United States)

    Bruchhausen, Matthias; Lapetite, Jean-Marc; Ripplinger, Stefan; Austin, Tim

    2016-12-01

    Raw data from small punch tensile/fracture tests at two displacement rates in the temperature range from -196 °C to room temperature on Grade 91 ferritic/martensitic steel are presented. A number of specimens were analyzed after testing by means of X-ray computed tomography (CT). Based on the CT volume data detailed 3D surface maps of the specimens were established. All data are open access and available from Online Data Information Network (ODIN)https://odin.jrc.ec.europa.eu. The data presented in the current work has been analyzed in the research article "On the determination of the ductile to brittle transition temperature from small punch tests on Grade 91 ferritic-martensitic steel" (M. Bruchhausen, S. Holmström, J.-M. Lapetite, S. Ripplinger, 2015) [1].

  10. Fracture toughness of the IEA heat of F82H ferritic/martensitic stainless steel as a function of loading mode

    Energy Technology Data Exchange (ETDEWEB)

    Li, Huaxin; Gelles, D.S. [Pacific Northwest Labs., Richland, WA (United States); Hirth, J.P. [Washington State Univ., Pullman, WA (United States)] [and others

    1997-04-01

    Mode I and mixed-mode I/III fracture toughness tests were performed for the IEA heat of the reduced activation ferritic/martensitic stainless steel F82H at ambient temperature in order to provide comparison with previous measurements on a small heat given a different heat treatment. The results showed that heat to heat variations and heat treatment had negligible consequences on Mode I fracture toughness, but behavior during mixed-mode testing showed unexpected instabilities.

  11. Development of ODS ferritic-martensitic steels for application to high temperature and irradiation environment

    International Nuclear Information System (INIS)

    Lambard, V.

    2000-01-01

    Iron oxide dispersion strengthened alloys are candidate for nuclear fuel cladding. Therefore, it is crucial to control their microstructure in order to optimise their mechanical properties at temperatures up to 700 deg C. The industrial candidates, ODS ferritic alloys, present an anisotropic microstructure which induces a weakening of mechanical properties in transversal direction as well as the precipitation of brittle phases under thermal aging and irradiation. For this purpose, we tried to develop a material with isotropic properties. We studied several 9Cr-1Mo ferritic/martensitic alloys, strengthened or not by oxide dispersion. The mechanical alloying was performed by attribution and powders were consolidated by hot extrusion. In this work, different metallurgical characterisation techniques and modelling were used to optimise a new martensitic ODS alloy. Microstructural and chemical characterization of matrix has been done. The effect of austenitizing and isochronal tempering treatments on microstructure and hardness has been studied. Oxide distribution, size and chemical composition have been studied before and after high temperature thermal treatment. The study of phase transformation upon heating has permitted the extrapolation to the equilibrium temperature formation of austenite. Phase transformation diagrams upon cooling have been determined and the transformation kinetics have been linked to austenite grain size by a simple relation. Fine grain size is unfavourable for the targeted application, so a particular thermal treatment inducing a coarser grain structure has been developed. Finally, tensile properties have been determined for the different microstructures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Tan, L., E-mail: tanl@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tavassoli, A.-A.F.; Henry, J. [DMN/Dir, DEN, CEA Saclay, 91191, Gif-sur-Yvette Cedex (France); Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe, 76021 (Germany); Sakasegawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tanigawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2016-10-15

    Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. In addition to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M{sub 23}C{sub 6} (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. Limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic are presented and compared with data for F82H and Eurofer97 irradiated up to ∼70 displacements per atom at ∼300–325 °C.

  13. Extraction residue analysis on F82H-BA07 heat and other reduced activation ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Nagasaka, Takuya; Hishinuma, Yoshimitsu; Muroga, Takeo; Li, Yanfen; Watanabe, Hideo; Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ando, Masami

    2011-01-01

    Extraction residue analysis was conducted on reduced activation ferritic/martensitic steels, such as F82H-BA07 heat, F82H-IEA heat, JLF-1 JOYO heat and CLAM steel. M 23 C 6 type precipitates, TaC precipitates and Fe 2 W Laves phase were identified in the present analyses. M 23 C 6 precipitates were coarsened in F82H-BA07 compared with the other steels at as-normalized and tempered (NT) condition. TaC precipitate formation was enhanced in JLF-1 and CLAM compared with F82H-BA07 and F82H-IEA at as-NT condition. Laves phase were detected in F82H-IEA after aging above 550 o C, where solid solution W was significantly decreased. F82H-IEA exhibited hardening after aging at 400 and 500 o C for 100 khr, whereas softening at 600 and 650 o C. This behavior is similar to JLF-1 and CLAM, and can be understood by precipitation of TaC and Laves phase.

  14. Microstructure anisotropy and its effect on mechanical properties of reduced activation ferritic/martensitic steel fabricated by selective laser melting

    Science.gov (United States)

    Huang, Bo; Zhai, Yutao; Liu, Shaojun; Mao, Xiaodong

    2018-03-01

    Selective laser melting (SLM) is a promising way for the fabrication of complex reduced activation ferritic/martensitic steel components. The microstructure of the SLM built China low activation martensitic (CLAM) steel plates was observed and analyzed. The hardness, Charpy impact and tensile testing of the specimens in different orientations were performed at room temperature. The results showed that the difference in the mechanical properties was related to the anisotropy in microstructure. The planer unmelted porosity in the interface of the adjacent layers induced opening/tensile mode when the tensile samples parallel to the build direction were tested whereas the samples vertical to the build direction fractured in the shear mode with the grains being sheared in a slant angle. Moreover, the impact absorbed energy (IAE) of all impact specimens was significantly lower than that of the wrought CLAM steel, and the IAE of the samples vertical to the build direction was higher than that of the samples parallel to the build direction. The impact fracture surfaces revealed that the load parallel to the build layers caused laminated tearing among the layers, and the load vertical to the layers induced intergranular fracture across the layers.

  15. Some initial considerations on the suitability of Ferritic/ martensitic stainless steels as first wall and blanket materials in fusion reactors

    International Nuclear Information System (INIS)

    Butterworth, G.J.

    1982-01-01

    The constitution of stainless iron alloys and the characteristic properties of alloys in the main ferritic, martensitic and austenitic groups are discussed. A comparison of published data on the mechanical, thermal and irradiation properties of typical austenitic and martensitic/ferritic steels shows that alloys in the latter groups have certain advantages for fusion applications. The ferromagnetism exhibited by martensitic and ferritic alloys has, however, been identified as a potentially serious obstacle to their utilisation in magnetic confinement devices. The paper describes measurements performed in other laboratories on the magnetic properties of two representative martensitic alloys 12Cr-1Mo and 9Cr-2Mo. These observations show that a modest bias magnetic field of magnitude 1 - 2 tesla induces a state of magnetic saturation in these materials. They would thus behave as essentially paramagnetic materials having a relative permeability close to unity when saturated by the toroidal field of a tokamak reactor. The results of computations by the General Atomic research group to assess the implications of such magnetic behaviour on reactor design and operation are presented. The results so far indicate that the ferromagnetism of martensitic/ferritic steels would not represent a major obstacle to their utilisation as first wall or blanket materials. (author)

  16. Low cycle fatigue design data for India-specific reduced activation ferritic-martensitic (IN-RAFM) steel

    Energy Technology Data Exchange (ETDEWEB)

    Mariappan, K.; Shankar, Vani, E-mail: vani@igcar.gov.in; Sandhya, R.; Laha, K.

    2016-03-15

    Highlights: • Generation of first set of experimental data related to LCF performance of the commercial heat of IN-RAFM steel. • Analysis of cyclic behavior from the perspective of both design and material characteristics. • Various correction factors to account for various plastic strain accumulations, change in Poisson’s ratio and asymmetry of loadings. • Low cycle fatigue design parameters and correction factor values were comparable with P91 steel as reported in RCC-MR design code. - Abstract: The objective of the present paper is to provide first hand experimental data and analysis on the low cycle fatigue (LCF) performance of a commercial heat of Indian reduced activation ferritic-martensitic (IN-RAFM) steel. Since this material is not yet codified in RCC-MR, cyclic properties were generated for the design of the structural material of the Test Blanket Modules (TBM) made of RAFM steel. Hence, as a part of the material development program, LCF experiments were conducted on IN-RAFM steel obtained in the normalized and tempered condition. Total axial strain controlled experiments were performed in air by employing strain amplitudes ranging from ±0.25 to ±1.0% and at temperatures of 300, 673, 723, 823, and 873 K and a nominal strain rate, 3 × 10{sup −3} s{sup −1}. In the present work, various cyclic parameters that are useful for the design oriented fatigue analysis are derived as per the systematic procedure given in the RCC-MR design code. The physical significance of each design parameter such as elasto-plastic corrections based on Neuber analysis has been explained and correlated with the material behavior such as the cyclic softening nature of the RAFM steel.

  17. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    International Nuclear Information System (INIS)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-01-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 .deg. C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 .deg. C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI)

  18. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Science.gov (United States)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen; Moon, Se Youn; Hong, Bong Guen

    2014-10-01

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 °C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 °C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI).

  19. Fabrication and integrity test preparation of HIP-joined W and ferritic-martensitic steel mockups for fusion reactor development

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Shin, Kyu In; Kim, Suk Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Choi, Bo Guen [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Moon, Se Youn; Hong, Bong Guen [Chonbuk National University, Jeonju (Korea, Republic of)

    2014-10-15

    Tungsten (W) and ferritic-martensitic steel (FMS) as armor and structural materials, respectively, are the major candidates for plasma-facing components (PFCs) such as the blanket first wall (BFW) and the divertor, in a fusion reactor. In the present study, three W/FMS mockups were successfully fabricated using a hot isostatic pressing (HIP, 900 .deg. C, 100 MPa, 1.5 hrs) with a following post-HIP heat treatment (PHHT, tempering, 750 .deg. C, 70 MPa, 2 hrs), and the W/FMS joining method was developed based on the ITER BFW and the test blanket module (TBM) development project from 2004 to the present. Using a 10-MHz-frequency flat-type probe to ultrasonically test of the joint, we found no defects in the fabricated mockups. For confirmation of the joint integrity, a high heat flux test will be performed up to the thermal lifetime of the mockup under the proper test conditions. These conditions were determined through a preliminary analysis with conventional codes such as ANSYS-CFX for thermal-hydraulic conditions considering the test facility, the Korea heat load test facility with an electron beam (KoHLT-EB), and its water coolant system at the Korea Atomic Energy Research Institute (KAERI)

  20. Mechanical properties of 9Cr–1W reduced activation ferritic martensitic steel weldment prepared by electron beam welding process

    Energy Technology Data Exchange (ETDEWEB)

    Das, C.R., E-mail: chitta@igcar.gov.in [Indira Gandhi Center for Atomic Research, Kalpakkam 603102 (India); Albert, S.K. [Indira Gandhi Center for Atomic Research, Kalpakkam 603102 (India); Sam, Shiju [Institute for Plasma Research, Gandhinagar (India); Mastanaiah, P. [Defense Research and Development Laboratory, Hyderabad (India); Chaitanya, G.M.S.K.; Bhaduri, A.K.; Jayakumar, T. [Indira Gandhi Center for Atomic Research, Kalpakkam 603102 (India); Murthy, C.V.S. [Defense Research and Development Laboratory, Hyderabad (India); Kumar, E. Rajendra [Institute for Plasma Research, Gandhinagar (India)

    2014-11-15

    Highlights: • Width of HAZ is smaller in the 9Cr–1W RAFM weldment prepared by EB process compared to that reported for TIG weldments in literature. • Weld joint is stronger than that of the base metal. • Toughness of weld metal prepared by EB welding process is comparable to that (in PWHT condition) prepared by TIG process. • DBTT of as-welded 9Cr–1W RAFM weldment prepared by EB process is comparable to that reported for TIG weld metal in PWHT condition. - Abstract: Microstructure and mechanical properties of the weldments prepared from 9Cr–1W reduced activation ferritic martensitic (RAFM) steel using electron beam welding (EBW) process were studied. Microstructure consists of tempered lath martensite where precipitates decorating the boundaries in post weld heat treated (PWHT) condition. Lath and precipitate sizes were found to be finer in the weld metal than in base metal. Accordingly, hardness of the weld metal was found to be higher than the base metal. Tensile strength of the cross weldment specimen was 684 MPa, which was comparable with the base metal tensile strength of 670 MPa. On the other hand, DBTT of 9Cr–1W weld metal in as-welded condition is similar to that reported for TIG weld metal in PWHT condition.

  1. Mechanical properties of 9Cr–1W reduced activation ferritic martensitic steel weldment prepared by electron beam welding process

    International Nuclear Information System (INIS)

    Das, C.R.; Albert, S.K.; Sam, Shiju; Mastanaiah, P.; Chaitanya, G.M.S.K.; Bhaduri, A.K.; Jayakumar, T.; Murthy, C.V.S.; Kumar, E. Rajendra

    2014-01-01

    Highlights: • Width of HAZ is smaller in the 9Cr–1W RAFM weldment prepared by EB process compared to that reported for TIG weldments in literature. • Weld joint is stronger than that of the base metal. • Toughness of weld metal prepared by EB welding process is comparable to that (in PWHT condition) prepared by TIG process. • DBTT of as-welded 9Cr–1W RAFM weldment prepared by EB process is comparable to that reported for TIG weld metal in PWHT condition. - Abstract: Microstructure and mechanical properties of the weldments prepared from 9Cr–1W reduced activation ferritic martensitic (RAFM) steel using electron beam welding (EBW) process were studied. Microstructure consists of tempered lath martensite where precipitates decorating the boundaries in post weld heat treated (PWHT) condition. Lath and precipitate sizes were found to be finer in the weld metal than in base metal. Accordingly, hardness of the weld metal was found to be higher than the base metal. Tensile strength of the cross weldment specimen was 684 MPa, which was comparable with the base metal tensile strength of 670 MPa. On the other hand, DBTT of 9Cr–1W weld metal in as-welded condition is similar to that reported for TIG weld metal in PWHT condition

  2. Triple ion-beam studies of radiation damage in 9Cr2WVTa ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Lee, E.H.; Hunn, J.D.; Rao, G.R.; Klueh, R.L.; Mansur, L.K.

    1997-01-01

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

  3. Boron-bearing Influences of 9Cr-0.5Mo-2W-V-Nb Ferritic/Martensitic Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Woo-Gon; Kim, Sung-Ho; Lee, Chan-Bock

    2008-01-01

    Currently the principal materials in a SFR (sodium-cooled fast reactor) of Gen-IV nuclear system are considering stainless steels (e.g. austenitic steels and ferritic/martensitic steels) for pressure boundary and structural applications in the primary circuit (cladding, duct, cold and hot leg piping, and pressure vessel). There are sound technical justifications for these material selections, and the adoption of these stainless steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for the Gen-IV SFR. The Gen-IV SFR is to have a considerable increase in safety and be economically competitive when compared with the conventional water reactors. To accomplish these objectives, the development of the fuel cladding material should be set forth as a premise because its integrity is directly related to those of the reactor system as well as the fuel in the Gen-IV SFR. Since last year, a R and D program was launched to develop the improved ferritic/martensitic steel for the Gen-IV SFR fuel cladding. Categories of materials considered in the program included 8 - 12% Cr ferritic/ martensitic steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel to offset the difficulties encountered with commercial available steels of the 8 - 12% Cr group. That is, since fuel cladding in the Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the cladding should thus sustain both superior irradiation and temperature stabilities during an operational life. The newly developed advanced steel should overcome the severe drawback; mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. In this study, as one of the composition

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

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang, E-mail: thaksang.byun@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Hoelzer, David T. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kim, Jeoung Han [Hanbat National University, Daejeon 305-719 (Korea, Republic of); Maloy, Stuart A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2017-02-15

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

  5. Investigating liquid-metal embrittlement of T91 steel by fracture toughness tests

    Energy Technology Data Exchange (ETDEWEB)

    Ersoy, Feyzan, E-mail: fersoy@sckcen.be [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400, Mol (Belgium); Department of Materials Science and Engineering, Ghent University (UGent), Technologiepark 903, B-9052, Ghent (Belgium); Gavrilov, Serguei [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400, Mol (Belgium); Verbeken, Kim [Department of Materials Science and Engineering, Ghent University (UGent), Technologiepark 903, B-9052, Ghent (Belgium)

    2016-04-15

    Heavy liquid metals such as lead bismuth eutectic (LBE) are chosen as the coolant to innovative Generation IV (Gen IV) reactors where ferritic/martensitic T91 steel is a candidate material for high temperature applications. It is known that LBE has a degrading effect on the mechanical properties of this steel. This degrading effect, which is known as liquid metal embrittlement (LME), has been screened by several tests such as tensile and small punch tests, and was most severe in the temperature range from 300 °C to 425 °C. To meet the design needs, mechanical properties such as fracture toughness should be addressed by corresponding tests. For this reason liquid-metal embrittlement of T91 steel was investigated by fracture toughness tests at 350 °C. Tests were conducted in Ar-5%H{sub 2} and LBE under the same experimental conditions Tests in Ar-5%H{sub 2} were used as reference. The basic procedure in the ASTM E 1820 standard was followed to perform tests and the normalization data reduction (NDR) method was used for the analysis. Comparison of the tests demonstrated that the elastic–plastic fracture toughness (J{sub 1C}) of the material was reduced by a factor in LBE and the fracture mode changed from ductile to quasi-cleavage. It was also shown that the pre-cracking environment played an important role in observing LME of the material since it impacts the contact conditions between LBE and steel at the crack tip. It was demonstrated that when specimens were pre-cracked in air and tested in LBE, wetting of the crack surface by LBE could not be achieved. When specimens were pre-cracked in LBE though, they showed a significant reduction in fracture toughness.

  6. Impact properties of reduced activation ferritic/martensitic steel, F82H jointed by hot isostatic pressing

    Energy Technology Data Exchange (ETDEWEB)

    Ogiwara, H.; Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Ibaraki-ken (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

    2007-07-01

    Full text of publication follows: Reduced-activation ferritic/martensitic steels are the leading candidate structural material for the blanket system of fusion reactors. The important issue at the current stage is the finalization of a detailed manufacturing specification for ITER test blanket module. Hot isostatic pressing (HIP) process is one of the most important methods to fabricate the first wall with cooling channels. The objective of this paper is to optimize HIP condition to obtain the excellent joints mechanical properties. The materials used were F82H steels. The joint was produced by solid state HIP method. Before HIP treatments, specimens were heated in vacuum condition to out-gas. This treatment was conducted to decrease oxidation on the surfaces. HIP treatments were carried out for 2 h at 1100 deg. C - 140 MPa. The specimens were normalized at 960 deg. C for 0.5 h and tempered at 750 deg. C for 1.5 h. The bonding interface was characterized by scanning electron microscopy. Charpy impact tests and tensile tests were conducted to evaluate the mechanical properties of the HIP joint. Impact tests revealed that there were no significant differences in the ductile-brittle transition temperatures of HIP jointed specimens and base metal specimens, but the upper-shelf energy (USE) of the HIP joint specimens at room temperature was only about 10% of that of the base metal specimens. SEM observations of the fracture surface of HIP joint specimens revealed that a large number of oxides were formed on the HIP joint. This result indicates that oxides formed on the HIP joint are the dominant factor of the impact properties. Based on these results, the pre-HIP treatment conditions had been optimized to reduce the number of oxides, and USE of HIP joint specimens increased to about 50% of that of the base metal. The detailed analyses on the HIP joint microstructure will be reported. (authors)

  7. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2010-01-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  8. Production and qualification for fusion applications, a steel of low activity ferritic-martensitic ASTURFER; Produccion y cualificacion, para aplicaciones de fusion, de un acero de baja actividad ferritico-martensitico, ASTURFER

    Energy Technology Data Exchange (ETDEWEB)

    Moran, A.; Belzunce, J.; Artimez, J. M.

    2011-07-01

    This article details the work carried out in the design and development pilot plant scale of a steel ferritic-martensitic of reduced activity, Asturfer, with a chemical composition and metallurgical properties similar to steel Eurofer. We describe the different stages of steel production and the results of the characterizations made in the context of an extensive test program.

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

    Science.gov (United States)

    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.

  10. Effects of impurity elements on mechanical properties and microstructures of reduced-activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sawahata, A. [Ibaraki Univ., Graduate School of Science and Engineering, Hitachi (Japan); Tanigawa, H.; Shiba, K. [Japan Atomic Energy Agency, Naga-gun, Ibaraki-ken (Japan); Enomoto, M. [Ibaraki Univ., Dept. of Materials Science, Faculty of Engineering, Hitachi (Japan)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFs), such as F82H (Fe-8Cr-2W-0.2V- 0.04Ta-0.1C, in wt%), are one of the leading candidates for structural materials of fusion reactors. Impact property of F82H can be improved by adjusting the amount of tantalum or titanium concentration. On the other hand, it was reported by microstructure analyses of IEA steel that tantalum has a tendency to form oxides and causes a large dispersion of fracture toughness. In this study, the correlation between titanium or tantalum concentration and the impact property were reported focusing on difference in microstructure. Charpy impact test and microstructure analyses were carried out against modified F82H series of which titanium, nitrogen and tantalum composition were controlled. Charpy impact test results showed that the ductile-brittle transition temperature (DBTT) of T05A (0.05Ta- 0.0014N-steels. The size distribution analyses of oxides indicate that the number density of composite oxides in T05B was higher than in T05A. In addition, EDX analyses showed that composite oxides in T05B had a strong peak of titanium, but the peak were not detected in the oxides in T05A. These results suggest that titanium has a significant influence on the formation of oxides, and affects the impact property. The influence of tantalum concentration on the formation of these oxides and mechanical properties will be reported. (authors)

  11. Effect of precipitate on yield strength of ferritic/martensitic steel exposed to 650 .deg. C liquid sodium

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Yong; Lee, Jeonghyeon; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of); Shin, Sang Hun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Ferritic/martensitic steels(FMS) which are used as one of cladding and structure materials in UCFR, have high creep strength at 600~650°C, low expansion coefficient, and dimensional stability with irradiation-induced void swelling in circumstance of fast neutrons compared to austenitic stainless steel. However, as exposed to high temperature liquid sodium during the design life time (30 to 60 years), the surface of FMS experienced Cr-depletion and decarburization by dissolution of components into sodium and formed oxidations by reacting with sodium. This changes chemical compositions of inter-surface and effects on behavior of precipitations. This change can cause a degradation of mechanical strength of structure material of UCFR. The research about FMS on effects of long term exposure in liquid sodium at 650 °C involve analysis of yield strength by change of microstructure, solid solution hardening and precipitation hardening. It shows how this three parts occupy total yield strength respectively and change over time. In a specific procedure, the microstructure and the surface phenomenon of FMS (Gr. 92) that are exposed to liquid sodium at 650°C, 20 ppm oxygen and are aged in high pure Argon gas environment to express bulk have been investigated by using scanning electron microscope (SEM) and transmission electron microscope (TEM). When specimens were exposed to 650 °C liquid sodium for 1583, 3095 hours and Ar-gas 1601, 2973 hours, mechanical properties of materials were analyzed quantitatively. After experiment, NaCrO{sub 2} oxidation was formed on the surface of Gr.92 at sodium environment. Also, change of microstructure, dissolution of elements, and nucleation and growth of precipitation was raised. During exposed to high temperature at sodium or Ar-gas, annealed lath structure as well as coarsening of tempered martensite structure affects reduction of mechanical properties. And dissolution of elements results in reduction of solid solution hardening. This

  12. The influence of inclusions on the low cycle fatigue properties of reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D.H.; Kima, S.W. [Kyoto Univ., Graduate School of Energy Science (Japan); Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic (RAFM) steels, such as F82H, are the primary near-term candidate for the blanket structural material of nuclear fusion reactors. During operation, blanket structural materials will be subjected to cyclic loading caused by start-up and shut-down procedure or plasma disruption. Therefore, investigation of fatigue property is essential to reactor design. It is considered that fatigue properties depend on the material factor such as the inclusion distribution, surface morphology and so on. Especially, many experimental results show that inclusions become the fracture origin in a given volume of material subjected to cyclic stress, and fracture failure is most likely to initiate at the largest inclusion in the volume. Therefore, the prediction of the size of maximum inclusion and its impact on fatigue properties would be essential to the fusion reactor materials development and application. This paper examines the possible relation between fatigue life and inclusion parameters such as size, shape, distribution and composition. The low cycle fatigue behavior of F82H steel at room temperature in air condition under fully reversed push-pull triangular wave was studied using miniaturized hourglass-type specimens with 1.25 mm in diameter. Total strain range is selected from 0.8% to 2.4%, and the strain rate was 0.04%/s. To examine the size and composition of the inclusions, fracture surfaces and crack initiation region were investigated by a scanning electron microscope (SEM) and EDS. The inclusions such as TaO{sub x}, TaO{sub x}- Al{sub 2}O{sub 3} and Al{sub 2}O{sub 3} with the size below 10 {mu}m are observed on specimen surface. The surface observation of the specimen which discontinued testing at 20 and 500 cycle tested at the strain range of 1.4% revealed that fatigue loading induced separation of inclusions from the matrix in initial stage, then micro-crack induced around the inclusions

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

    International Nuclear Information System (INIS)

    Vasantharaja, P.; Vasudevan, M.

    2012-01-01

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

  14. Recent results of the reduced activation ferritic/martensitic steel development

    Energy Technology Data Exchange (ETDEWEB)

    Jitsukawa, S. E-mail: jitsukawa@ifmif.tokai.jaeri.go.jp; Kimura, A.; Kohyama, A.; Klueh, R.L.; Tavassoli, A.A.; Schaaf, B. van der; Odette, G.R.; Rensman, J.W.; Victoria, M.; Petersen, C

    2004-08-01

    Significant progress has been achieved in the international research effort on reduced-activation steels. Extensive tensile, fracture toughness, fatigue, and creep properties in unirradiated and irradiated conditions have been performed and evaluated. Since it is not possible to include all work in this limited review, selected areas will be presented to indicate the scope and progress of recent international efforts. These include (1) results from mechanical properties studies that have been combined in databases to determine materials design limits for the preliminary design of an ITER blanket module. (2) Results indicate that the effect of transmutation-produced helium on fracture toughness is smaller than indicated previously. (3) Further efforts to reduce irradiation-induced degradation of fracture toughness. (4) The introduction of a post-irradiation constitutive equation for plastic deformation. (5) The production of ODS steels that have been used to improve high-temperature strength. (6) The method developed to improve fracture toughness of ODS steels.

  15. Creep strength of reduced activation ferritic/martensitic steel Eurofer'97

    International Nuclear Information System (INIS)

    Fernandez, P.; Lancha, A.M.; Lapena, J.; Lindau, R.; Rieth, M.; Schirra, M.

    2005-01-01

    Creep rupture strength of tempered martensitic steel Eurofer'97 has been investigated. Different products form (plate and bar) have been tested in the temperature range from 450 deg. C to 650 deg. C at different loads. No significant differences in the creep rupture properties have been found between the studied product forms. The Eurofer'97 has shown adequate creep rupture strength levels at short creep rupture tests, similar to those of the F-82 H mod. steel. However, for long testing times (>9000 h) the results available up to now at 500 deg. C and 550 deg. C seem to indicate a change in the creep degradation mechanism

  16. Microstructural and mechanical characterization of the ferritic martensitic steel eurofer'97 after simulated service conditions

    International Nuclear Information System (INIS)

    Fernandez, P.; Lancha, A. M.; Lapena, J.

    2002-01-01

    This report details the metallurgical characterization of the Eurofer'97 steel thermally aged in the range of temperatures from 400 degree centigrade to 600 degree centigrade up to 10000 H, microstructural studies and mechanical testing (hardness, tensile, Charpy and low cycle fatigue test) have been carried out

  17. Effect of ferrite-martensite interface morphology on bake hardening response of DP590 steel

    International Nuclear Information System (INIS)

    Chakraborty, Arnab; Adhikary, Manashi; Venugopalan, T.; Singh, Virender; Nanda, Tarun; Kumar, B. Ravi

    2016-01-01

    The effect of martensite spatial distribution and its interface morphology on the bake hardening characteristics of a dual phase steel was investigated. In one case, typical industrial continuous annealing line parameters were employed to anneal a 67% cold rolled steel to obtain a dual phase microstructure. In the other case, a modified annealing process with changed initial heating rates and peak annealing temperature was employed. The processed specimens were further tensile pre-strained within 1–5% strain range followed by a bake hardening treatment at 170 °C for 20 min. It was observed that industrial continuous annealing line processed specimen showed a peak of about 70 MPa in bake-hardening index at 2% pre-strain level. At higher pre-strain values a gradual drop in bake-hardening index was observed. On the contrary, modified annealing process showed near uniform bake-hardening response at all pre-strain levels and a decrease could be noted only above 4% pre-strain. The evolving microstructure at each stage of annealing process and after bake-hardening treatment was studied using field emission scanning electron microscope. The microstructure analysis distinctly revealed differences in martensite spatial distribution and interface morphologies between each annealing processes employed. The modified process showed predominant formation of martensite within the ferrite grains with serrated lath martensite interfaces. This nature of the martensite was considered responsible for the observed improvement in the bake-hardening response. Furthermore, along with improved bake-hardening response negligible loss in tensile ductility was also noted. This behaviour was correlated with delayed micro-crack initiation at martensite interface due to serrated nature.

  18. Irradiation effects on tensile ductility and dynamic toughness of ferritic-martensitic 7-12 Cr steels

    International Nuclear Information System (INIS)

    Preininger, D.

    2006-01-01

    The superimposed effect of irradiation-induced hardening by small defects (clusters, dislocation loops) and chromium-rich - precipitate formations on tensile ductility and Charpy-impact behaviour of various ferritic-martensitic (7-13)CrWVTa(Ti)-RAFM steels have been examined by micro-mechanical deformation and ductile/dynamic fracture models. Analytical relations have been deduced describing irradiation-induced changes of uniform ductility and fracture strain as well as ductile-to-brittle transition temperature DBTT and ductile upper shelf energy USE observed from impact tests. The models apply work-hardening with competitive action of relevant dislocation multiplication and annihilation reactions. The impact model takes into account stress intensity with local plasticity and fracture within the damage zone of main crack. Especially, the influences of radiation-induced changes in ductile and dynamic fracture stresses have been considered together with effects from strain rate sensitivity of strength, precipitate morphology as mean size dp and volume fraction fv as well as deformation temperature and strain rate. For these, particularly the correlation between tensile ductility and impact properties have been examined. Strengthening by clusters and loops generally reduces uniform ductility, and more stronger fracture strain as well as ductile upper shelf energy USE and additionally increases DBTT for constant fracture stresses. A superimposed precipitation hardening by formation of 3-6 nm, f v 6 nm, which clear above the sharable limit of coherent precipitates increases with increasing fraction fv and but strongly reduces with increasing matrix strength due to full martensitic structure, higher C, N alloying contents and pronounced hardening by irradiation-induced cluster and loop formations. A combined increase of fracture stresses due to irradiation-induced changes of the grain boundary structure diminishes the strength-induced increase in DBTT and more stronger

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

    International Nuclear Information System (INIS)

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

    1993-10-01

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

  20. Effects of corrosion and precipitates on mechanical properties in the ferritic/martensitic steel cladding under ultra-long cycle fast reactor environment at 650 .deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Yong; Lee, Jeong Hyeon; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of); Shin, Sang Hun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This changes chemical compositions of inter-surface and effects on behavior of precipitations. NaCrO{sub 2} which is ternary sodium compound occurs intergranular corrosion resulting in thickness reduction. This change can cause a degradation of mechanical strength of structure material of UCFR. Therefore, we should consider longterm compatibility with sodium and study about life prediction. The research about ferritic/martensitic steel on effects of long term exposure in liquid sodium at 650 .deg. C, 20ppm oxygen includes weight loss of test material (Gr. 92) by corrosion and mechanism about nucleation and growth of precipitates like Laves-phase in bulk. There are many changes such as segregation of component to nucleate precipitates, affecting into microstructural evolution of the steel. Therefore, the thermochemical reaction research to predict behavior about precipitates should be performed. In a specific procedure, the micro-structure and the surface phenomenon of ferritic/martensitic steels (Gr. 92) that are exposed to liquid sodium at 650 .deg. C, 20 ppm oxygen and aged in high pure Argon gas environment to express bulk have been investigated by using scanning electron microscope (SEM) and transmission electron microscope (TEM). At 10 ppm oxygen designed oxygen value for UCFR, there is 107μm thickness reduction for 30 years. Thus, if there is no degradation of mechanical strength caused by aging effect, the tolerance of load of initial cladding should be higher than real load at least 23.6 %. Compared to specimens exposed to Ar-gas environment, Specimen which solutions are leaded into sodium has degradation of strength by reduction of solution hardening.

  1. Quantitative analysis of tensile deformation behavior by in-situ neutron diffraction for ferrite-martensite type dual-phase steels

    International Nuclear Information System (INIS)

    Morooka, Satoshi; Umezawa, Osamu; Harjo, Stefanus; Hasegawa, Kohei; Toji, Yuki

    2012-01-01

    The yielding and work-hardening behavior of ferrite-martensite type dual-phase (DP) alloys were clearly analyzed using the in-situ neutron diffraction technique. We successfully established a new method to estimate the stress and strain partitioning between ferrite and martensite phase during loading. Although these phases exhibit the same lattice structure with similar lattice parameters, their lattice strains on (110), (200) and (211) are obviously different from each other under an applied stress. The misfit strains between those phases were clearly accompanied with the phase-scaled internal stream (phase stress). Thus, the martensite phase yielded by higher applied stress than macro-yield stress, which resulted in high work-hardening rate of the DP steel. We also demonstrated that ferrite phase fraction influenced work-hardening behavior. (author)

  2. Effect of microstructural evolution by isothermal aging on the mechanical properties of 9Cr-1WVTa reduced activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min-Gu [Korea Institute of Materials Science, Changwon 642-831 (Korea, Republic of); Lee, Chang-Hoon, E-mail: lee1626@kims.re.kr [Korea Institute of Materials Science, Changwon 642-831 (Korea, Republic of); Moon, Joonoh; Park, Jun Young; Lee, Tae-Ho [Korea Institute of Materials Science, Changwon 642-831 (Korea, Republic of); Kang, Namhyun [Pusan National University, Busan 609-735 (Korea, Republic of); Chan Kim, Hyoung [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of)

    2017-03-15

    The influence of microstructural changes caused by aging condition on tensile and Charpy impact properties was investigated for reduced activation ferritic-martensitic (RAFM) 9Cr-1WVTa steels having single martensite and a mixed microstructure of martensite and ferrite. For the mixed microstructure of martensite and ferrite, the Charpy impact properties deteriorated in both as-normalized and tempered conditions due to the ferrite and the accompanying M{sub 23}C{sub 6} carbides at the ferrite grain boundaries which act as path and initiation sites for cleavage cracks, respectively. However, aging at 550 °C for 20–100 h recovered gradually the Charpy impact toughness without any distinct drop in strength, as a result of the spheroidization of the coarse M{sub 23}C{sub 6} carbides at the ferrite grain boundaries, which makes crack initiation more difficult.

  3. Microstructure and mechanical properties in the weld heat affected zone of 9Cr-2W-VTa reduced activation ferritic/martensitic steel for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joonoh; Lee, Changhoon; Lee, Taeho; Jang, Minho; Park, Mingu [Korea Institute of Materials Science, Changwon (Korea, Republic of); Kim, Hyoung Chan [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Reduced activation ferritic/martensitic (RAFM) steel demonstrated excellent resistance to the neutron irradiation and mechanical properties. The investigation of weldability in company with the development of RAFM steel is essential for construction of the fusion reactor. Generally, the superior mechanical properties of the RAFM steel can be upset during welding process due to microstructural change by rapid heating and cooling in the weld heat affected zone (HAZ). The phase transformation and mechanical properties in the weld HAZ of RAFM steel were investigated. The base steel consisted of tempered martensite and two carbides. During rapid welding thermal cycle, the microstructure of the base steel was transformed into martensite and δ-ferrite. In addition, the volume fraction of δ-ferrite and grain size increased with increase in the peak temperature and heat input. The strength of the HAZs was higher than that of the base steel due to the formation of martensite, whereas the impact properties of the HAZs deteriorated as compared with the base steel due to the formation of δ-ferrite. The PWHT improved the impact properties of the HAZs, resulting from the formation of tempered martensite.

  4. 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

    International Nuclear Information System (INIS)

    Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

    2003-01-01

    Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-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-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-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C 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 <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

  5. Effects of Ti and Ta addition on microstructure stability and tensile properties of reduced activation ferritic/martensitic steel for nuclear fusion reactors

    Science.gov (United States)

    Kim, Han Kyu; Lee, Ji Won; Moon, Joonoh; Lee, Chang Hoon; Hong, Hyun Uk

    2018-03-01

    The effects of Ti and Ta addition on microstructure stability and tensile properties of a reduced activation ferritic/martensitic (RAFM) steel have been investigated. Ti addition of 0.06 wt% to conventional RAFM reference base steel (Fe-9.3Cr-0.93W-0.22V-0.094Ta-0.1C) was intended to promote the precipitation of nano-sized (Ti,W) carbides with a high resistance to coarsening. In addition, the Ti addition was substituted for 0.094 wt% Ta. The Ti-added RAFM steel (Ti-RAFM) exhibited a higher yield strength (ΔYS = 32 MPa) at 600 °C than the reference base steel due to additional precipitation hardening by (Ti,W)-rich MX with an average size of 6.1 nm and the area fraction of 2.39%. However, after thermal exposure at 600 °C for 1000 h, this Ti-RAFM was more susceptible to degradation than the reference base steel; the block width increased by 77.6% in Ti-RAFM after thermal exposure while the reference base steel showed only 9.1% increase. In order to suppress diffusion rate during thermal exposure, the large-sized Ta element with low activation was added to Ti-RAFM. The Ta-added Ti-RAFM steel exhibited good properties with outstanding microstructure stability. Quantitative comparison in microstructures was discussed with a consideration of Ti and Ta addition.

  6. Characteristics and Liquid Metal Embrittlement of the steel T91 in contact with Lead–Bismuth Eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Hojna, Anna, E-mail: anna.hojna@cvrez.cz; Di Gabriele, Fosca; Klecka, Jakub

    2016-04-15

    This paper summarizes results of the work carried out on the evaluation of the susceptibility to LME (Liquid Metal Embrittlement) of the ferritic/martensitic steel T91 in contact with LBE (Lead–Bismuth Eutectic). The influence of LBE on the fracture toughness of the steel was studied using 0.5T CT specimen at 355 °C, pre-cracked by cyclic loading in the liquid metal. Tests were carried out in well-defined conditions and according to ASTM standard. It was observed that the LBE decreased the apparent fracture toughness, J{sub IC}, by more than 30%, compared to the value in air. The results are discussed based on examinations of the fracture surface evidencing LME occurrence. The stretch zone accompanying the pre-crack tip blunting was not observed in the specimens exhibiting LME. Therefore, a new fracture toughness, J{sub map}, determined as J integral at the maximum applied load, is proposed to be the appropriate value for fracture resistance evaluation in LBE. The J{sub map} can be applied for the assessment of a pre-existing LME crack stability.

  7. Phase transformation and impact properties in the experimentally simulated weld heat-affected zone of a reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joonoh, E-mail: mjo99@kims.re.kr [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondaero, Seongsangu, Changwon, Gyeongnam 642-831 (Korea, Republic of); Lee, Chang-Hoon; Lee, Tae-Ho [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondaero, Seongsangu, Changwon, Gyeongnam 642-831 (Korea, Republic of); Jang, Min-Ho [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondaero, Seongsangu, Changwon, Gyeongnam 642-831 (Korea, Republic of); Division of Materials Science and Engineering, Hanyang University, Seongdong-ku, Seoul 133-791 (Korea, Republic of); Park, Min-Gu [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondaero, Seongsangu, Changwon, Gyeongnam 642-831 (Korea, Republic of); Department of Material Science and Engineering, Pusan National University, 30 Jangjeon-Dong, Geumjeong-gu, Pusan 609-735 (Korea, Republic of); Han, Heung Nam [Department of Materials Science and Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2014-12-15

    In this work, the phase transformation and impact properties in the weld heat-affected zone (HAZ) of a reduced activation ferritic/martensitic (RAFM) steel are investigated. The HAZs were experimentally simulated using a Gleeble simulator. The base steel consisted of tempered martensite through normalizing at 1000 °C and tempering at 750 °C, while the HAZs consisted of martensite, δ-ferrite and a small volume of autotempered martensite. The impact properties using a Charpy V-notch impact test revealed that the HAZs showed poor impact properties due to the formation of martensite and δ-ferrite as compared with the base steel. In addition, the impact properties of the HAZs further deteriorated with an increase in the δ-ferrite fraction caused by increasing the peak temperature. The impact properties of the HAZs could be improved through the formation of tempered martensite after post weld heat treatment (PWHT), but they remained lower than that of the base steel because the δ-ferrite remained in the tempered HAZs.

  8. Reduced Antivation Ferritic/Martensitic Steel Eurofer 97 as Possible Structural Material for Fusion Devices. Metallurgical Characterization on As-Received Condition and after Simulated Services Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, P.; Lancha, A. M.; Lapena, J.; Serrano, M.; Hernandez-Mayoral, M.

    2004-07-01

    Metallurgical Characterization of the reduced activation ferritic/martensitic steel Eurofer'97, on as-received condition and after thermal ageing treatment in the temperature range from 400 degree centigree to 600 degree centigree for periods up to 10.000 h, was carried out. The microstructure of the steel remained stable (tempered martensite with M{sub 2}3 C{sub 6} and MX precipitates) after the thermal ageing treatments studied in this work. In general, this stability was also observed in the mechanical properties. The Eurofer'97 steel exhibited similar values of hardness, ultimate tensile stress, 0,2% proof stress, USE and T{sub 0}3 regardless of the investigated material condition. However, ageing at 600 degree centigree for 10.000 ha caused a slight increase in the DBTT, of approximately 23. In terms of creep properties, the steel shows in general adequate creep rupture strength levels for short rupture times. However, the results obtained up to now for long time creep rupture tests at 500 degree centigree suggests a change in the deformation mechanisms. (Author) 62 refs.

  9. Reduced Activation Ferritic/Martensitic Steel Eurofer 97 as Possible Structural Material for Fusion Devices. Metallurgical Characterization on As-Received Condition and after Simulated Services Conditions

    International Nuclear Information System (INIS)

    Fernandez, P.; Lancha, A. M.; Lapena, J.; Serrano, M.; Hernandez-Mayoral, M.

    2004-01-01

    Metallurgical Characterization of the reduced activation ferritic/martensitic steel Eurofer'97, on as-received condition and after thermal ageing treatment in the temperature range from 400 degree centigree to 600 degree centigree for periods up to 10.000 h, was carried out. The microstructure of the steel remained stable (tempered martensite with M 2 3 C 6 and MX precipitates) after the thermal ageing treatments studied in this work. In general, this stability was also observed in the mechanical properties. The Eurofer'97 steel exhibited similar values of hardness, ultimate tensile stress, 0,2% proof stress, USE and T 0 3 regardless of the investigated material condition. However, ageing at 600 degree centigree for 10.000 ha caused a slight increase in the DBTT, of approximately 23 . In terms of creep properties, the steel shows in general adequate creep rupture strength levels for short rupture times. However, the results obtained up to now for long time creep rupture tests at 500 degree centigree suggests a change in the deformation mechanisms. (Author) 62 refs

  10. Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

    International Nuclear Information System (INIS)

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

    2007-01-01

    Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

  11. Comparative study of eddy current and Barkhausen noise nondestructive testing methods in microstructural examination of ferrite-martensite dual-phase steel

    Science.gov (United States)

    Ghanei, S.; Kashefi, M.; Mazinani, M.

    2014-04-01

    The magnetic properties of ferrite-martensite dual-phase steels were evaluated using eddy current and Barkhausen noise nondestructive testing methods and correlated with their microstructural changes. Several routes were used to produce different microstructures of dual-phase steels. The first route was different heat treatments in γ region to vary the ferrite grain size (from 9.47 to 11.12 in ASTM number), and the second one was variation in intercritical annealing temperatures (from 750 to 890 °C) in order to produce different percentages of martensite in dual-phase microstructure. The results concerning magnetic Barkhausen noise are discussed in terms of height, position and shape of Barkhausen noise profiles, taking into account two main aspects: ferrite grain size, and different percentages of martensite. Then, eddy current testing was used to study the mentioned microstructural changes by detection of impedance variations. The obtained results show that microstructural changes have a noticeable effect on the magnetic properties of dual-phase steels. The results reveal that both magnetic methods have a high potential to be used as a reliable nondestructive tool to detect and monitor microstructural changes occurring during manufacturing of dual-phase steels.

  12. Processing, Microstructure, and Material Property Relationships Following Friction Stir Welding of Oxide Dispersion Strengthened Steels

    Science.gov (United States)

    2013-09-01

    Fast, 200 Ferritic- martensitic steels , ODS alloys Stainless steels Lead fast reactor Lead or lead- bismuth 800 Fast, 150 Ferritic- martensitic ...from Zinkle [from 1]. T22, T9, T91, E911, NF12, NF616, and SAVE12 are all Ferritic or Martensitic steels with variations in alloy concentrations and...manufacturing techniques. Similarly HCM12 and HCM12A are High Chromium Martensitic steels

  13. Behaviour of the steel T91 under uniaxial and multiaxial slow loading in contact with liquid lead

    Energy Technology Data Exchange (ETDEWEB)

    Hojná, Anna, E-mail: anna.hojna@cvrez.cz [Centrum vyzkumu Rez – CVR, Hlavni 130, 250 68, Husinec, Rez (Czech Republic); Di Gabriele, Fosca; Klecka, Jakub [Centrum vyzkumu Rez – CVR, Hlavni 130, 250 68, Husinec, Rez (Czech Republic); Burda, Jaroslav [UJV Rez, a. s., Hlavni 130, 250 68, Husinec, Rez (Czech Republic)

    2015-11-15

    This work deals with the interaction between liquid lead and the ferritic-martensitic steel, T91. Mechanical properties of specimens loaded in contact with liquid lead were tested in laboratory and boundary conditions necessary to ascertain the sensitivity to Liquid Metal Embrittlement (LME) were studied. Three effects, temperature, deformation rate and surface treatment were selected to stimulate the LME initiation on smooth tensile specimens, then the notch effect was selected as an additional factor. Some specimens were pre-treated by application of a flux followed by dipping into liquid lead to simulate wetting. Slow strain rate tests (SSRT) of specimens immersed in liquid lead were performed applying strain rates from 10{sup −2} to 10{sup −8} 1/s from 350° to 450 °C in test cell CALLISTO. Two types of tensile specimens were tested, smooth and notched. After tests, the fracture mode and the status of specimen surface was examined. Results of the specimens immersed in lead were compared with the results obtained in air. The most significant was the strain rate effect on the stress-strain curves of smooth specimens. Moreover, while no LME was observed for the smooth specimens, clear evidence of LME was observed for the notched specimens.

  14. Effect of Microstructures and Tempering Heat Treatment on the Mechanical Properties of 9Cr-2W Reduced-Activation Ferritic-Martensitic Steel

    International Nuclear Information System (INIS)

    Park, Min-Gu; Kang, Nam Hyun; Moon, Joonoh; Lee, Tae-Ho; Lee, Chang-Hoon; Kim, Hyoung Chan

    2015-01-01

    The aim of this study was to investigate the effect of microstructures (martensite, ferrite, or mixed ferrite and martensite) on the mechanical properties. Of particular interest was the Charpy impact results for 9Cr-2W reduced-activation ferritic-martensitic (RAFM) steels. Under normalized conditions, steel with martensitic microstructure showed superior tensile strength and Charpy impact results. This may result from auto-tempering during the transformation of martensite. On the other hand, both ferrite, and ferrite mixed with martensite, showed unusually poor Charpy impact results. This is because the ferrite phases, and coarse M_23C_6 carbides at the ferrite-grain boundaries acted as cleavage crack propagation paths, and as preferential initiation sites for cleavage cracks, respectively. After the tempering heat treatment, although tensile strength decreased, the energy absorbed during the Charpy impact test drastically increased for martensite, and ferrite mixed with martensite. This was due to the tempered martensite. On the other hand, there were no distinctive differences in tensile and Charpy impact properties of steel with ferrite microstructure, when comparing normalized and tempered conditions.

  15. A study on influence of heat input variation on microstructure of reduced activation ferritic martensitic steel weld metal produced by GTAW process

    International Nuclear Information System (INIS)

    Arivazhagan, B.; Srinivasan, G.; Albert, S.K.; Bhaduri, A.K.

    2011-01-01

    Reduced activation ferritic martensitic (RAFM) steel is a major structural material for test blanket module (TBM) to be incorporated in International Thermonuclear Experimental Reactor (ITER) programme to study the breeding of tritium in fusion reactors. This material has been mainly developed to achieve significant reduction in the induced radioactivity from the structural material used. Fabrication of TBM involves extensive welding, and gas tungsten arc welding (GTAW) process is one of the welding processes being considered for this purpose. In the present work, the effect of heat input on microstructure of indigenously developed RAFM steel weld metal produced by GTAW process has been studied. Autogenous bead-on-plate welding, autogenous butt-welding, butt-welding with filler wire addition, and pulsed welding on RAFMS have been carried out using GTAW process respectively. The weld metal is found to contain δ-ferrite and its volume fraction increased with increase in heat input. This fact suggests that δ-ferrite content in the weld metal is influenced by the cooling rate during welding. It was also observed that the hardness of the weld metal decreased with increase in δ-ferrite content. This paper highlights the effect of heat input and PWHT duration on microstructure and hardness of welds.

  16. Evaluation of mechanically alloyed Cu-based powders as filler alloy for brazing tungsten to a reduced activation ferritic-martensitic steel

    Science.gov (United States)

    de Prado, J.; Sánchez, M.; Ureña, A.

    2017-07-01

    80Cu-20Ti powders were evaluated for their use as filler alloy for high temperature brazing of tungsten to a reduced activation ferritic/martensitic steel (Eurofer), and its application for the first wall of the DEMO fusion reactor. The use of alloyed powders has not been widely considered for brazing purposes and could improve the operational brazeability of the studied system due to its narrower melting range, determined by DTA analysis, which enhances the spreading capabilities of the filler. Ti contained in the filler composition acts as an activator element, reacting and forming several interfacial layers at the Eurofer-braze, which enhances the wettability properties and chemical interaction at the brazing interface. Brazing thermal cycle also activated the diffusion phenomena, which mainly affected to the Eurofer alloying elements causing in it a softening band of approximately 400 μm of thickness. However, this softening effect did not degrade the shear strength of the brazed joints (94 ± 23 MPa), because failure during testing was always located at the tungsten-braze interface.

  17. Microstructural evolution in modified 9Cr-1Mo ferritic/martensitic steel irradiated with mixed high-energy proton and neutron spectra at low temperatures

    International Nuclear Information System (INIS)

    Sencer, B.H.; Garner, F.A.; Gelles, D.S.; Bond, G.M.; Maloy, S.A.

    2002-01-01

    Modified 9Cr-1Mo ferritic/martensitic steel was exposed at 32-57 deg. C to a mixed proton/neutron particle flux and spectrum at the Los Alamos Neutron Science Center. The microstructure of unirradiated 9Cr-1Mo consists of laths, dislocations and carbides. Examination of electron diffraction patterns obtained from extraction replicas of unirradiated 9Cr-1Mo revealed that the precipitate microstructure was primarily dominated by M 23 C 6 carbides. The post-irradiation microstructure contained black-spot damage in addition to precipitates and dislocations. Examination of electron diffraction patterns revealed diffuse rings from M 23 C 6 carbides, indicating amorphization and/or nanocrystallinity. Crystalline MC carbides were also found. No cavity formation was found although a significant amount of helium and hydrogen generation had been generated. TEM-EDS examination of extraction replicas for carbides from unirradiated and irradiated samples did not show any detectable changes in composition of either M 23 C 6 or MC carbides. There was also no evident change in carbide size. Lattice images of M 23 C 6 carbides revealed an amorphous microstructure following irradiation, but MC carbides were still crystalline

  18. Evaluation of mechanically alloyed Cu-based powders as filler alloy for brazing tungsten to a reduced activation ferritic-martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Prado, J. de, E-mail: javier.deprado@urjc.es; Sánchez, M.; Ureña, A.

    2017-07-15

    80Cu-20Ti powders were evaluated for their use as filler alloy for high temperature brazing of tungsten to a reduced activation ferritic/martensitic steel (Eurofer), and its application for the first wall of the DEMO fusion reactor. The use of alloyed powders has not been widely considered for brazing purposes and could improve the operational brazeability of the studied system due to its narrower melting range, determined by DTA analysis, which enhances the spreading capabilities of the filler. Ti contained in the filler composition acts as an activator element, reacting and forming several interfacial layers at the Eurofer-braze, which enhances the wettability properties and chemical interaction at the brazing interface. Brazing thermal cycle also activated the diffusion phenomena, which mainly affected to the Eurofer alloying elements causing in it a softening band of approximately 400 μm of thickness. However, this softening effect did not degrade the shear strength of the brazed joints (94 ± 23 MPa), because failure during testing was always located at the tungsten-braze interface. - Highlights: •W-Eurofer brazed joints, manufactured using Cu-based mechanically alloyed powders as filler is proposed. •The benefits derivate from the alloyed composition could improve the operational brazeability of the studied system. •Tested pre-alloyed fillers have a more homogeneous melting stage which enhances its spreading and flowing capabilities. •This behaviour could lead to work with higher heating rates and lower brazing temperatures.

  19. Energy-filtered TEM imaging and EELS study of ODS particles and argon-filled cavities in ferritic-martensitic steels.

    Science.gov (United States)

    Klimiankou, M; Lindau, R; Möslang, A

    2005-01-01

    Oxide-dispersion-strengthened (ODS) ferritic-martensitic steels with yttrium oxide (Y(2)O(3)) have been produced by mechanical alloying and hot isostatic pressing for use as advanced material in fusion power reactors. Argon gas, usually widely used as inert gas during mechanical alloying, was surprisingly detected in the nanodispersion-strengthened materials. Energy-filtered transmission electron microscopy (EFTEM) and electron energy loss spectroscopy (EELS) led to the following results: (i) chemical composition of ODS particles, (ii) voids with typical diameters of 1-6 nm are formed in the matrix, (iii) these voids are filled with Ar gas, and (iv) the high-density nanosized ODS particles serve as trapping centers for the Ar bubbles. The Ar L(3,2) energy loss edge at 245 eV as well as the absorption features of the ODS particle elements were identified in the EELS spectrum. The energy resolution in the EEL spectrum of about 1.0 eV allows to identify the electronic structure of the ODS particles.

  20. Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Wenhui [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Luo, Fengfeng; Li, Tiecheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-04-15

    Highlights: • Helium/hydrogen synergistic effect can increase irradiation swelling of RAFM steel. • Hydrogen can be trapped to the outer surface of helium bubbles. • Too large a helium bubble can become movable. • Point defects would become mobile and annihilate at dislocations at high temperature. • The peak swelling temperature for RAFM steel is 450 °C. - Abstract: In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He{sup +} beam and sequential He{sup +} and H{sup +} beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.

  1. Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

    International Nuclear Information System (INIS)

    Hu, Wenhui; Guo, Liping; Chen, Jihong; Luo, Fengfeng; Li, Tiecheng; Ren, Yaoyao; Suo, Jinping; Yang, Feng

    2014-01-01

    Highlights: • Helium/hydrogen synergistic effect can increase irradiation swelling of RAFM steel. • Hydrogen can be trapped to the outer surface of helium bubbles. • Too large a helium bubble can become movable. • Point defects would become mobile and annihilate at dislocations at high temperature. • The peak swelling temperature for RAFM steel is 450 °C. - Abstract: In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He + beam and sequential He + and H + beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C

  2. Investigation on microstructure and properties of narrow-gap laser welding on reduced activation ferritic/martensitic steel CLF-1 with a thickness of 35 mm

    Science.gov (United States)

    Wu, Shikai; Zhang, Jianchao; Yang, Jiaoxi; Lu, Junxia; Liao, Hongbin; Wang, Xiaoyu

    2018-05-01

    Reduced activation ferritic martensitic (RAFM) steel is chosen as a structural material for test blanket modules (TBMs) to be constructed in International Thermonuclear Experimental Reactor (ITER) and China Fusion Engineering Test Reactor (CFETR). Chinese specific RAFM steel named with CLF-1 has been developed for CFETR. In this paper, a narrow-gap groove laser multi-pass welding of CLF-1 steel with thickness of 35 mm is conduced by YLS-15000 fiber laser. Further, the microstructures of different regions in the weld joint were characterized, and tensile impact and micro-hardness tests were carried out for evaluating the mecharical properties. The results show that the butt weld joint of CLF-1 steel with a thickness of 35 mm was well-formed using the optimal narrow-gap laser filler wire welding and no obvious defects was found such as incomplete fusion cracks and pores. The microstructures of backing layer is dominated by lath martensites and the Heat-Affected Zone (HAZ) was mainly filled with two-phase hybrid structures of secondary-tempering sorbites and martensites. The filler layer is similar to the backing layer in microstructures. In tensile tests, the tensile samples from different parts of the joint all fractured at base metal (BM). The micro-hardness of weld metal (WM) was found to be higher than that of BM and the Heat-Affected Zone (HAZ) exhibited no obvious softening. After post weld heat treatment (PWHT), it can be observed that the fusion zone of the autogenous welding bead and the upper filling beads mainly consist of lath martensites which caused the lower impact absorbing energy. The HAZ mainly included two-phase hybrid structures of secondary-tempering sorbites and martensites and exhibited favorable impact toughness.

  3. The effect of microstructural change on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS) irradiated in JOYO/MARICO-1

    International Nuclear Information System (INIS)

    Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro

    2004-03-01

    It is well known that the irradiation embrittlement is one of the most important issues to apply ferritic steels for FBR core materials, although ferritic steels have been considered to be candidate core materials of the commercialized FBR core material because of their superior swelling resistance. In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS), microstructural observations were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1. The results obtained in this study are as follows: (1) There was remarkable disappearance of the lath of martensite in the samples irradiated at 650degC, although there was no significant change in microstructures, especially the lath of martensite between the samples irradiated at 500degC and unirradiated. The disappearance of martensitic lath in the samples irradiated at 650degC was larger than that of the samples thermally aged at 650degC. (2) The ductile-brittle transition temperature (DBTT) of irradiated PNC-FMS is judged to increase with the disappearance of martensitic lath and to decrease with the recovery in dislocations. (3) The decrease in the upper shelf energy (USE) of irradiated PNC-FMS is significantly accompanied by the change of precipitation behavior. (4) The Charpy impact properties and microstructures of PNC-FMS irradiated at 500degC were superior under these irradiation conditions. In future, it is necessary to establish how to evaluate Charpy impact properties in a high fluence region, based on theoretical methods introduced from the data gained in low fluence experiments, in addition to expanding the data area widely. (author)

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

    International Nuclear Information System (INIS)

    Sham, T.-L.; Tan, L.; Yamamoto, Y.

    2013-01-01

    Summary of creep, thermal aging and weldability aspects: • The creep resistance of advanced 9Cr FM steels was greatly enhanced by optimizing their compositions as well as by using TMT. – Up to about 700 times increase in creep life, compared to Gr91, was achieved under the accelerated test conditions at 600°C. • The increased density of ultrafine precipitates facilitated the increase in strength and thermal aging resistance, leading to the improved creep resistance. • Properties of four candidate austenitic alloys, HT-UPS, NF709, and two modified HT-UPS alloy (designated Alloys A and B), have been evaluated and compared with 316H. – Alloys A and B showed successful improvement in weldability. – Only a little difference in thermal stability of the alloys in solution annealed conditions. 10% cold work increased the yield strength of the alloys for more than 200% compared to the HT-UPS without cold work. – HT-UPS exhibited the best creep properties among the alloys with and without cold work, and NF709 followed

  5. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER; Hacia la demostracion de capacidad de produccion industrial del acero ferritico-martensitico espanol ASTURFER a escala de demanda ITER

    Energy Technology Data Exchange (ETDEWEB)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-07-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

  6. Development of filler wires for welding of reduced activation ferritic martensitic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2010-07-01

    Indigenous development of reduced activation ferritic-martensitic (RAFM) steel has become necessary for India as a participant in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFM steel is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFM steel filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFM steel. The purpose of this study is to develop filler wires that can be directly used for both gas tungsten arc welding (GTAW) and for narrow-gap gas tungsten arc welding (NG-GTAW) that reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser-MIG welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using GTAW process at various heat inputs with a preheat temperature of 250 C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some amount of delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimized to qualify the filler wires without the presence of delta-ferrite in the weld metal and with optimized mechanical properties. Results showed that the weld metals are free from delta-ferrite. Tensile properties at ambient temperature and at 500 C are well above the specified values, and are much higher than the base metal values. Ductile Brittle Transition Temperature (DBTT) has been evaluated as -81 C based on the 68 J criteria. The present study highlights the basis and methodology

  7. In-situ neutron diffraction study on work-hardening behavior in a ferrite-martensite dual phase steel

    OpenAIRE

    Morooka, Satoshi; Sato, Naoko; Ojima, Mayumi; Harjo, Stefanus; Adachi, Yoshitaka; Tomota, Yo; Umezawa, Osamu

    2011-01-01

    Strength and work-hardening in steels have been discussed from the viewpoint of heterogeneous deformation. In-situ neutron diffraction techniques made clear that the misfit strains between grains accompanied with the grain-scaled internal stress (intergranular stress). In the dual phase steel, the intergranular stress was superposed on phase stress. Both long-range internal stress and short-range one like forest dislocation hardening may cause the resistance for dislocation motion in the steels.

  8. Precipitation behavior of Laves phase and its effect on toughness of 9Cr-2Mo ferritic-martensitic steel

    International Nuclear Information System (INIS)

    Hosoi, Y.; Wade, N.; Kunimitsu, S.; Urita, T.

    1986-01-01

    This study clarified the relationship between the toughness of a 9Cr-2Mo dual phase steel and precipitates formed during aging, with special attention to the Laves phase (Fe 2 Mo). The ductile-brittle transition temperature (DBTT) is increased and the upper shelf energy decreased when the Laves phase begins to precipitate during aging. Electron microscopy and X-ray diffraction indicate that elimination of Si in the steel reduces the precipitation of the Laves phase and results in maintaining good toughness. It is also noted that the toughness of the steel is controlled by the total amount of precipitates (Laves + carbides) in the aging at 873 K for more than 3.6x10 3 ks. A time-temperature-precipitation diagram for the Laves phase is established and it clearly shows that the precipitation of the Laves phase is markedly retarded by the decrease of Si content. In Si-free steel, no Laves phase is observed in the temperature and time range investigated. (orig.)

  9. Metallurgical characterization of the reduced activation ferritic/martensitic steel Eurofer'97 on as-received condition

    International Nuclear Information System (INIS)

    Fernandez, P.; Lancha, A.M.; Lapena, J.; Hernandez-Mayoral, M.

    2001-01-01

    A new European reduced activation ferrous alloy (denominated Eurofer'97) developed as possible first wall and breeder blanket structural material for fusion applications is being characterized. In this paper, activities specially focussed to investigate the microstructural and mechanical properties of this material on the as-received state (normalized at 980 degree sign C/27' plus tempered at 760 degree sign C/90'/air cooled) are presented. Chemical analyses, a detailed microstructural study, hardness, tensile and Charpy tests have been carried out and are compared to the reduced activation material F-82H modified previously studied. The results show that the Eurofer'97 is a fully martensitic steel free of δ-ferrite with similar tensile and better impact properties than the F-82H modified steel. Two types of carbides have been observed in the Eurofer'97, namely, Cr rich precipitates and Ta/V rich precipitates, tentatively identified as M 23 C 6 type and (Ta,V)C type, respectively

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

    International Nuclear Information System (INIS)

    Gelles, D.S.; Shibayama, T.

    1998-01-01

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

  11. Microstructural Variations Across a Dissimilar 316L Austenitic: 9Cr Reduced Activation Ferritic Martensitic Steel Weld Joint

    Science.gov (United States)

    Thomas Paul, V.; Karthikeyan, T.; Dasgupta, Arup; Sudha, C.; Hajra, R. N.; Albert, S. K.; Saroja, S.; Jayakumar, T.

    2016-03-01

    This paper discuss the microstructural variations across a dissimilar weld joint between SS316 and 9Cr-RAFM steel and its modifications on post weld heat treatments (PWHT). Detailed characterization showed a mixed microstructure of austenite and martensite in the weld which is in agreement with the phases predicted using Schaeffler diagram based on composition measurements. The presence of very low volume fraction of δ-ferrite in SS316L has been identified employing state of the art electron back-scattered diffraction technique. PWHT of the ferritic steel did not reduce the hardness in the weld metal. Thermal exposure at 973 K (700 °C) showed a progressive reduction in hardness of weld joint with duration of treatment except in austenitic base metal. However, diffusion annealing at 1073 K (800 °C) for 100 hours resulted in an unexpected increase in hardness of weld metal, which is a manifestation of the dilution effects and enrichment of Ni on the transformation characteristics of the weld zone. Migration of carbon from ferritic steel aided the precipitation of fine carbides in the austenitic base metal on annealing at 973 K (700 °C); but enhanced diffusion at 1073 K (880 °C) resulted in coarsening of carbides and thereby reduction of hardness.

  12. Influence of Tool Rotational Speed and Post-Weld Heat Treatments on Friction Stir Welded Reduced Activation Ferritic-Martensitic Steel

    Science.gov (United States)

    Manugula, Vijaya L.; Rajulapati, Koteswararao V.; Reddy, G. Madhusudhan; Mythili, R.; Bhanu Sankara Rao, K.

    2017-08-01

    The effects of tool rotational speed (200 and 700 rpm) on evolving microstructure during friction stir welding (FSW) of a reduced activation ferritic-martensitic steel (RAFMS) in the stir zone (SZ), thermo-mechanically affected zone (TMAZ), and heat-affected zone (HAZ) have been explored in detail. The influence of post-weld direct tempering (PWDT: 1033 K (760 °C)/ 90 minutes + air cooling) and post-weld normalizing and tempering (PWNT: 1253 K (980 °C)/30 minutes + air cooling + tempering 1033 K (760 °C)/90 minutes + air cooling) treatments on microstructure and mechanical properties has also been assessed. The base metal (BM) microstructure was tempered martensite comprising Cr-rich M23C6 on prior austenite grain and lath boundaries with intra-lath precipitation of V- and Ta-rich MC precipitates. The tool rotational speed exerted profound influence on evolving microstructure in SZ, TMAZ, and HAZ in the as-welded and post-weld heat-treated states. Very high proportion of prior austenitic grains and martensite lath boundaries in SZ and TMAZ in the as-welded state showed lack of strengthening precipitates, though very high hardness was recorded in SZ irrespective of the tool speed. Very fine-needle-like Fe3C precipitates were found at both the rotational speeds in SZ. The Fe3C was dissolved and fresh precipitation of strengthening precipitates occurred on both prior austenite grain and sub-grain boundaries in SZ during PWNT and PWDT. The post-weld direct tempering caused coarsening and coalescence of strengthening precipitates, in both matrix and grain boundary regions of TMAZ and HAZ, which led to inhomogeneous distribution of hardness across the weld joint. The PWNT heat treatment has shown fresh precipitation of M23C6 on lath and grain boundaries and very fine V-rich MC precipitates in the intragranular regions, which is very much similar to that prevailed in BM prior to FSW. Both the PWDT and PWNT treatments caused considerable reduction in the hardness of SZ

  13. Investigation of a Novel NDE Method for Monitoring Thermomechanical Damage and Microstructure Evolution in Ferritic-Martensitic Steels for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nagy, Peter

    2013-09-30

    The main goal of the proposed project is the development of validated nondestructive evaluation (NDE) techniques for in situ monitoring of ferritic-martensitic steels like Grade 91 9Cr-1Mo, which are candidate materials for Generation IV nuclear energy structural components operating at temperatures up to ~650{degree}C and for steam-generator tubing for sodium-cooled fast reactors. Full assessment of thermomechanical damage requires a clear separation between thermally activated microstructural evolution and creep damage caused by simultaneous mechanical stress. Creep damage can be classified as "negligible" creep without significant plastic strain and "ordinary" creep of the primary, secondary, and tertiary kind that is accompanied by significant plastic deformation and/or cavity nucleation and growth. Under negligible creep conditions of interest in this project, minimal or no plastic strain occurs, and the accumulation of creep damage does not significantly reduce the fatigue life of a structural component so that low-temperature design rules, such as the ASME Section III, Subsection NB, can be applied with confidence. The proposed research project will utilize a multifaceted approach in which the feasibility of electrical conductivity and thermo-electric monitoring methods is researched and coupled with detailed post-thermal/creep exposure characterization of microstructural changes and damage processes using state-of-the-art electron microscopy techniques, with the aim of establishing the most effective nondestructive materials evaluation technique for particular degradation modes in high-temperature alloys that are candidates for use in the Next Generation Nuclear Plant (NGNP) as well as providing the necessary mechanism-based underpinnings for relating the two. Only techniques suitable for practical application in situ will be considered. As the project evolves and results accumulate, we will also study the use of this technique for monitoring other GEN IV

  14. Fabrication and characterization of reference 9Cr and 12Cr-ODS low activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Muroga, T., E-mail: muroga@nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Nagasaka, T. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Li, Y.; Abe, H. [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan); Ukai, S. [Graduate School of Engineering, Hokkaido University, N13, W8, Kita-ku, Sapporo 060-8628 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Okuda, T. [Kobelco Research Institute, 1-5-5 Takatsukadai, Nishi-ku, Kobe, Hyogo 651-2271 (Japan)

    2014-10-15

    For the purpose of arranging reference alloys available for various characterization efforts by Japanese fusion research groups, fabrication of reference 9Cr and 12Cr-ODS steels have been carried out with similar manufacturing processes followed by various characterizations. The fabrication proceeded with powder mixing, MA, encapsulation into mild steel cases, hot extrusion and hot forging, followed by final heat treatments. Each alloy was extruded into three bars. The characterization included chemical composition analysis, SEM and TEM microstructural observations, hardness tests, tensile tests at RT and 973 K, and relatively short-term thermal creep tests at 973 K. Room temperature hardness for 9Cr-ODS was larger than 12Cr-ODS, the former showing large increase when annealing temperature exceeded 1200 K and the latter showing no significant change with annealing temperature. Tensile strength of 9Cr-ODS was significantly larger than that of 12Cr-ODS at RT but comparable at 973 K. 9Cr-ODS showed longer and shorter creep rupture time than 12Cr-ODS at high and low stress levels, respectively. The mechanism of the difference in creep properties of the two alloys was discussed.

  15. Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique

    International Nuclear Information System (INIS)

    Otsuka, Teppei; Tanabe, Tetsuo

    2013-01-01

    Highlights: ► We applied a tritium imaging plate technique to depth profiling of hydrogen in bulk. ► Changes of hydrogen depth profiles in the steel by thermal annealing were examined. ► We proposed a release model of plasma-loaded hydrogen in the steel. ► Hydrogen is trapped at trapping sites newly developed by plasma loading. ► Hydrogen is also trapped at surface oxides and hardly desorbed by thermal annealing. -- Abstract: In order to understand how hydrogen loaded by plasma in F82H is removed by annealing at elevated temperatures in vacuum, depth profiles of plasma-loaded hydrogen were examined by means of a tritium imaging plate technique. Owing to large hydrogen diffusion coefficients in F82H, the plasma-loaded hydrogen easily penetrates into a deeper region becoming solute hydrogen and desorbs by thermal annealing in vacuum. However the plasma-loading creates new hydrogen trapping sites having larger trapping energy than that for the intrinsic sites beyond the projected range of the loaded hydrogen. Some surface oxides also trap an appreciable amount of hydrogen which is more difficult to remove by the thermal annealing

  16. Viscoplastic equations incorporated into a finite element model to predict deformation behavior of irradiated reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yuanyuan, E-mail: 630wyy@163.com [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), Dalian University of Technology, Dalian 116024 (China); Zhao, Jijun, E-mail: zhaojj@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), Dalian University of Technology, Dalian 116024 (China); Zhang, Chi [Key Laboratory of Advanced Materials of Ministry of Education, School of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • The initial internal variable in the Anand model is modified by considering both temperature and irradiation dose. • The tensile stress-strain response is examined and analyzed under different temperatures and irradiation doses. • Yield strengths are predicted as functions of strain rate, temperature and irradiation dose. - Abstract: The viscoplastic equations with a modified initial internal variable are implemented into the finite element code to investigate stress-strain response and irradiation hardening of the materials under increased temperature and at different levels of irradiated dose. We applied this model to Mod 9Cr-1Mo steel. The predicted results are validated by the experimentally measured data. Furthermore, they show good agreement with the previous data from a constitutive crystal plasticity model in account of dislocation and interstitial loops. Three previous hardening models for predicting the yield strength of the material are discussed and compared with our simulation results.

  17. Effect of Proton Irradiation on the Corrosion Behaviors of Ferritic/Martensitic Steel in Liquid Metal Environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeonghyeon; Kim, Tae Yong; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Liquid metal fast breeder reactors (LMFBRs) such as sodium-cooled fast reactor (SFR) and lead-cooled fast reactor (LFR) are the candidates of GEN-IV nuclear energy systems. Among various liquid metals that can be used as primary coolant material, sodium is a world widely used coolant for GEN-IV reactors. In this study, as-received Gr.92 and irradiated Gr.92 specimen in the oxygen-saturated liquid sodium were examined at high temperature for 300h. The microstructure results reveal the information of the effect of irradiation and effect of the chrome concentration in specimen. From the SRIM result, penetration distance of 40 μm in stainless steel and nominal sample thickness of 30 μm was used to avoid the damage peak and any proton implantation and From the microstructural evaluation, chromium-rich zones existed under the surface of the both of non-irradiated and irradiated materials. The irradiated materials showed chromium-rich zones with larger depths than the non-irradiated specimens.

  18. Thermal-Fatigue Analysis of W-coated Ferritic-Martensitic Steel Mockup for Fusion Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Kim, Suk Kwon; Park, Seong Dae; Kim, Dong Jun [KAERI, Daejeon (Korea, Republic of); Moon, Se Yeon; Hong, Bong Guen [Chonbuk University, Jeonju (Korea, Republic of)

    2016-05-15

    In this study, commercial ANSYS-CFX for thermalhydraulic analysis and ANSYS-mechanical for the thermo-mechanical analysis are used to evaluate the thermal-lifetime of the mockup to determine the test conditions. Also, the Korea Heat Load Test facility with an Electron Beam (KoHLT-EB) will be used and its water cooling system is considered to perform the thermal-hydraulic analysis especially for considering the two-phase analysis with a higher heat flux conditions. Through the ITER blanket first wall (BFW) development project in Korea, the joining methods were developed with a beryllium (Be) layer as a plasma-facing material, a copper alloy (CuCrZr) layer as a heat sink, and type 316L austenitic stainless steel (SS316L) as a structural material. And joining methods were developed such as Be as an armor and FMS as a structural material, or W as an armor and FMS as a structural material were developed through the test blanket module (TBM) program. As a candidate of PFC for DEMO, a new W/FMS joining methods, W coating with plasma torch, have been developed. The HHF test conditions are found by performing a thermal-hydraulic and thermo-mechanical analysis with the conventional codes such as ANSYSCFX and .mechanical especially for considering the two-phase condition in cooling tube.

  19. Thermal-Fatigue Analysis of W-joined Ferritic-Martensitic Steel Mockup for Fusion Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon; Park, Seong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Kyu In [Gentec Co., Daejeon (Korea, Republic of); Moon, Se Yeon; Hong, Bong Guen [Chonbuk National University, Chonbuk (Korea, Republic of)

    2015-10-15

    Through the ITER blanket first wall (BFW) development project in Korea, the joining methods were developed with a beryllium (Be) layer as a plasma-facing material, a copper alloy (CuCrZr) layer as a heat sink, and type 316L austenitic stainless steel (SS316L) as a structural material. And joining methods were developed such as Be as an armor and FMS as a structural material, or W as an armor and FMS as a structural material were developed through the test blanket module (TBM) program. As a candidate of PFC for DEMO, W/FMS joining methods have been developed and a new Ti interlayer was applied differently from the previous work. In the present study, the W/FMS PFC development was introduced with the following procedure to apply to the PFCs for a fusion reactor: (1) Three W/FMS mockups were fabricated using the developed HIP followed by a post-HIP heat treatment (PHHT). (2) Because the High Heat Flux (HHF) test should be performed over the thermal lifetime of the mockup under the proper test conditions to confirm the joint's integrity, the test conditions were determined through a preliminary analysis. In this study, commercial ANSYS-CFX for thermalhydraulic analysis and ANSYS-mechanical for the thermo-mechanical analysis are used to evaluate the thermal-lifetime of the mockup to determine the test conditions. Also, the Korea Heat Load Test facility with an Electron Beam (KoHLT-EB) will be used and its water cooling system is considered to perform the thermal-hydraulic analysis especially for considering the two-phase analysis with a higher heat flux conditions. From the analysis, the heating and the cooling conditions were determined for 0.5- and 1.0-MW/m{sup 2} heat fluxes, respectively. Elastic-plastic analysis is performed to determine the lifetime and finally, the 1.0 MW/m{sup 2} heat flux conditions are determined up to 4,306 cycles. The test will be done in the near future and the measured temperatures will be compared with the present simulation results.

  20. Orientation relationship in various 9% Cr ferritic/martensitic steels-EBSD comparison between Nishiyama-Wassermann, Kurdjumov-Sachs and Greninger-Troiano

    International Nuclear Information System (INIS)

    Barcelo, F.; Bechade, J. L.; Fournier, B.

    2010-01-01

    EBSD measurements were carried out on four different martensitic steels (T91, P92, EM10 and Eurofer) in various metallurgical conditions (nine different microstructural states). The usual orientation relationships (ORs) between the parent austenitic phase and the resulting martensite in martensitic steels are those of Nishiyama-Wassermann (NW) and Kurjumov-Sachs (KS). The present study first proposes a methodology based on the combined analysis of the misorientation distribution, the pole figures (PFs) and the angle/axis pairs. This methodology leads to the conclusion that neither NW nor KS relationships are able to account for all the features observed whatever the material under study. A third OR proposed by Greninger and Troiano (GT) proves to describe the relationship between austenite and ferrite in all four different martensitic steels much more accurately. (authors)

  1. Mechanical characterization of a reduced activation 9 Cr ferritic/martensitic steel of spanish production; Caracterizacion mecanica de un acero ferritico/martenitico de activacion reducida de produccion espanola

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, D.; Serrano, M.

    2012-07-01

    This paper shows the first results concerning the characterization of two heats of a reduced activation 9 Cr ferritic/martensitic steel (RAFM) made in Spain, called AF1B and AF2A. The results of this characterization are compared with their European counterparts, EUROFER97-2, which was chosen as reference material. All activities described were performed in the Structural Materials Unit of CIEMAT, within the national project TECNO-FUS CONSOLIDER INGENIO.The two Spanish heats have the same production process and heat treatment. Both heats have a similar tensile behaviour similar to EUROFER97-2, but on the other hand impact properties are lower. The microstructure of AF1B reveals large biphasic inclusions that affecting its mechanical properties, especially the impact properties. AF2A casting was free of these inclusions. (Author) 24 refs.

  2. Mechanical behavior and fracture characterization of the T91 martensitic steel in liquid sodium environment

    International Nuclear Information System (INIS)

    Hamdane, Ouadie

    2012-01-01

    The T91 martensitic steel is designed to constitute structural material of future sodium fast reactors of fourth generation, where it will be subjected to stresses in presence of liquid sodium. This study presents a qualitative and quantitative estimate of the sensitivity of T91 steel towards the phenomenon of liquid metal embrittlement. The effect of liquid sodium on T91 steel was studied and quantified according to the temperature and the cross head rate displacement, by using a set-up of Small Punch Test, three and four bending test, developed in laboratory. Mechanical tests in sodium environment are carried out inside a Plexiglas cell, conceived and developed at the laboratory. The atmosphere inside this cell is severely purified and controlled, in order to avoid on the one hand an explosive reaction of sodium with moisture, or an ignition with oxygen, and on the other hand to minimize the presence of impurities in liquid sodium used. The presence of sodium accelerates T91 steel fracture at low temperature, without modifying its ductile character. The T91 pre-immersion in sodium makes it possible to dissolve the protective layer of chromium oxide, and to obtain an intimate contact with the molten metal. However, pre-immersion generates a surface defects which cause a partial embrittlement by sodium. The hardening of T91 steel by heat treatment with a tempering temperature of 550 C (T91-TR550) causes a total embrittlement of steel in presence of sodium, with and without pre-immersion. The rupture of the T91-TR550 steel takes then place by intergranular de-cohesion, corresponding to the crack initiation phase, followed by laths de-cohesion, corresponding to the phase of propagation of these cracks. The mechanism suggested in this study is based on the intergranular penetration of sodium, supported by the presence of segregated impurities such phosphorus, and by the plastic deformation [fr

  3. Study of MHD Corrosion and Transport of Corrosion Products of Ferritic/Martensitic Steels in the Flowing PbLi and its Application to Fusion Blanket

    Science.gov (United States)

    Saeidi, Sheida

    Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. The research performed here is aimed at: (1) better understanding of corrosion processes in the system including RAFM steel and flowing PbLi in the presence of a strong magnetic field and (2) prediction of corrosion losses in conditions of a Dual Coolant Lead Lithium (DCLL) blanket, which is at present the key liquid metal blanket concept in the US. To do this, numerical and analytical tools have been developed and then applied to the analysis of corrosion processes. First, efforts were taken to develop a computational suite called TRANSMAG (Transport phenomena in Magnetohydrodynamic Flows) as an analysis tool for corrosion processes in the PbLi/RAFM system, including transport of corrosion products in MHD laminar and turbulent flows. The computational approach in TRANSMAG is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. Then, the new computational tool was used to solve an inverse mass transfer problem where the saturation concentration of iron in PbLi was reconstructed from the experimental data resulting in the following correlation: CS = e 13.604--12975/T, where T is the temperature of PbLi in K and CS is in wppm. The new correlation for saturation concentration was then used in the analysis of corrosion processes in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown in this study, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely

  4. Parameters promoting liquid metal embrittlement of the T91 steel in lead-bismuth eutectic alloy

    International Nuclear Information System (INIS)

    Proriol Serre, I.; Ye, C.; Vogt, J.B.

    2015-01-01

    The use of liquid lead-bismuth eutectic (LBE) as a spallation target and a coolant in accelerator-driven systems raises the question of the reliability of structural materials, such as T91 martensitic steel in terms of liquid metal assisted damage and corrosion. In this study, the mechanical behaviour of the T91 martensitic steel was examined in liquid lead-bismuth eutectic (LBE) and in inert atmosphere. Several conditions showed the most sensitive embrittlement factor. The Small Punch Test technique was employed using smooth specimens. In this standard heat treatment, T91 appeared in general as a ductile material, and became brittle in the considered conditions if the test was performed in LBE. It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement (LME) of the T91 steel in LBE. Loading the T91 very slowly instead of rapidly in oxygen saturated LBE resulted in brittle fracture. Furthermore, low-oxygen content in LBE and an increase in temperature promote LME. (authors)

  5. Characterization of films formed by the aluminizing of T91 steel

    Science.gov (United States)

    Sanabria Cala, J. A.; Conde Rodríguez, G. R.; Y Peña Ballesteros, D.; Laverde Cataño, D.; Quintero Rangel, L. S.

    2017-12-01

    The aluminizing of a T91 martensitic ferritic steel was carried out by a novel modification to the traditional technique of packed cementation, with the objective of producing a diffusion coating of aluminum in a shorter time and operating cost, from a technique that allows the reuse of powder packaging and which the coating of metal parts with complex shapes can be secured. As an aluminum source, commercial foil is used to wrap the piece to be coated, while the powder packaging contains aluminum oxide Al2O3 and an activating salt, ammonium chloride NH4Cl. During the deposition process of the coating, the NH4Cl is decomposed by reacting with foil, and thus, aluminum halides can be transferred to the metallic substrate, which deposit aluminum on the T91 steel surface while Al2O3 can be recycled for subsequent processes. The results of the diffractograms and micrographs indicated the strong influence of temperature, exposure time and ammonium chloride concentration in the formation and growth evolution of a stable coating of iron-aluminum and iron-aluminum-nickel on the T91 steel surface, which was effectively deposited at a temperature of 700°C and an exposure period of 9 hours. The coating formed on the T91 steel surface could play a protective role towards the material by acting as a physical barrier between the alloy and other corrosive species in high temperature operated systems.

  6. Thermal Effects on Thin Laser-Peened Ferritic-Martensitic Samples

    International Nuclear Information System (INIS)

    Caro, M; Zalesky, T; Hosemann, P; El-dasher, B S; Halsey, W G; Stuart, B

    2007-01-01

    Laser peening is emerging as a promising technique to improve the corrosion resistance of cladding material candidates for lead-cooled fast reactors (LFRs). The challenge is in the performance capability of ∼1 mm-thick fuel-pin cladding. Ferritic-martensitic (F/M) steels are foreseen as possible candidates that stand severe conditions of high dose (150 dpa), and high temperature (∼500-600 C) under the corrosive environments of molten lead or lead-bismuth. In this paper, we present the results of experiments carried on laser peened (LP) samples of F/M steels HT9, T91, EP823, as well as the austenitic material 316 L. The samples underwent a thermal treatment in an oven at 520 C, and XRD compressive stress results indicate that the F/M samples do not retain the residual stress after 2 weeks of heat treatment. The corrosion behavior in flowing lead-bismuth eutectic (LBE) at 535 C has been investigated as well. Also, irradiation experiments of LP samples to a maximum dose of ∼10 dpa are foreseen

  7. The evolution of internal stress and dislocation during tensile deformation in a 9Cr ferritic/martensitic (F/M) ODS steel investigated by high-energy X-rays

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guangming; Zhou, Zhangjian; Mo, Kun; Miao, Yinbin; Liu, Xiang; Almer, Jonathan; Stubbins, James F.

    2015-12-01

    An application of high-energy wide angle synchrotron X-ray diffraction to investigate the tensile deformation of 9Cr ferritic/martensitic (F/M) ODS steel is presented. With tensile loading and in-situ Xray exposure, the lattice strain development of matrix was determined. The lattice strain was found to decrease with increasing temperature, and the difference in Young's modulus of six different reflections at different temperatures reveals the temperature dependence of elastic anisotropy. The mean internal stress was calculated and compared with the applied stress, showing that the strengthening factor increased with increasing temperature, indicating that the oxide nanoparticles have a good strengthening impact at high temperature. The dislocation density and character were also measured during tensile deformation. The dislocation density decreased with increasing of temperature due to the greater mobility of dislocation at high temperature. The dislocation character was determined by best-fit methods for different dislocation average contrasts with various levels of uncertainty. The results shows edge type dislocations dominate the plastic strain at room temperature (RT) and 300 C, while the screw type dislocations dominate at 600 C. The dominance of edge character in 9Cr F/M ODS steels at RT and 300 C is likely due to the pinning effect of nanoparticles for higher mobile edge dislocations when compared with screw dislocations, while the stronger screw type of dislocation structure at 600 C may be explained by the activated cross slip of screw segments.

  8. Postirradiation thermocyclic loading of ferritic-martensitic structural materials

    Science.gov (United States)

    Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

    Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

  9. Thermal expansion characteristics of Fe-9Cr-0.12C-0.56Mn-0.24V-1.38W-0.06Ta (wt.%) reduced activation ferritic-martensitic steel

    Science.gov (United States)

    Subramanian, Raju; Tripathy, Haraprasanna; Rai, Arun Kumar; Hajra, Raj Narayan; Saibaba, Saroja; Jayakumar, Tammana; Rajendra Kumar, Ellappan

    2015-04-01

    The lattice and bulk thermal expansion behavior of an Indian version of reduced activation ferritic-martensitic (INRAFM) steel has been quantified using high temperature X-ray diffraction and dilatometry. The lattice parameter of tempered α-ferrite phase exhibited a smooth quadratic increase with temperature, while that of γ-austenite remained fairly linear up to 1273 K. The results suggest that α-ferrite + Carbides → γ-austenite transformation occurs upon continuous heating in the temperature range, 1146 ⩽ T ⩽ 1173 K. Further, this transformation is found to be accompanied by a reduction in average atomic volume. The mean linear thermal expansion coefficients of tempered α-ferrite and γ-austenite phases are estimated to be about 1.48 × 10-5 and 2.4 × 10-5 K-1 respectively. The magnetic contribution to relative thermal dilatation (Δl/l298)mag is found to be small and negative, as compared to phonon contribution.

  10. The evolution of internal stress and dislocation during tensile deformation in a 9Cr ferritic/martensitic (F/M) ODS steel investigated by high-energy X-rays

    International Nuclear Information System (INIS)

    Zhang, Guangming; Zhou, Zhangjian; Mo, Kun; Miao, Yinbin; Liu, Xiang; Almer, Jonathan; Stubbins, James F.

    2015-01-01

    An application of high-energy wide angle synchrotron X-ray diffraction to investigate the tensile deformation of 9Cr ferritic/martensitic (F/M) ODS steel is presented. With tensile loading and in-situ X-ray exposure, the lattice strain development of matrix was determined. The lattice strain was found to decrease with increasing temperature, and the difference in Young's modulus of six different reflections at different temperatures reveals the temperature dependence of elastic anisotropy. The mean internal stress was calculated and compared with the applied stress, showing that the strengthening factor increased with increasing temperature, indicating that the oxide nanoparticles have a good strengthening impact at high temperature. The dislocation density and character were also measured during tensile deformation. The dislocation density decreased with increasing of temperature due to the greater mobility of dislocation at high temperature. The dislocation character was determined by best-fit methods for different dislocation average contrasts with various levels of uncertainty. The results shows edge type dislocations dominate the plastic strain at room temperature (RT) and 300 °C, while the screw type dislocations dominate at 600 °C. The dominance of edge character in 9Cr F/M ODS steels at RT and 300 °C is likely due to the pinning effect of nanoparticles for higher mobile edge dislocations when compared with screw dislocations, while the stronger screw type of dislocation structure at 600 °C may be explained by the activated cross slip of screw segments. - Highlights: • The tensile deformation of 9Cr ODS steel was studied by synchrotron irradiation. • The evolution of internal mean stress was calculated. • The evolution of dislocation character was determined by best-fit method. • Edge type dominates plasticity at RT and 300 °C, while screw type dominates at 600 °C.

  11. The evolution of internal stress and dislocation during tensile deformation in a 9Cr ferritic/martensitic (F/M) ODS steel investigated by high-energy X-rays

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guangming [School of Materials Science and Engineering, University of Science and Technology, Beijing, Beijing 100083 (China); Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL 61801 (United States); Zhou, Zhangjian, E-mail: zhouzhj@mater.ustb.edu.cn [School of Materials Science and Engineering, University of Science and Technology, Beijing, Beijing 100083 (China); Mo, Kun [Nuclear Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Miao, Yinbin; Liu, Xiang [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL 61801 (United States); Almer, Jonathan [X-ray Science Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Stubbins, James F. [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL 61801 (United States)

    2015-12-15

    An application of high-energy wide angle synchrotron X-ray diffraction to investigate the tensile deformation of 9Cr ferritic/martensitic (F/M) ODS steel is presented. With tensile loading and in-situ X-ray exposure, the lattice strain development of matrix was determined. The lattice strain was found to decrease with increasing temperature, and the difference in Young's modulus of six different reflections at different temperatures reveals the temperature dependence of elastic anisotropy. The mean internal stress was calculated and compared with the applied stress, showing that the strengthening factor increased with increasing temperature, indicating that the oxide nanoparticles have a good strengthening impact at high temperature. The dislocation density and character were also measured during tensile deformation. The dislocation density decreased with increasing of temperature due to the greater mobility of dislocation at high temperature. The dislocation character was determined by best-fit methods for different dislocation average contrasts with various levels of uncertainty. The results shows edge type dislocations dominate the plastic strain at room temperature (RT) and 300 °C, while the screw type dislocations dominate at 600 °C. The dominance of edge character in 9Cr F/M ODS steels at RT and 300 °C is likely due to the pinning effect of nanoparticles for higher mobile edge dislocations when compared with screw dislocations, while the stronger screw type of dislocation structure at 600 °C may be explained by the activated cross slip of screw segments. - Highlights: • The tensile deformation of 9Cr ODS steel was studied by synchrotron irradiation. • The evolution of internal mean stress was calculated. • The evolution of dislocation character was determined by best-fit method. • Edge type dominates plasticity at RT and 300 °C, while screw type dominates at 600 °C.

  12. Study mechanism of growth and spallation of oxide scales formed after T91 steel oxidation in water vapor at 550 C

    International Nuclear Information System (INIS)

    Demizieux, Marie-Christine

    2015-01-01

    In the framework of the development of Generation IV reactors and specifically in the new Sodium Fast Reactor (SFR) project, Fe-9Cr ferritic-martensitic steels are candidates as structural materials for steam generators. Indeed, Fe-9Cr steels are already widely used in high temperature steam environments - like boilers and steam turbines- for their combination of creep strength and high thermal properties. Many studies have been focused on Fe-9Cr steels oxidation behavior between 550 C-700 C.Depending on the oxidizing environment, formation of a triplex (Fe-Cr spinel/magnetite/hematite) or duplex (Fe-Cr spinel/magnetite) oxide scales are reported.. Besides, for long time exposure in steam, the exfoliation of oxide scales can cause serious problems such as tube obstruction and steam turbine erosion. Consequently, this work has been dedicated to study, on the one hand the oxidation kinetics of T91 steel in water vapor environments, and on the other hand, the mechanisms leading to the spallation of the oxide scale. Oxidation tests have been carried out at 550 C in pure water vapor and in Ar/D_2O/H_2 environments with different hydrogen contents. Based on an analytical resolution, a quantitative modeling has shown that the 'available space model' proposed in the literature for duplex oxide scale formation well reproduces both scales growth kinetics and spinel oxide stoichiometry. Then, oxidized samples have been precisely characterized and it turns out that buckling then spalling of the oxide scale is always located in the magnetite layer. Voids observed in the magnetite layer are major initiation sites of de-cohesion of the outer oxide scale. A mechanism of formation of these voids has been proposed, in accordance with the mechanism of duplex scale formation. The derived model based on the assumption that vacancies accumulate where the iron vacancies flux divergence is maximal gives a good estimation of the location of pores inside the magnetite layer. Then, in order

  13. Brittle fracture of T91 steel in liquid lead–bismuth eutectic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Changqing, E-mail: Changqing.ye@ed.univ-lille1.fr; Vogt, Jean-Bernard, E-mail: jean-bernard.vogt@univ-lille1.fr; Proriol-Serre, Ingrid, E-mail: ingrid.proriol-serre@univ-lille1.fr

    2014-12-15

    Highlights: • Tempering temperature is important for LBE embrittlement occurrence. • Brittle behaviour in LBE evidenced by small punch test and fatigue test. • Brittle behaviour in low oxygen LBE observed for low loading rate. - Abstract: The mechanical behaviour of the T91 martensitic steel has been studied in liquid lead–bismuth eutectic (LBE) and in inert atmosphere. Several conditions were considered to point out the most sensitive embrittling factors. Smooth and notched specimens were employed for respectively monotonic and cyclic loadings. The present investigation showed that T91 appeared in general as a ductile material, and became brittle in the considered conditions only if at least tests were performed in LBE. It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement of T91 in LBE. For the standard heat treatment condition, loading monotonically the T91 very slowly instead of rapidly in LBE resulted in brittle fracture. Also, under cyclic loading, the crack propagated in a brittle manner in LBE.

  14. Tensile and Charpy impact properties of an ODS ferritic/martensitic steel 9Cr–1.8W–0.5Ti–0.35Y{sub 2}O{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guangming; Zhou, Zhangjian, E-mail: zhouzhj@mater.ustb.edu.cn; Wang, Man; Li, Shaofu; Zou, Lei; Zhang, Liwei

    2014-04-15

    Highlights: • The tensile property and Charpy impact were tested. • Both strength and plasticity in LT direction are better than that of TL direction. • The LSE was more than 65% of the USE from absorbed energy curve. • The initiation and propagation energy at different temperatures were calculated. • High LSE and dimples on the fracture surface indicated good toughness at −60 °C. - Abstract: A 9Cr-ODS ferritic/martensitic steel with a composition of 9Cr–1.8W–0.5Ti–0.35Y{sub 2}O{sub 3} was fabricated by mechanical alloying and hot isostatic pressing, followed by hot rolling. Tensile properties were measured at room temperature (23 °C) and 700 °C in the rolling direction (LT) and the transverse direction (TL). The ultimate tensile strength (UTS) of the as-rolled samples in both directions reached 990 MPa at 23 °C, and still maintained at 260 MPa at 700 °C. The tensile strength and elongation of the rolling direction was greater than that of the transverse direction. The Charpy impact was tested from −100 to 100 °C in the LT direction. The lower shelf energy (LSE) was more than 65% of the upper shelf energy (USE). The total absorbed energy was separated into the energies for crack initiation and propagation. The propagation energy was always higher than the initiation energy in the range of temperatures tested. The ductile-to-brittle transition temperature (DBTT) of the rolled 9Cr ODS evaluated by an absorbed energy curve was about 0 °C. However, the high LSE and the fracture surface that still contained dimples at lower shelf indicated good toughness of the as-rolled 9Cr ODS steels at temperature of −60 °C.

  15. Creep Tests and Modeling Based on Continuum Damage Mechanics for T91 and T92 Steels

    Science.gov (United States)

    Pan, J. P.; Tu, S. H.; Zhu, X. W.; Tan, L. J.; Hu, B.; Wang, Q.

    2017-12-01

    9-11%Cr ferritic steels play an important role in high-temperature and high-pressure boilers of advanced power plants. In this paper, a continuum damage mechanics (CDM)-based creep model was proposed to study the creep behavior of T91 and T92 steels at high temperatures. Long-time creep tests were performed for both steels under different conditions. The creep rupture data and creep curves obtained from creep tests were captured well by theoretical calculation based on the CDM model over a long creep time. It is shown that the developed model is able to predict creep data for the two ferritic steels accurately up to tens of thousands of hours.

  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

    Science.gov (United States)

    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. 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

    Energy Technology Data Exchange (ETDEWEB)

    Kupriiyanova, Y.E., E-mail: fomenkoj@kipt.kharkov.ua [National Science Centre Kharkov Institute of Physics and Technology, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Bryk, V.V.; Borodin, O.V.; Kalchenko, A.S.; Voyevodin, V.N.; Tolstolutskaya, G.D. [National Science Centre Kharkov Institute of Physics and Technology, 1, Akademicheskaya St., Kharkov, 61108 (Ukraine); Garner, F.A. [Radiation Effects Consulting, Richland, WA 99354 (United States)

    2016-01-15

    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{sup +3}, 40 keV He{sup +}, and 20 keV H{sup +}. 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.

  18. Modelling of liquid sodium induced crack propagation in T91 martensitic steel: Competition with ductile fracture

    Energy Technology Data Exchange (ETDEWEB)

    Hemery, Samuel [Institut PPRIME, CNRS, Université de Poitiers, ISAE ENSMA, UPR 3346, Téléport 2, 1 Avenue Clément Ader, BP 40109, 86961 Futuroscope Chasseneuil Cedex (France); Berdin, Clotilde, E-mail: clotilde.berdin@u-psud.fr [Univ Paris-Sud, SP2M-ICMMO, CNRS UMR 8182, F-91405 Orsay Cedex (France); Auger, Thierry; Bourhi, Mariem [Ecole Centrale-Supelec, MSSMat CNRS UMR 8579, F-92295 Chatenay Malabry Cedex (France)

    2016-12-01

    Liquid metal embrittlement (LME) of T91 steel is numerically modeled by the finite element method to analyse experimental results in an axisymmetric notched geometry. The behavior of the material is identified from tensile tests then a crack with a constant crack velocity is introduced using the node release technique in order to simulate the brittle crack induced by LME. A good agreement between the simulated and the experimental macroscopic behavior is found: this suggests that the assumption of a constant crack velocity is correct. Mechanical fields during the embrittlement process are then extracted from the results of the finite element model. An analysis of the crack initiation and propagation stages: the ductile fracture probably breaks off the LME induced brittle fracture. - Highlights: • T91 martensitic steel is embrittled by liquid sodium depending on the loading rate at 573 K. • The mechanical behavior is modeled by a von Mises elastic-plastic law. • The LME induced crack propagates at a constant velocity. • The mechanical state at the crack tip does not explain a brittle crack arrest. • The occurrence of the ductile fracture breaks off the brittle fracture.

  19. Influence of dynamic sodium environment on the creep-fatigue behaviour of Modified 9Cr-1Mo ferritic-martensitic steel

    International Nuclear Information System (INIS)

    Kannan, R.; Ganesan, V.; Mariappan, K.; Sukumaran, G.; Sandhya, R.; Mathew, M.D.; Bhanu Sankara Rao, K.

    2011-01-01

    Highlights: → The effects of dynamic sodium on the CFI behaviour of Mod. 9Cr-1Mo steel has investigated. → The cyclic stress response of Mod. 9Cr-1Mo steel under flowing sodium environment is similar to that of air environment. → The creep-fatigue endurance of the alloy is found to decrease with introduction of hold time and with increase in the duration of hold time and the factor of life increase in sodium compared to air environment is reduced with increase in hold time. → In contrast to air environment, tensile holds were found to be more damaging than compression hold in sodium environment. → Design rules based on air environment can be safely applied for the components operating in sodium environment. - Abstract: The use of liquid sodium as a heat transfer medium for sodium-cooled fast reactors (SFRs) necessitates a clear understanding of the effects of dynamic sodium on low cycle fatigue (LCF), creep and creep-fatigue interaction (CFI) behaviour of reactor structural materials. Mod. 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of sodium cooled fast reactors. The steam generator has a design life of 30-40 years. The effects of dynamic sodium on the LCF and CFI behaviour of Mod. 9Cr-1Mo steel have been investigated at 823 and 873 K. The CFI life of the steel showed marginal increase under flowing sodium environment when compared to air environment. Hence, the design rules for creep-fatigue interaction based on air tests can be safely applied for components operating in sodium environment. This paper attempts to explain the observed LCF and CFI results based on the detailed metallography and fractography conducted on the failed samples.

  20. Finite element modelling of the creep deformation of T91 steel weldments at 600 C

    Energy Technology Data Exchange (ETDEWEB)

    Bhadrui, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Gaudig, W. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Theofel, H. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Maile, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt

    1996-05-01

    Finite element modelling of the creep deformation of T91 steel weldments, welded using the manual metal arc (MMA) and submerged arc (SA) welding processes, was carried out to predict creep curves for both of the weldments under different stresses and compared with the experimental data. The stress and strain redistribution across the length of the transverse-weld specimens has also been predicted. Data of creep tests at 600 C at stresses between 90-130 MPa for the base metal, the MMA and SA weld metals, and the simulated heat-affected zone were used to determine Garofalo`s equation for creep strain. Finite element meshes for both of the weldments were constructed after calculating the HAZ locations using Rosenthal`s heat flow equation. (orig.)

  1. Study of the influence of liquid sodium on the mechanical behavior of T91 steel in liquid sodium

    International Nuclear Information System (INIS)

    Hemery, S.

    2013-01-01

    We studied the sensitivity of T91 steel to embrittlement by liquid sodium. An experimental procedure was set up to proceed to mechanical testing in sodium under an inert atmosphere. The introduction of a liquid sodium pre-exposure step prior to mechanical testing enabled the study of both the wettability of T91 by sodium and the structure of the sodium steel/interface as a function of the exposure parameters. The mechanical properties of T91 steel are significantly reduced in liquid sodium provided the wetting conditions are good. The use of varying oxygen and hydrogen concentrations suggests that oxygen plays a major role in enhancing the wettability of T91. The sensitivity of the embrittlement to strain rate and temperature was characterized. These results showed the existence of a ductile to brittle transition depending on both parameters. Its characterization suggests that a diffusion step is the limiting rate phenomenon of this embrittlement case. TEM and EBSD analysis of arrested cracks enabled us to establish that the fracture mode is inter-lath or intergranular. This characteristic is coherent with the crack path commonly reported in liquid metal embrittlement. A similar procedure was applied to the unalloyed XC10 steel. The results show a behavior which is similar to the one of T91 steel and suggest a common mechanism for liquid sodium embrittlement of body centered cubic steels. Moreover, they confirm that the ductile to brittle transition seems associated with a limited crack propagation rate. The propagation is thermally activated with activation energy of about 50 kJ/mol. Finally, it was shown that 304L austenitic steel is sensitive to liquid sodium embrittlement as well. Some fracture surfaces testify of an intergranular fracture mode, but some questions still remain about the crack path. (author) [fr

  2. Development of New Heats of Advanced Ferritic/Martensitic Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pestovich, Kimberly Shay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-23

    The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements in low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.

  3. Correlation between aging grade of T91 steel and spectral characteristics of the laser-induced plasma

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jun [Power Electric College in South China University of Technology, Guangzhou 510640 (China); Lu, Jidong, E-mail: jdlu@scut.edu.cn [Power Electric College in South China University of Technology, Guangzhou 510640 (China); Dai, Yuan [Electric Power Research Institute of Guangdong Power Grid Company, Guangzhou 510080 (China); Dong, Meirong [Power Electric College in South China University of Technology, Guangzhou 510640 (China); Zhong, Wanli [Electric Power Research Institute of Guangdong Power Grid Company, Guangzhou 510080 (China); Yao, Shunchun [Power Electric College in South China University of Technology, Guangzhou 510640 (China)

    2015-08-15

    Highlights: • Laser-induced breakdown spectroscopy (LIBS) was first employed to estimate the aging grade of T91 steel. • The differences from the ionic-to-atomic ratio of the same elements indicated that the grade is the reason to cause the difference of the plasma characteristics. • Good unique value correlations between the peak intensity ratio of CrI/FeI, MoI/FeI and the aging grade were found. • The research indicates that LIBS technique is a potential way to estimate the aging grade of T91 steel. - Abstract: T91 steel with favorable mechanical performance has become the representative heat-resistant steel used as heat exchange surfaces in supercritical and ultra-supercritical boilers. The organizational structure and mechanical properties change during the service period, called material aging, which affects the service life and the equipment safety. To develop a fast and easy aging predictive technique of heat exchange metal surfaces, laser-induced breakdown spectroscopy (LIBS) was applied to investigate the plasma characteristics of T91 steel specimens with different aging grades. The metallographic structure, mechanical properties and spectral characteristics of the specimens were analyzed. Then, the correlations between the spectral characteristics and the aging grade were established. The analysis results show that the martensite substructure disappears, and the dimension of the carbide particles among the crystal lattices increases with aging. At the same time, the hardness of the samples gradually decreases. The peak intensities of both the matrix and the alloying element increases then decreases with aging, owing to the change of the metallography structure and mechanical properties. Furthermore, good unique value correlations between the intensity ratio of CrI/FeI, MoI/FeI and the aging grade are found. This demonstrates that LIBS is a possible new way to estimate the aging grade of metal materials.

  4. Correlation between aging grade of T91 steel and spectral characteristics of the laser-induced plasma

    International Nuclear Information System (INIS)

    Li, Jun; Lu, Jidong; Dai, Yuan; Dong, Meirong; Zhong, Wanli; Yao, Shunchun

    2015-01-01

    Highlights: • Laser-induced breakdown spectroscopy (LIBS) was first employed to estimate the aging grade of T91 steel. • The differences from the ionic-to-atomic ratio of the same elements indicated that the grade is the reason to cause the difference of the plasma characteristics. • Good unique value correlations between the peak intensity ratio of CrI/FeI, MoI/FeI and the aging grade were found. • The research indicates that LIBS technique is a potential way to estimate the aging grade of T91 steel. - Abstract: T91 steel with favorable mechanical performance has become the representative heat-resistant steel used as heat exchange surfaces in supercritical and ultra-supercritical boilers. The organizational structure and mechanical properties change during the service period, called material aging, which affects the service life and the equipment safety. To develop a fast and easy aging predictive technique of heat exchange metal surfaces, laser-induced breakdown spectroscopy (LIBS) was applied to investigate the plasma characteristics of T91 steel specimens with different aging grades. The metallographic structure, mechanical properties and spectral characteristics of the specimens were analyzed. Then, the correlations between the spectral characteristics and the aging grade were established. The analysis results show that the martensite substructure disappears, and the dimension of the carbide particles among the crystal lattices increases with aging. At the same time, the hardness of the samples gradually decreases. The peak intensities of both the matrix and the alloying element increases then decreases with aging, owing to the change of the metallography structure and mechanical properties. Furthermore, good unique value correlations between the intensity ratio of CrI/FeI, MoI/FeI and the aging grade are found. This demonstrates that LIBS is a possible new way to estimate the aging grade of metal materials

  5. Creep Rupture of the Simulated HAZ of T92 Steel Compared to that of a T91 Steel

    Directory of Open Access Journals (Sweden)

    Yu-Quan Peng

    2017-02-01

    Full Text Available The increased thermal efficiency of fossil power plants calls for the development of advanced creep-resistant alloy steels like T92. In this study, microstructures found in the heat-affected zone (HAZ of a T92 steel weld were simulated to evaluate their creep-rupture-life at elevated temperatures. An infrared heating system was used to heat the samples to 860 °C (around AC1, 900 °C (slightly below AC3, and 940 °C (moderately above AC3 for one minute, before cooling to room temperature. The simulated specimens were then subjected to a conventional post-weld heat treatment (PWHT at 750 °C for two hours, where both the 900 °C and 940 °C simulated specimens had fine grain sizes. In the as-treated condition, the 900 °C simulated specimen consisted of fine lath martensite, ferrite subgrains, and undissolved carbides, while residual carbides and fresh martensite were found in the 940 °C simulated specimen. The results of short-term creep tests indicated that the creep resistance of the 900 °C and 940 °C simulated specimens was poorer than that of the 860 °C simulated specimens and the base metal. Moreover, simulated T92 steel samples had higher creep strength than the T91 counterpart specimens.

  6. Creep Rupture of the Simulated HAZ of T92 Steel Compared to that of a T91 Steel.

    Science.gov (United States)

    Peng, Yu-Quan; Chen, Tai-Cheng; Chung, Tien-Jung; Jeng, Sheng-Long; Huang, Rong-Tan; Tsay, Leu-Wen

    2017-02-08

    The increased thermal efficiency of fossil power plants calls for the development of advanced creep-resistant alloy steels like T92. In this study, microstructures found in the heat-affected zone (HAZ) of a T92 steel weld were simulated to evaluate their creep-rupture-life at elevated temperatures. An infrared heating system was used to heat the samples to 860 °C (around A C1 ), 900 °C (slightly below A C3 ), and 940 °C (moderately above A C3 ) for one minute, before cooling to room temperature. The simulated specimens were then subjected to a conventional post-weld heat treatment (PWHT) at 750 °C for two hours, where both the 900 °C and 940 °C simulated specimens had fine grain sizes. In the as-treated condition, the 900 °C simulated specimen consisted of fine lath martensite, ferrite subgrains, and undissolved carbides, while residual carbides and fresh martensite were found in the 940 °C simulated specimen. The results of short-term creep tests indicated that the creep resistance of the 900 °C and 940 °C simulated specimens was poorer than that of the 860 °C simulated specimens and the base metal. Moreover, simulated T92 steel samples had higher creep strength than the T91 counterpart specimens.

  7. Development of ODS ferritic-martensitic steels for application to high temperature and irradiation environment; Developpement d'une nouvelle nuance martensitique ODS pour utilisation sous rayonnement a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Lambard, V

    2000-07-01

    Iron oxide dispersion strengthened alloys are candidate for nuclear fuel cladding. Therefore, it is crucial to control their microstructure in order to optimise their mechanical properties at temperatures up to 700 deg C. The industrial candidates, ODS ferritic alloys, present an anisotropic microstructure which induces a weakening of mechanical properties in transversal direction as well as the precipitation of brittle phases under thermal aging and irradiation. For this purpose, we tried to develop a material with isotropic properties. We studied several 9Cr-1Mo ferritic/martensitic alloys, strengthened or not by oxide dispersion. The mechanical alloying was performed by attribution and powders were consolidated by hot extrusion. In this work, different metallurgical characterisation techniques and modelling were used to optimise a new martensitic ODS alloy. Microstructural and chemical characterization of matrix has been done. The effect of austenitizing and isochronal tempering treatments on microstructure and hardness has been studied. Oxide distribution, size and chemical composition have been studied before and after high temperature thermal treatment. The study of phase transformation upon heating has permitted the extrapolation to the equilibrium temperature formation of austenite. Phase transformation diagrams upon cooling have been determined and the transformation kinetics have been linked to austenite grain size by a simple relation. Fine grain size is unfavourable for the targeted application, so a particular thermal treatment inducing a coarser grain structure has been developed. Finally, tensile properties have been determined for the different microstructures. (author)

  8. Estimation of the aging grade of T91 steel by laser-induced breakdown spectroscopy coupled with support vector machines

    Science.gov (United States)

    Lu, Shengzi; Dong, Meirong; Huang, Jianwei; Li, Wenbing; Lu, Jidong; Li, Jun

    2018-02-01

    T91 steel is a representative martensitic heat-resistant steel widely used in high temperature compression components of industrial equipment. During the service period, the operation safety and the service life of the equipment will be affected by the change of structure and mechanical properties of the steel components, which is called material aging. In order to develop a rapid in-situ aging estimation technology of high temperature compression components surface, laser-induced breakdown spectroscopy (LIBS) coupled with support vector machine (SVM) was employed in this paper. The spectral characteristics of 10 T91 steel specimens with different aging grades were analyzed. Line intensities and the line intensity ratios (ionic/atomic and alloying element/matrix element) that indicate the change of metallographic structure were used to establish SVM models, and the results using different variable sets were compared. The model was optimized by comparing different pulse number for practical effectiveness, and the robustness of the model was investigated in dealing with the inhomogeneity of steel composition. The study results show that the estimation model obtained the best performance using line intensities and line intensity ratios averaged from 31st-60th laser pulses as input variables. The estimation accuracy of validation set was greatly improved from 75.8% to 95.3%. In addition, the model showed the outstanding capacity for handling the fluctuations of spectral signals between measuring-points (spots), which indicated that the aging estimation based on a few measuring-points is feasible. The studies presented here demonstrate that the LIBS coupled with SVM is a new useful technique for the aging estimation of steel, and would be well-suited for fast safety assessment in industrial field.

  9. Role of grain boundary engineering in the SCC behavior of ferritic-martensitic alloy HT-9

    International Nuclear Information System (INIS)

    Gupta, G.; Ampornrat, P.; Ren, X.; Sridharan, K.; Allen, T.R.; Was, G.S.

    2007-01-01

    This paper focuses on the role of grain boundary engineering (GBE) in stress corrosion cracking (SCC) of ferritic-martensitic (F-M) alloy HT-9 in supercritical water (SCW) at 400 deg. C and 500 deg. C. Constant extension rate tensile (CERT) tests were conducted on HT-9 in as-received (AR) and coincident site lattice enhanced (CSLE) condition. Both unirradiated and irradiated specimens (irradiated with 2 MeV protons at 400 deg. C and 500 deg. C to a dose of 7 dpa) were tested. Ferritic-martensitic steel HT-9 exhibited intergranular stress corrosion cracking when subjected to CERT tests in an environment of supercritical water at 400 deg. C and 500 deg. C and also in an inert environment of argon at 500 deg. C. CSL-enhancement reduces grain boundary carbide coarsening and cracking susceptibility in both the unirradiated and irradiated condition. Irradiation enhanced coarsening of grain boundary carbides and cracking susceptibility of HT-9 for both the AR and CSLE conditions. Intergranular (IG) cracking of HT-9 results likely from fracture of IG carbides and seems consistent with the mechanism that coarser carbides worsen cracking susceptibility. Oxidation in combination with wedging stresses is the likely cause of the observed environmental enhancement of high temperature IG cracking in HT-9

  10. Microstructure and Nano-Hardness of 10 MeV Cl-Ion Irradiated T91 Steel

    International Nuclear Information System (INIS)

    Hu Jing; Wang Xianping; Gao Yunxia; Zhuang Zhong; Zhang Tao; Fang Qianfeng; Liu Changsong

    2015-01-01

    Hardening and elemental segregation of T91 martenstic steel irradiated by 10 MeV Cl ions to doses from 0.06 dpa to 0.83 dpa were investigated with the nanoindentation technique and transmission electron microscopy (TEM). The results demonstrated that the irradiation hardening was closely related with irradiation dose. By increasing the dose, the hardness increased rapidly at first from the initial value of 3.15 GPa before irradiation, and then tended to saturate at a value of 3.58 GPa at the highest dose of 0.83 dpa. Combined with TEM observation, the mechanism of hardening was preliminary attributed to the formation of M(Fe,Cr) 2 3C 6 carbides induced by the high energy Cl-ion irradiation. (paper)

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

    Science.gov (United States)

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

    2017-12-01

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

  12. TEM Characterization of Helium Bubbles in T91 and MNHS Steels Implanted with 200 keV He Ions at Different Temperatures

    International Nuclear Information System (INIS)

    Wang Ji; Gao Xing; Wang Zhi-Guang; Wei Kong-Fang; Yao Cun-Feng; Cui Ming-Huan; Sun Jian-Rong; Li Bing-Sheng; Pang Li-Long; Zhu Ya-Bin; Luo Peng; Chang Hai-Long; Zhang Hong-Peng; Zhu Hui-Ping; Wang Dong; Du Yang-Yang; Xie Er-Qing

    2015-01-01

    Modified novel high silicon steel (MNHS, a newly developed reduced-activation martensitic alloy) and commercial alloy T91 are implanted with 200 keV He"2"+ ions to a dose of 5 × 10"2"0 ions/m"2 at 300, 450 and 550°C. Transmission electron microscopy (TEM) is used to characterize the size and morphology of He bubbles. With the increase of the implantation temperature, TEM observations indicate that bubbles increase in size and the proportion of ‘brick shaped’ cuboid bubbles increases while the proportion of polyhedral bubbles decreases in both the steel samples. For the samples implanted at the same temperature, the average size of He bubbles in MNHS is smaller than that in T91. This might be due to the abundance of boundaries and precipitates in MNHS, which provide additional sites for the trapping of He atoms, thus reduce the susceptibility of MNHS to He embrittlement. (paper)

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

    Science.gov (United States)

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

    2018-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-15

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

  15. A comparative study on the compatibility of liquid lead–gold eutectic and liquid lead–bismuth eutectic with T91 and SS 316LN steels

    International Nuclear Information System (INIS)

    Dai, Y.; Gao, W.; Zhang, T.; Platacis, E.; Heinitz, S.; Thomsen, K.

    2012-01-01

    Liquid lead–gold eutectic (LGE) is considered as a potential target material for high power spallation sources. In the present work, the corrosion effects of LGE on T91 and SS 316LN steels have been investigated in comparison with that of liquid lead–bismuth eutectic (LBE) under the same testing conditions. Two tests were conducted at 400 °C for 1800 h and at 450 °C for 1300 h, in which specimens of the two steels were exposed to 1 m/s flowing LGE and LBE. Surface inspections showed that the specimens underwent a mixed corrosion mode of dissolution and oxidation. The results obtained from the SS 316LN specimens are very interesting. Firstly, EDX (electron energy dispersion X-ray spectrometry) analyses revealed that Ni, Cr and Mn have a higher dissolution rate in LGE than in LBE. Secondly, it was observed that LBE attacked strongly on grain-boundaries (GBs) and twin-boundaries (TBs), while LGE did not preferentially attack GBs and TBs. Further, the diffusion or penetration paths of LBE look straight, while those of LGE look like a complex network. In the attacked regions the chemical composition of the steel did not change much in the LBE case, but changed greatly in the LGE case. The T91 specimens exhibited considerably weaker corrosion effects under the present testing conditions.

  16. Elevated Temperature Corrosion Studies of AlCrN and TiAlN Coatings by PAPVD on T91 Boiler Steel

    Science.gov (United States)

    Goyal, Lucky; Chawla, Vikas; Hundal, Jasbir Singh

    2017-11-01

    The present investigation discusses the hot corrosion behavior of AlCrN and TiAlN nano-coatings on T91 boiler steel by PAPVD process subjected to molten salt of Na2SO4-60%V2O5 at 900 °C for 50 cycles. Surface and cross-sectional studies were performed by AFM, SEM/EDS and XRD techniques to understand the corrosion kinetics and mechanism. T91 bare boiler steel as well as TiAlN-coated specimen has shown higher internal oxidation as well as weight gain. The better corrosion resistance of AlCrN-coated specimen has been observed by virtue of higher availability of Cr and Al in the oxide scale as well as adherent and dense coating. The betterment of AlCrN coating can be attributed to low internal oxidation as well as movement of Cr and Al toward oxide scale to form protective corrosion barriers.

  17. Radiation response of alloy T91 at damage levels up to 1000 peak dpa

    Energy Technology Data Exchange (ETDEWEB)

    Gigax, J.G., E-mail: gigaxj@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Chen, T.; Kim, Hyosim [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Wang, J. [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Pacific Northwest National Lab, Richland, WA 99354 (United States); Price, L.M. [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Aydogan, E. [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Los Alamos National Lab, Los Alamos, NM 87545 (United States); Maloy, S.A. [Los Alamos National Lab, Los Alamos, NM 87545 (United States); Schreiber, D.K.; Toloczko, M.B. [Pacific Northwest National Lab, Richland, WA 99354 (United States); Garner, F.A. [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States); Radiation Effects Consulting, Richland, WA 99354 (United States); Shao, Lin [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77840 (United States)

    2016-12-15

    Ferritic/martensitic alloys are required for advanced reactor components to survive 500–600 neutron-induced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 °C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500–600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

  18. Radiation response of alloy T91 at damage levels up to 1000 peak dpa

    Energy Technology Data Exchange (ETDEWEB)

    Gigax, J. G.; Chen, T.; Kim, Hyosim; Wang, J.; Price, L. M.; Aydogan, E.; Maloy, S. A.; Schreiber, D. K.; Toloczko, M. B.; Garner, F. A.; Shao, Lin

    2016-12-01

    Ferritic/martensitic alloys are required for advanced reactor components to survive 500e600 neutroninduced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500e600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

  19. Effect of welding thermal cycles on the structure and properties of simulated heat-affected zone areas in X10CrMoVNb9-1 (T91) steel at a state after 100,000 h of operation

    Energy Technology Data Exchange (ETDEWEB)

    Łomozik, Mirosław, E-mail: miroslaw.lomozik@is.gliwice.pl [Instytut Spawalnictwa, Testing of Materials Weldability and Welded Constructions Department, 44-100 Gliwice, Bł. Czesława 16-18 (Poland); Hernas, Adam, E-mail: adam.hernas@polsl.pl [Silesian University of Technology, Faculty of Materials Engineering and Metallurgy, 40-019 Katowice, Krasińskiego 8 str. (Poland); Zeman, Marian L., E-mail: marian.zeman@is.gliwice.pl [Instytut Spawalnictwa, Testing of Materials Weldability and Welded Constructions Department, 44-100 Gliwice, Bł. Czesława 16-18 (Poland)

    2015-06-18

    The article presents results of structural tests (light, scanning electron and scanning transmission electron microscopy) of X10CrMoVNb9-1 (T91) creep-resisting steel after approximately 100,000 h of operation. It was ascertained that the parent metal of T91 steel is characterized by the microstructure of tempered martensite with M{sub 23}C{sub 6} carbide precipitates and few dispersive precipitates of MX-type niobium and vanadium carbonitrides. The most inconvenient change in T91 steel precipitate morphology due to long-term operation is the appearance of the Laves Fe{sub 2}Mo phase which along with M{sub 23}C{sub 6} carbide particles forms elongated blocks and conglomerates on grain boundaries. The article also presents results of tests related to the effect of simulated welding thermal cycles on selected properties of X10CrMoVNb9-1 (T91) grade steel at a state after approximately 100,000 h of operation. The tests involved the determination of the chemical composition of the steel tested as well as impact tests, hardness measurements and microscopic metallographic examination (based on light microscopy) of simulated heat-affected zone (HAZ) areas for a cooling time (t{sub 8/5}) restricted within a range between 3 s and 120 s, with and without heat treatment. The tests revealed that, among other results, hardness values of simulated HAZ areas in X10CrMoVNb9-1 (T91) steel do not guarantee cold crack safety of the steel at the state without additional heat treatment. It was also observed that simulated welding thermal cycles of cooling times t{sub 8/5}=3, 12, 60 and 120 s do not significantly affect the toughness and hardness of simulated HAZ areas of the steel tested.

  20. Long-term strength and allowable stresses of grade 10Kh9MFB and X10CrMoVNb9-1 (T91/P91) chromium heat-resistant steels

    Science.gov (United States)

    Skorobogatykh, V. N.; Danyushevskiy, I. A.; Schenkova, I. A.; Prudnikov, D. A.

    2015-04-01

    Currently, grade X10CrMoVNb9-1 (T91, P91) and 10Kh9MFB (10Kh9MFB-Sh) chromium steels are widely applied in equipment manufacturing for thermal power plants in Russia and abroad. Compilation and comparison of tensile, impact, and long-term strength tests results accumulated for many years of investigations of foreign grade X10CrMoVNb9-1, T91, P91, and domestic grade 10Kh9MFB (10Kh9MFB-Sh) steels is carried out. The property identity of metals investigated is established. High strength and plastic properties of steels, from which pipes and other products are made, for operation under creep conditions are confirmed. Design characteristics of long-term strength on the basis of tests with more than one million of hour-samples are determined ( and at temperatures of 500-650°C). The table of recommended allowable stresses for grade 10Kh9MFB, 10Kh9MFB-SH, X10CrMoVNb9-1, T91, and P91 steels is developed. The long-time properties of pipe welded joints of grade 10Kh9MFB+10Kh9MFB, 10Kh9MFB-Sh+10Kh9MFB-Sh, X10CrMoVNb9-1+X10CrMoVNb9-1, P91+P91, T91+T91, 10Kh9MFB (10Kh9MFB-Sh)+X10CrMoVNb9-1(T/P91) steels is researched. The welded joint reduction factor is experimentally determined.

  1. Corrosion mechanism of T91 steel by Pb-Bi eutectic used as spallation target: importance for accelerator driven system; Mecanisme de corrosion de l'acier T91 par l'eutectique Pb-Bi utilise comme materiau de cible de spallation: importance pour les reacteurs hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Martinelli, L

    2005-10-15

    The aim of this work has been to determine the oxidation mechanism of the martensitic steel T91 in the Pb-Bi liquid eutectic alloy, saturated in oxygen, at 470 C, in order to develop a long-term predictive model of the oxidation kinetics of the steel. This work enters in the framework of the lifetime studies of the spallation module demonstrator: MEGAPIE for the researches on hybrid reactors. An experimental characterization of the oxide layers has been carried out as well as the oxidation kinetics of the T91 steel. An oxidation mechanism has been elaborated from these experimental results and then simulated. The oxide layer formed at the T91 surface presents a duplex structure constituted by a magnetite external layer and a spinel Fe-Cr internal layer. A growth mechanism of the oxide layers has been proposed: the growth of the magnetite layer seems to be limited by the iron diffusion in the lattice of the duplex oxide layer. In parallel, an auto-regulation mechanism seems to govern the growth of the Fe-Cr spinel layer. This mechanism includes a non-limiting step of the oxygen diffusion in the oxide layer (by liquid way in the nano-channels of lead), as well as a limiting step of iron diffusion in the lattice of the oxide layer. In considering the proposed oxidation mechanisms, a simulation of the growth of the two oxide layers is carried out and compared to the long-time oxidation growth kinetics. The good agreement between the experimental results allows, finally, to strengthen the proposition of a long-term growth kinetic oxidation mechanism of the oxide layers. (O.M.)

  2. High burnup performance of an advanced oxide fuel assembly in FFTF [Fast Flux Test Facility] with ferritic/martensitic materials

    International Nuclear Information System (INIS)

    Bridges, A.E.; Saito, G.H.; Lovell, A.J.; Makenas, B.J.

    1986-05-01

    An advanced oxide fuel assembly with ferritic/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd/KgM and a peak fast fluence greater than 15 x 10 22 n/cm 2 . The cladding, wire-wrap, and duct material for the ACO-1 test assembly is the ferritic/martensitic alloy, HT9, which was chosen for use in long-lifetime fuel assemblies because of its good nominal temperature creep strength and low swelling rate. Valuable experience on the performance of HT9 materials has been gained from this test, advancing our quest for long-lifetime fuel. Pertinent data, obtained from the ACO-1 test assembly, will support the irradiation of the Core Demonstration Experiment in FFTF

  3. Interdiffusion behavior of Al-rich oxidation resistant coatings on ferritic-martensitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Velraj, S.; Zhang, Y.; Hawkins, E.W. [Department of Mechanical Engineering, Tennessee Technological University, Cookeville, TN 38505-0001 (United States); Pint, B.A. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6156 (United States)

    2012-10-15

    Interdiffusion of thin Al-rich coatings synthesized by chemical vapor deposition (CVD) and pack cementation on 9Cr ferritic-martensitic alloys was investigated in the temperature range of 650-700 C. The compositional changes after long-term exposures in laboratory air and air + 10 vol% H{sub 2}O were examined experimentally. Interdiffusion was modeled by a modified coating oxidation and substrate interdiffusion model (COSIM) program. The modification enabled the program to directly input the concentration profiles of the as-deposited coating determined by electron probe microanalysis (EPMA). Reasonable agreement was achieved between the simulated and experimental Al profiles after exposures. The model was also applied to predict coating lifetime at 650-700 C based on a minimum Al content (C{sub b}) required at the coating surface to re-form protective oxide scale. In addition to a C{sub b} value established from the failure of a thin CVD coating at 700 C, values reported for slurry aluminide coatings were also included in lifetime predictions. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  4. Corrosion behavior of steels in flowing lead-bismuth under abnormal conditions

    International Nuclear Information System (INIS)

    Doubkova, A.; Di Gabriele, F.; Brabec, P.; Keilova, E.

    2008-01-01

    The project IP EUROTRANS, domain DEMETRA, is primary focused on the study of the technology of the interaction between steels and heavy liquid metals. The characterization of the metal response to sudden changes, simulating accidental conditions in liquid lead-bismuth eutectic was carried out. This paper reports the results of two hot-spot simulations with two different oxygen concentrations (10 -8 wt%, 10 -6 wt%). Each experiment was divided in two main periods: the initial, long period at the standard operating temperature 550 deg. C; the second, short period, at higher temperature, 650 deg. C. The damage that occurs on the austenitic steel AISI 316L and the ferritic-martensitic steel T91 was investigated. The amount of damage for both steels was higher at lower oxygen contents and the short, hot spot simulation, markedly affected the T91. At higher oxygen content the amount of damage decreased. A few, localized pits, were observed; however, there was no visible increment in the amount of damage after the hot spot simulation

  5. Oxidation behavior of steels and Alloy 800 in supercritical water

    International Nuclear Information System (INIS)

    Olmedo, A.M.; Bordoni, R.; Dominguez, G.; Alvarez, M.G.

    2011-01-01

    The oxidation behavior of a ferritic-martensitic steel T91 and a martensitic steel AISI 403 up to 750 h, and of AISI 316L and Alloy 800 up to 336 h in deaerated supercritical water, 450ºC-25 MPa, was investigated in this paper. After exposure up to 750 h, the weight gain data, for steels T91 and AISI 403, was fitted by ∆W=k t n , were n are similar for both steels and k is a little higher for T91. The oxide films grown in the steels were characterized using gravimetry, scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction. The films were adherent and exhibited a low porosity. For this low oxygen content supercritical water exposure, the oxide scale exhibited a typical duplex structure, in which the scale is composed of an outer iron oxide layer of magnetite (Fe 3 O 4 ) and an inner iron/chromium oxide layer of a non-stoichiometric iron chromite (Fe,Cr) 3 O 4 . Preliminary results, with AISI 316L and Alloy 800, for two exposure periods (168 and 336 h), are also reported. The morphology shown for the oxide films grown on both materials up to 336 h of oxidation in supercritical water, resembles that of a duplex layer film like that shown by stainless steels and Alloy 800 oxide films grown in a in a high temperature and pressure (220-350ºC) of a primary or secondary coolant of a plant. (author) [es

  6. Effect of irradiation temperature on microstructure of ferritic-martensitic ODS steel

    Science.gov (United States)

    Klimenkov, M.; Lindau, R.; Jäntsch, U.; Möslang, A.

    2017-09-01

    The EUROFER-ODS alloy with 0.5% Y2O3 was neutron irradiated with doses up to 16.2 dpa at 250 °C, 350 °C and 450 °C. The radiation induced changes in the microstructure (e.g. dislocation loops and voids) were investigated using transmission electron microscopy (TEM). The number density of radiation induced defects was found to be significantly lower than in EUROFER 97 irradiated at the same conditions. It was found that the appearance and extent of radiation damage strongly depend not only on the irradiation temperature but also on the local number density and size distribution of ODS particles. The higher number density of dislocation loops and voids was found in the local areas with low number density of ODS particles. The interstitial loops with Burgers vector of both ½ and types were detected by imaging using different diffraction conditions.

  7. Microstructural Evolution of Advanced Radiation-Resistant ODS Steel with Different Lengths of Mechanical Alloying Time

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon; Kim, Ga Eon; Kang, Suk Hoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/ martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, powder properties and microstructures of the ODS steel with different length of mechanical alloying time was investigated. The ODS steel milled 5h showed homogeneous grain structure with the highest hardness.

  8. Microstructural Evolution of Advanced Radiation-Resistant ODS Steel with Different Lengths of Mechanical Alloying Time

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Kim, Ga Eon; Kang, Suk Hoon; Kim, Tae Kyu

    2015-01-01

    Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/ martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, powder properties and microstructures of the ODS steel with different length of mechanical alloying time was investigated. The ODS steel milled 5h showed homogeneous grain structure with the highest hardness

  9. Influence of the initial metallurgical state and the austenizing conditions on the distribution of austenitc grain size of the martensitic-ferritic steel T91(9%Cr-1%Mo-V-Nb)

    International Nuclear Information System (INIS)

    Zavaleta Gutierrez, N; Luppo, M.I; Danon, C.A; Garcia de Andres, C

    2006-01-01

    It is a known fact that the austenizing conditions (speed of heating to the austenite temperature and austenizing time) as well as the initial metallurgical state of the material strongly influence the distribution of austenitic grain size in steels. This distribution will be one of the parameters that will control the behavior of the material in a later transformation from the austenite -by continuous cooling or in the isothermal case - and this behavior will determine the product's final mechanical properties. Based on the published literature, we have studied the influence of the initial metallurgical state and the speed of heating to austenite on the distribution of austenitic grain size for a certain austenizing temperature and time for a martensitic-iron ASTM A213 grade T91 steel. Two-stage thermal cycles were designed for this, that is, tempering for a variable period of time at the industrial tempering temperature (780 o C) followed by the austenizing (1050 o C, 30 minutes) 'in situ'. We have analyzed the following as a whole: 1) the role of the stabilizing elements (Nb, V) that eventually control the anchoring of the austenitic grain boundary by carbide or carbonitride precipitation. Therefore, we have tried to vary the fraction of these elements present in solid solution by annealing before austenizing. 2) the role of the speed of heating to austenite. In this case, we have considered two different values (1 and 30 o C/s), previously reported as inferior and superior, respectively, to the speed of 'critical' heat needed to produce a distribution of heterogeneous austenitic grain size when the metallurgical state before the austenizing is quenched and tempered. Preliminary results suggest that a annealing stage after tempering in the plant and prior to eventual austenizing significantly reduces the influence of the heating to austenite speed in the development of a heterogeneous structure of austenitic grains (CW)

  10. Nanostructures in a ferritic and an oxide dispersion strengthened steel induced by dynamic plastic deformation

    DEFF Research Database (Denmark)

    Zhang, Zhenbo

    fission and fusion reactors. In this study, two candidate steels for nuclear reactors, namely a ferritic/martensitic steel (modified 9Cr-1Mo steel) and an oxide dispersion strengthened (ODS) ferritic steel (PM2000), were nanostructured by dynamic plastic deformation (DPD). The resulting microstructure...

  11. Effect of M{sub 3}C on the precipitation behavior of M{sub 23}C{sub 6} phase during early stage of tempering in T91 ferritic steel

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Chenxi; Liu, Yongchang; Zhang, Dantian; Ning, Baoqun; Yan, Zesheng [School of Material Science and Engineering, Tianjin Key Laboratory of Advanced Jointing Technology, Tianjin University, Tianjin (China)

    2011-12-15

    Tempered martensitic structure is the service condition of T91 ferritic steel after adopting the austenitizing followed by tempering. Needle-like M{sub 3}C particles are precipitated during air cooling after austenization, while the precipitation of M{sub 3}C is suppressed during the water cooling. The effect of existence of M{sub 3}C on the precipitation behaviors of M{sub 23}C{sub 6} during the early stage of tempering, as nucleation site, number density and size distribution, was investigated by means of TEM observation. The TEM results indicate that, upon the same tempering time, the size of M{sub 23}C{sub 6} is smaller and its number density is higher in the sample pre-existing M{sub 3}C than in the sample without M{sub 3}C. This can be explained that existence of M{sub 3}C results in more M{sub 23}C{sub 6} precipitates forming inside of grain, where a relatively low self-diffusion coefficient of alloy element leads to M{sub 23}C{sub 6} hardly coarsening. However, with the prolongation of tempering time, this effect becomes weaken. Microhardness results indicate that the existence of M{sub 3}C phase results in the increase of hardness after tempering due to the precipitation of finer and denser M{sub 23}C{sub 6} particles. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. High temperature creep strength of Advanced Radiation Resistant Oxide Dispersion Strengthened Steels

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling well occurred to 120dpa at high temperatures and this leads the decrease of the mechanical properties and dimensional stability. Compared to this, ferritic/martensitic steel is a good solution because of excellent thermal conductivity and good swelling resistance. Unfortunately, the available temperature range of ferritic/martensitic steel is limited up to 650 .deg. C. ODS steel is the most promising structural material because of excellent creep and irradiation resistance by uniformly distributed nano-oxide particles with a high density which is extremely stable at the high temperature in ferritic/martensitic matrix. In this study, high temperature strength of advanced radiation resistance ODS steel was investigated for the core structural material of next generation nuclear systems. ODS martensitic steel was designed to have high homogeneity, productivity and reproducibility. Mechanical alloying, hot isostactic pressing and hot rolling processes were employed to fabricate the ODS steels, and creep rupture test as well as tensile test were examined to investigate the behavior at high temperatures. ODS steels were fabricated by a mechanical alloying and hot consolidation processes. Mechanical properties at high temperatures were investigated. The creep resistance of advanced radiation resistant ODS steels was more superior than those of ferritic/ martensitic steel, austenitic stainless steel and even a conventional ODS steel.

  13. Hydrogen effects in stainless steel

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1983-01-01

    The effects of hydrogen on stainless steels have been reviewed and are summarized in this paper. Discussion covers hydrogen solution and transport in stainless steels as well as the effects of hydrogen on deformation and fracture under various loading conditions. Damage is caused also by helium that arises from decay of the hydrogen isotope tritium. Austenitic, ferritic, martensite, and precipitation-hardenable stainless steels are included in the discussion. 200 references

  14. 75 FR 19369 - Certain Hot-Rolled Flat-Rolled Carbon Quality Steel Products from Brazil: Preliminary Results of...

    Science.gov (United States)

    2010-04-14

    .... Hot-rolled dual phase steel, phase-hardened, primarily with a ferritic-martensitic microstructure.... See Preliminary Results of Antidumping Duty Administrative Review: Stainless Steel Sheet and Strip in... Antidumping Duty Administrative Review: Stainless Steel Sheet and Strip in Coils From France, 68 FR 69379...

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

    International Nuclear Information System (INIS)

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

    1985-03-01

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

  16. Effects of welding on toughness of Mod. 9Cr-1Mo steel

    International Nuclear Information System (INIS)

    Ryu, W. S.; Kim, S. H.; Yoon, J. H.

    2008-01-01

    Nuclear energy is being seriously considered to meet the increasing demand for a world-wide energy supply without environmental effects. Generation IV reactors are being developed to produce a reliable energy safely and with an economic benefit. Since these new reactors require an elevated temperature, ferritic/martensitic steels are attracting attention as candidate materials for the reactor vessel of a very high temperature reactor (VHTR) and the cladding of a sodium fast reactor (SFR,) due to their high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. in recent years, new ferritic/martensitic steels have been developed for ultra supercritical fossil power plants. Advanced technologies for a steel fabrication have improved the elevated temperature properties of ferritic/martensitic steels to make them comparable with austenitic stainless steels. The microstructural stability of the pressure vessel, cladding and core structural materials of the VHTR and SCWR is very important. Welding process affects the microstructure and residual stress, so the toughness of ferritic/martensitic steels decreases in general. In this paper; Mod. 9Cr-1Mo steel is welded by SMAW with V-groove, and the effects of welding on tensile and impact properties are evaluated. The upper self energy of the weldment was only 57% of that of the base metal, and the DBTT T 41J and T 68J index temperatures of the weldment were higher than those of the base metal by 17 deg. C, 38 deg. C and 37 deg. C, respectively. (authors)

  17. Oxide Evolution in ODS Steel Resulting From Friction Stir Welding

    Science.gov (United States)

    2014-06-01

    the SZ on both the AS and RS of 304L stainless steel , from [16]. ...........................................12  Figure 7.  Past research conditions...being done on void swelling and embrittlement effects. Reduced activation ferritic/ martensitic (RAFM) steels and oxide dispersion strengthened (ODS...growth by grain boundary pinning at higher temperatures. Another type of ODS steel is 9-Cr martensitic steel , which is not considered in this research

  18. Modification of vacuum plasma sprayed tungsten coating on reduced activation ferritic/martensitic steels by friction stir processing

    International Nuclear Information System (INIS)

    Tanigawa, Hiroyasu; Ozawa, Kazumi; Morisada, Yoshiaki; Noh, Sanghoon; Fujii, Hidetoshi

    2015-01-01

    Highlights: • Friction stir processing (FSP) was applied on vacuum plasma spray (VPS) W to improve its low thermal conductivity and weakness due to high porosity. • FSP can achieve significant improvement both in mechanical and thermal properties of VPS-W coating. • It was indicated that the double pass FSP at 600 rpm/50 mm/min/2 ton on VPS-W show the most dense microstructure and hardest mechanical property. • Hardness test over FSPed VPS-W layer revealed that the hardness of W becomes higher than that of bulk W. • The thermal conductivity of double pass FSPed VPS-W was about 80% of bulk W at 200 °C, and it becomes equivalent to that of bulk W over 800 °C. - Abstract: Tungsten (W) is the primary candidate material as a plasma facing material in fusion devices, as for its high melting temperature, good thermal conductivity and low sputtering rate, and vacuum plasma spray (VPS) technique is preferred as it is applicable for large area without brittle interlayer, but the thermal conductivity of W layer is very poor, and easy to detach, mainly caused by its porous structure. W Friction stir processing (FSP) was applied on VPS-W to improve these poor properties, and it was suggested that FSP can contribute to significant improvement in both mechanical and thermal properties of the VPS-W coating.

  19. Ultrafine grained steels processed by equal channel angular pressing

    International Nuclear Information System (INIS)

    Shin, Dong Hyuk; Park, Kyung-Tae

    2005-01-01

    Recent development of ultrafine grained (UFG) low carbon steels by using equal channel angular pressing (ECAP) and their room temperature tensile properties are reviewed, focusing on the strategies overcoming their inherent mechanical drawbacks. In addition to ferrite grain refinement, when proper post heat treatments are imposed, carbon atom dissolution from pearlitic cementite during ECAP can be utilized for microstructural modification such as uniform distribution of nano-sized cementite particles or microalloying element carbides inside UFG ferrite grains and fabrication of UFG ferrite/martensite dual phase steel. The utilization of nano-sized particles is effective on improving thermal stability of UFG low carbon ferrite/pearlite steel but less effective on improving its tensile properties. By contrast, UFG ferrite/martensite dual phase steel exhibits an excellent combination of ultrahigh strength, large uniform elongation and extensive strain hardenability

  20. Mechanical properties of a 316L/T91 weld joint tested in lead-bismuth liquid

    International Nuclear Information System (INIS)

    Serre, Ingrid; Vogt, Jean-Bernard

    2009-01-01

    The mechanical strength of T91/316L weld joint assembled by electron beam process is investigated in air and in a liquid lead bismuth bath at 300 and 380 o C using the small punch test. It is shown that the mechanical response in air of the weld joint is similar to that of the T91 base material. The plastic deformation is mainly concentrated in the T91 part of the weld joint which promotes cracking in this material. Testing in liquid lead bismuth bath results in a reduction in ductility and the formation of brittle cracks. The T91/weld interface is found to be rather resistant as it cracks late in the test and after a large crack propagated in the T91 steel.

  1. Temperature dependence of liquid metal embrittlement susceptibility of a modified 9Cr–1Mo steel under low cycle fatigue in lead–bismuth eutectic at 160–450 °C

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Xing, E-mail: gongxingzfl@hotmail.com [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400 Mol (Belgium); KU Leuven, Department of Materials Engineering (MTM), Kasteelpark Arenberg 44, Box 2450, B-3001 Heverlee (Belgium); Marmy, Pierre, E-mail: pmarmy@sckcen.be [SCK-CEN (Belgian Nuclear Research Centre), Boeretang 200, B-2400 Mol (Belgium); Qin, Ling; Verlinden, Bert; Wevers, Martine [KU Leuven, Department of Materials Engineering (MTM), Kasteelpark Arenberg 44, Box 2450, B-3001 Heverlee (Belgium); Seefeldt, Marc, E-mail: Marc.Seefeldt@mtm.kuleuven.be [KU Leuven, Department of Materials Engineering (MTM), Kasteelpark Arenberg 44, Box 2450, B-3001 Heverlee (Belgium)

    2016-01-15

    Low cycle fatigue properties of a 9Cr–1Mo ferritic-martensitic steel (T91) have been tested in a low oxygen concentration (LOC) lead–bismuth eutectic (LBE) environment and in vacuum at 160–450 °C. The results show a clear fatigue endurance “trough” in LOC LBE, while no such a strong temperature dependence of the fatigue endurance is observed when the steel is tested in vacuum. The fractographic observations by means of scanning electron microscopy (SEM) show that ductile microdimples are prevalent on the fracture surfaces of the specimens tested in vacuum, whereas the fracture surfaces produced in LOC LBE at all the temperatures are characterized by quasi-cleavage. Interestingly, using electron backscatter diffraction (EBSD), martensitic laths close to the fatigue crack walls or to the fracture surfaces of the specimens tested in vacuum are found to have transformed into very fine equiaxed subgrains. Nevertheless, such microstructural modifications do not happen to the specimens tested in LOC LBE at 160–450 °C. These interesting microstructural distinctions indicate that liquid metal embrittlement (LME) is able to occur throughout the fatigue crack propagation phase in the full range of the temperatures investigated, i.e. LME is not very sensitive to temperature during the fatigue crack propagation.

  2. Temperature dependence of liquid metal embrittlement susceptibility of a modified 9Cr-1Mo steel under low cycle fatigue in lead-bismuth eutectic at 160-450 °C

    Science.gov (United States)

    Gong, Xing; Marmy, Pierre; Qin, Ling; Verlinden, Bert; Wevers, Martine; Seefeldt, Marc

    2016-01-01

    Low cycle fatigue properties of a 9Cr-1Mo ferritic-martensitic steel (T91) have been tested in a low oxygen concentration (LOC) lead-bismuth eutectic (LBE) environment and in vacuum at 160-450 °C. The results show a clear fatigue endurance "trough" in LOC LBE, while no such a strong temperature dependence of the fatigue endurance is observed when the steel is tested in vacuum. The fractographic observations by means of scanning electron microscopy (SEM) show that ductile microdimples are prevalent on the fracture surfaces of the specimens tested in vacuum, whereas the fracture surfaces produced in LOC LBE at all the temperatures are characterized by quasi-cleavage. Interestingly, using electron backscatter diffraction (EBSD), martensitic laths close to the fatigue crack walls or to the fracture surfaces of the specimens tested in vacuum are found to have transformed into very fine equiaxed subgrains. Nevertheless, such microstructural modifications do not happen to the specimens tested in LOC LBE at 160-450 °C. These interesting microstructural distinctions indicate that liquid metal embrittlement (LME) is able to occur throughout the fatigue crack propagation phase in the full range of the temperatures investigated, i.e. LME is not very sensitive to temperature during the fatigue crack propagation.

  3. Study on the Hot Extrusion Process of Advanced Radiation Resistant Oxide Dispersion Strengthened Steel Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byoungkwon; Noh, Sanghoon; Kim, Kibaik; Kang, Suk Hoon; Chun, Youngbum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Ferritic/martensitic steel has a better thermal conductivity and swelling resistance than austenitic stainless steel. Unfortunately, the available temperature range of ferritic/martensitic steel is limited at up to 650 .deg. C. Oxide dispersion strengthened (ODS) steels have been developed as the most prospective core structural materials for next generation nuclear systems because of their excellent high strength and irradiation resistance. The material performances of this new alloy are attributed to the existence of uniformly distributed nano-oxide particles with a high density, which is extremely stable at high temperature in a ferritic/martensitic matrix. This microstructure can be very attractive in achieving superior mechanical properties at high temperatures, and thus, these favorable microstructures should be obtained through the controls of the fabrication process parameters during the mechanical alloying and hot consolidation procedures. In this study, a hot extrusion process for advanced radiation resistant ODS steel tube was investigated. ODS martensitic steel was designed to have high homogeneity, productivity, and reproducibility. Mechanical alloying and hot consolidation processes were employed to fabricate the ODS steels. A microstructure observation and creep rupture test were examined to investigate the effects of the optimized fabrication conditions. Advanced radiation resistant ODS steel has been designed to have homogeneity, productivity, and reproducibility. For these characteristics, modified mechanical alloying and hot consolidation processes were developed. Microstructure observation revealed that the ODS steel has uniformly distributed fine-grain nano-oxide particles. The fabrication process for the tubing is also being propelled in earnest.

  4. Study on the Hot Extrusion Process of Advanced Radiation Resistant Oxide Dispersion Strengthened Steel Tubes

    International Nuclear Information System (INIS)

    Choi, Byoungkwon; Noh, Sanghoon; Kim, Kibaik; Kang, Suk Hoon; Chun, Youngbum; Kim, Tae Kyu

    2014-01-01

    Ferritic/martensitic steel has a better thermal conductivity and swelling resistance than austenitic stainless steel. Unfortunately, the available temperature range of ferritic/martensitic steel is limited at up to 650 .deg. C. Oxide dispersion strengthened (ODS) steels have been developed as the most prospective core structural materials for next generation nuclear systems because of their excellent high strength and irradiation resistance. The material performances of this new alloy are attributed to the existence of uniformly distributed nano-oxide particles with a high density, which is extremely stable at high temperature in a ferritic/martensitic matrix. This microstructure can be very attractive in achieving superior mechanical properties at high temperatures, and thus, these favorable microstructures should be obtained through the controls of the fabrication process parameters during the mechanical alloying and hot consolidation procedures. In this study, a hot extrusion process for advanced radiation resistant ODS steel tube was investigated. ODS martensitic steel was designed to have high homogeneity, productivity, and reproducibility. Mechanical alloying and hot consolidation processes were employed to fabricate the ODS steels. A microstructure observation and creep rupture test were examined to investigate the effects of the optimized fabrication conditions. Advanced radiation resistant ODS steel has been designed to have homogeneity, productivity, and reproducibility. For these characteristics, modified mechanical alloying and hot consolidation processes were developed. Microstructure observation revealed that the ODS steel has uniformly distributed fine-grain nano-oxide particles. The fabrication process for the tubing is also being propelled in earnest

  5. Substitution of modified 9 Cr-1 Mo steel for austentic stainless steels

    International Nuclear Information System (INIS)

    Sikka, V.K.

    1982-04-01

    This report describes the current program to develop a high-strength ferritic-martensitic steel. The alloy is essentially Fe-9% Cr-1% Mo with small additions of V and Nb and is known as modifed 9 Cr-1 Mo steel. Its elevated-temperature properties and design allowable stresses match those of type 304 stainless steel for temperatures up to 600 0 C and exceed those of other ferritic steels by factors of 2 to 3. The improved strength of this alloy permits its use in place of stainless steels for many applications

  6. Metallurgical Characterization of Reduced Activation Martensitic Steel F-82H Modified

    International Nuclear Information System (INIS)

    Fernandez, P.; Lapena, J.; Lancha, A.M.; Gomez-Briceno, D.; Schirra, M.

    1999-12-01

    During 1995-1998 within of research and development programs on reduced ferritic/martensitic steels for fusion, metallurgical characterization of 8Cr-2WVTa steel, denominated F-28H modified, have been carried out. The work has focused on studying the microstructural and mechanical (tensile, creep, low cycle fatigue and charpy) characteristics of as-received state and aged material in the temperature range 300 degree centigrade to 600 degree centigrade for periods up to 5000 h. (Author) 45 refs

  7. Investigation of the relationships between mechanical properties and microstructure in a Fe-9%Cr ODS steel

    OpenAIRE

    Hary Benjamin; Guilbert Thomas; Wident Pierre; Baudin Thierry; Logé Roland; de Carlan Yann

    2016-01-01

    Ferritic-martensitic Oxide Dispersion Strengthened (ODS) steels are potential materials for fuel pin cladding in Sodium Fast Reactor (SFR) and their optimisation is essential for future industrial applications. In this paper, a feasibility study concerning the generation of tensile specimens using a quenching dilatometer is presented. The ODS steel investigated contains 9%Cr and exhibits a phase transformation between ferrite and austenite around 870 °C. The purpose was to generate different ...

  8. A comparative study of the mechanical properties and the behavior of carbon and boron in stainless steel cladding tubes fabricated by PM HIP and traditional technologies

    Science.gov (United States)

    Shulga, A. V.

    2013-03-01

    The ring tensile test method was optimized and successfully used to obtain precise data for specimens of the cladding tubes of AISI type 316 austenitic stainless steels and ferritic-martensitic stainless steel. The positive modifications in the tensile properties of the stainless steel cladding tubes fabricated by powder metallurgy and hot isostatic pressing of melt atomized powders (PM HIP) when compared with the cladding tubes produced by traditional technology were found. Presently, PM HIP is also used in the fabrication of oxide dispersion strengthened (ODS) ferritic-martensitic steels. The high degree of homogeneity of the distribution of carbon and boron as well the high dispersivity of the phase-structure elements in the specimens manufactured via PM HIP were determined by direct autoradiography methods. These results correlate well with the increase of the tensile properties of the specimens produced by PM HIP technology.

  9. Steels from materials science to structural engineering

    CERN Document Server

    Sha, Wei

    2013-01-01

    Steels and computer-based modelling are fast growing fields in materials science as well as structural engineering, demonstrated by the large amount of recent literature. Steels: From Materials Science to Structural Engineering combines steels research and model development, including the application of modelling techniques in steels.  The latest research includes structural engineering modelling, and novel, prototype alloy steels such as heat-resistant steel, nitride-strengthened ferritic/martensitic steel and low nickel maraging steel.  Researchers studying steels will find the topics vital to their work.  Materials experts will be able to learn about steels used in structural engineering as well as modelling and apply this increasingly important technique in their steel materials research and development. 

  10. Fatigue cracks in Eurofer 97 steel: Part II. Comparison of small and long fatigue crack growth

    Czech Academy of Sciences Publication Activity Database

    Kruml, Tomáš; Hutař, Pavel; Náhlík, Luboš; Seitl, Stanislav; Polák, Jaroslav

    2011-01-01

    Roč. 412, 1 (2011), s. 7-12 ISSN 0022-3115 R&D Projects: GA ČR GA106/09/1954; GA ČR GA101/09/0867 Institutional research plan: CEZ:AV0Z20410507 Keywords : ferritic-martensitic steel * long crack growth * small crack growth * crack closure Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 2.052, year: 2011

  11. Fatigue cracks in Eurofer 97 steel: Part I. Nucleation and small crack growth kinetics

    Czech Academy of Sciences Publication Activity Database

    Kruml, Tomáš; Polák, Jaroslav

    2011-01-01

    Roč. 412, 1 (2011), s. 2-6 ISSN 0022-3115 R&D Projects: GA ČR GA106/09/1954; GA ČR GA101/09/0867 Institutional research plan: CEZ:AV0Z20410507 Keywords : ferritic-martensitic steel * low cycle fatigue * small crack growth * fatigue life prediction Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 2.052, year: 2011

  12. Electroplating of erbium on steel surface in ErCl3 doped LiCl-KCl

    International Nuclear Information System (INIS)

    Kondo, Masatoshi; Tanaka, Teruya; Muroga, Takeo; Tsujimura, Hiroyuki; Ito, Yasuhiko

    2012-01-01

    The electroplating of Er metal on the reduced activation ferritic martensitic steel, JLF-1 (Fe-9Cr-2W-0.1C), in a molten salt was studied. The specimen was immersed in the molten ErCl 3 doped LiCl-KCl electrolyte. The electroplating was carried out by a constant potential electrolysis method and a pulsed current electrolysis method. It was found that the Er metal was deposited on the specimen surface due to the electrochemical reaction. (author)

  13. The effect of microstructural stability on long-term creep behaviour of 11 %Cr steels for steam power plants with operating steam temperatures up to 650 C

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y.; Scholz, A.; Berger, C. [Technische Univ. Darmstadt (DE). Inst. fuer Werkstoffkunde (IFW); Kauffmann, F.; Maile, K. [Stuttgart Univ. (DE). Materialpruefungsanstalt (MPA); Mayer, K.H. [Alstom Power, Nuernberg (Germany)

    2010-07-01

    The investigations of advanced ferritic/martensitic steels for 650 C power plant components focus on the improvement of high-temperature creep properties with respect to chemical composition. This study deals with the development of new heat resistant 11-12%Cr ferritic-martensitic steels with sufficient creep and oxidation resistance up to 650 application by using basic principles and concepts of physical metallurgy. The highest creep strength could be achieved with a 0.04% Nb alloyed 11%CrWCoMoVB melt, which is in addition alloyed with a higher C and B content as well as with lower W and Co portions. The microstructure evolution during creep of this newly developed steel was investigated in comparison to a sister alloy which comprises 0.06% Ta instead of the Nb. (orig.)

  14. Ductility of high chromium stainless steels

    International Nuclear Information System (INIS)

    Peretyat'ko, V.N.; Kazantsev, A.A.

    1997-01-01

    Aimed to optimize the hot working conditions for high chromium stainless steels the experiments were carried in the temperature range of 800-1300 deg C using hot torsion tests and cylindrical specimens of ferritic and ferritic-martensitic steels 08Kh13, 12Kh13, 20Kh13, 30Kh13 and 40Kh13. Testing results showed that steel plasticity varies in a wide range depending on carbon content. Steels of lesser carbon concentration (08Kh13 and 12Kh13) exhibit a sharp increase in plasticity with a temperature rise, especially in the interval of 1200-1250 deg C. Steels 20Kh13 and 30Kh13 display insignificant plasticity increasing, whereas plastic properties of steel 40Kh13 increase noticeably in the range of 1000-1300 deg C. It is shown that optimal hot working conditions for specific steel must be selected with account of steel phase composition at high temperatures

  15. Effects of alloying and processing modifications on precipitation and strength in 9%Cr ferritic/martensitic steels for fast reactor cladding

    Science.gov (United States)

    Tippey, Kristin E.

    P92 was modified with respect to alloying and processing in the attempt to enhance high-temperature microstructural stability and mechanical properties. Alloying effects were modeled in ThermoCalcRTM and analyzed with reference to literature. ThermoCalcRTM modeling was conducted to design two low-carbon P92-like low-carbon alloys with austenite stabilized by alternative alloying; full conversion to austenite allows for a fully martensitic structure. Goals included avoidance of Z-phase, decrease of M23C6 phase fraction and maintained or increased MX phase fraction. Fine carbonitride precipitation was optimized by selecting alloying compositions such that all V and Nb could be solutionized at temperatures outside the delta-ferrite phase field. A low-carbon alloy (LC) and a low-carbon-zero-niobium alloy (0Nb) were identified and fabricated. This low-carbon approach stems from the increased creep resistance reported in several low-carbon alloys, presumably from reduced M23C6 precipitation and maintained MX precipitation [1], although these low-carbon alloys also contained additional tungsten (W) and cobalt (Co) compared to the base P92 alloy. The synergistic effect of Co and W on the microstructure and mechanical properties are difficult to deconvolute. Higher solutionizing temperatures allow more V and Nb into solution and increase prior austenite grain size; however, at sufficiently high temperatures delta-ferrite forms. Optimal solutionizing temperatures to maximize V and Nb in solution, while avoiding the onset of the delta ferrite phase field, were analyzed in ThermoCalcRTM. Optical microscopy showed ThermoCalc RTM predicted higher delta-ferrite onset temperatures of 20 °C in P92 alloys to nearly 50 °C in the designed alloys of the critical temperature. Identifying the balance where maximum fine precipitation is achieved and delta-ferrite avoided is a key factor in the design of an acceptable P92-like alloy for Generation IV reactor cladding. Processing was further modified utilizing thermomechanical processing (TMP) simulations with the GleebleRTM 3500. Hardness increased substantially in thermomechanically processed alloys, with increased hardness strongly correlating to decreased TMP temperature. The most significant difference between low- and high-temperature thermomechanically processed specimens was an increase in crystallite size at the higher temperature. The fundamental reason for higher strength in the TMP conditions is higher dislocation density, as precipitate volume fraction was not specifically improved in TMP conditions. Thermal stability of the base P92 and of the experimental alloys was analyzed by aging the alloys for times ranging from 500 to 10,000 h at 550, 600, 650, and 700 °C. Results suggest the hardness and thermal stability of LC is greater than that of 0Nb at lower aging temperatures and shorter times, with 0Nb surpassing LC microhardness at 10,000 h at 650 °C and for most conditions aged at 700 °C. Small- and wide-angle x-ray scattering (SAXS/WAXS) was conducted at Argonne National Laboratory (ANL). Atom probe tomography (APT) and scanning transmission electron microscopy high-angle annular dark field (STEM-HAADF) in conjunction with EDS were used to elucidate x-ray findings. These microstructural characteristics were then correlated with mechanical properties, including Vickers microhardness testing, elevated-temperature tensile testing, and creep rupture testing. The designed alloys exhibited less stable microstructures leading to less favorable mechanical potencies, as compared to the base P92 alloy. It is posited that factors other than inclination towards MX over M23C6 precipitation are important in generating thermal stability and high-temperature strength, i.e. perhaps the solid solution or diffusion controlling effects of Co in the low-carbon variation of Taneike's alloys [1] delay martensite recovery. The refined thermal profiles, however, put both P92 and LC creep strengths beyond those found in literature.

  16. Corrosion behavior of steels in liquid lead bismuth with low oxygen concentrations

    Science.gov (United States)

    Kurata, Yuji; Futakawa, Masatoshi; Saito, Shigeru

    2008-02-01

    Corrosion tests in pots were conducted to elucidate corrosion behavior of various steels in liquid lead-bismuth for 3000 h under the condition of an oxygen concentration of 5 × 10 -8 wt% at 450 °C and an oxygen concentration of 3 × 10 -9 wt% at 550 °C, respectively. Significant corrosion was not observed at 450 °C for ferritic/martensitic steels, F82H, Mod.9Cr-1Mo steel, 410SS, 430SS except 2.25Cr-1Mo steel. Pb-Bi penetration into steels and dissolution of elements into Pb-Bi were severe at 550 °C even for ferritic/martensitic steels. Typical dissolution attack occurred for pure iron both at 550 °C without surface Fe 3O 4 and at 450 °C with a thin Fe 3O 4 film. Ferritization due to dissolution of Ni and Cr, and Pb-Bi penetration were recognized for austenitic stainless steels, 316SS and 14Cr-16Ni-2Mo steel at both temperatures of 450 °C and 550 °C. The phenomena were mitigated for 18Cr-20Ni-5Si steel. In some cases oxide films could not be a corrosion barrier in liquid lead-bismuth.

  17. FATIGUE BEHAVIOR OF HOT-ROLLED STEEL INTENDED FOR COLD FORMING

    Directory of Open Access Journals (Sweden)

    Gejza Rosenberg

    2011-07-01

    Full Text Available In the work, there are presented measured tension and fatigue properties of eight low-carbon steels moulded in form of 20 kg ingots that were processed by controlled regime of rolling /cooling and then exposed to simulated effect of two coiling temperatures. The experimental results presented in the work show, that steels with ferrite-martensite or ferrite-bainitic microstructure have in comparison to ferrite-pearlitic or ferrite-carbidic microstructure better strength-plastic properties, but worse resistance to cyclic loading.

  18. Evaluation of factors affecting the edge formability of two hot rolled multiphase steels

    Science.gov (United States)

    Mukherjee, Monideepa; Tiwari, Sumit; Bhattacharya, Basudev

    2018-02-01

    In this study, the effect of various factors on the hole expansion ratio and hence on the edge formability of two hot rolled multiphase steels, one with a ferrite-martensite microstructure and the other with a ferrite-bainite microstructure, was investigated through systematic microstructural and mechanical characterization. The study revealed that the microstructure of the steels, which determines their strain hardening capacity and fracture resistance, is the principal factor controlling edge formability. The influence of other factors such as tensile strength, ductility, anisotropy, and thickness, though present, are secondary. A critical evaluation of the available empirical models for hole expansion ratio prediction is also presented.

  19. Thermoelastoviscoplastic modeling of RAFM steel JLF-1 using tensile and low cycle fatigue experiments

    Energy Technology Data Exchange (ETDEWEB)

    Msolli, S., E-mail: sabeur.msolli@univ-lorraine.fr

    2014-08-01

    In this paper, a modeling of the elastoviscoplastic behavior of a Reduced Activation Ferritic Martensitic (RAFM) steel JLF-1 is presented. The modeling of this material was based on various Low Cycle Fatigue (LCF) and tensile tests performed in air and vacuum using different imposed strain rates and temperature ranges going from ambient temperature to 873 K. The coupled viscoplastic model is coded in FORTRAN program, implemented into the finite elements code ABAQUS and used to predict the thermomechanical behavior of a fatigue specimen made of RAFM steel JLF-1. Good agreements were found between numerical results and experimental data.

  20. Design of duplex low carbon steels for improved strength: weight applications

    Energy Technology Data Exchange (ETDEWEB)

    Koo, J.

    1977-08-01

    Duplex ferrite-martensite (DFM) steels with excellent combinations of strength and formability have been selected from first principles, and their microstructure-property relationships established through a systematic investigation of a series of ternary Fe/X/0.1 C steels (X being varying amount of Cr and Si). Duplex processing consists of initial austenitization and quenching to form 100% martensite, followed by annealing in the (..cap alpha.. + ..gamma..) range and subsequent quenching. The resultant DFM morphology is controlled by the type and amount of alloying element X.

  1. Metallurgical comparison between the experimental ED-ODS and commercial ODS steels

    International Nuclear Information System (INIS)

    Fernandez, P.; Serrano, M.; Lapena, J.

    2007-01-01

    Full text of publication follows: Recently, reduced activation oxide dispersion strengthened ferritic/martensitic steels have been identified as potential structural material candidates for first wall and blanket structures of fusion devices. These steels allow to increase the operation temperature of the fusion reactor, around of 100 deg. C or even more, because of their excellent thermal creep resistance. ODS steels are being developed and investigated for nuclear fission and fusion applications in Japan, Europe and the United States. Commercial ODS products, such as MA956 and PM2000 are available and are being used for high temperature applications. Since no reduced activation ferritic/martensitic steels are commercially available at present, the European Materials Fusion Programme has considered in the last years initiate adequate research activities to produce and characterize these materials to evaluate the feasibility of their use in the different blanket designs. Nowadays, the ODS steel that are being extensively investigated in Europe is the denominated EU-ODS. This alloy has the basic composition of the Eurofer'97 with 0.3% of Y 2 O 3 particles. In this paper, the metallurgical properties (microstructural, Charpy, etc) of the newly developed EU-ODS steel are presented and discussed together with the properties of the commercial MA956 and PM2000 ODS steels, also studied in this work, in order to show the differences and similarities between these ODS alloys. (authors)

  2. Evolution of stainless steels in nuclear industry

    International Nuclear Information System (INIS)

    Tavassoli, Farhad

    2010-01-01

    Starting with the stainless steels used in the conventional industry, their adoption and successive evolutions in the nuclear industry, from one generation of nuclear reactors to another, is presented. Specific examples for several steels are given, covering fabrication procedures, qualification methods, property databases and design allowable stresses, to show how the ever-increasing demands for better performance and reliability, in particular under neutron irradiation, have been met. Particular attention is paid to the austenitic stainless steels types 304L, 316L, 316L(N), 316L(N)-IG, titanium stabilized grade 321, precipitation strengthened alloy 800, conventional and low activation ferritic/martensitic steels and their oxygen dispersion strengthening (ODS) derivatives. For each material, the evolution of the associated filler metal and welding techniques are also presented. (author)

  3. A Critical Analysis of Grain-Size and Yield-Strength Dependence of Near-Threshold Fatigue-Crack Growth in Steels.

    Science.gov (United States)

    1981-07-15

    of high-strength steel ), and a host of microstructural types (ferritic, martensitic, pearlitic, bainitic , austenitic). Accession For NTIS GRA&l DTIC...IN STEELS , : RPRNGO.RPRTNMR 1~A Tw.R CONTRACT OA4A&XMUt8~ G/~ ! R./koderl L.A./Cooleyad T.W./Crooker 2 .{I 9PERFORMING ORGANIZATION NAME AND A10R4SI...growth Steels Microstructure Ferrous alloys Structure-sensitive crack growth 20 ABSTRACT (Con~tinue an r*,er.. side it necesar and Identity by black

  4. Microstructure refinement and strengthening mechanisms of a 9Cr oxide dispersion strengthened steel by zirconium addition

    International Nuclear Information System (INIS)

    Xu, Hai Jian; Lu, Zheng; Wang, Dong Mei; Liu, Chunming

    2017-01-01

    To study the effects of zirconium (Zr) addition on the microstructure, hardness and the tensile properties of oxide dispersion strengthened (ODS) ferritic-martensitic steels, two kinds of 9Cr-ODS ferritic-martensitic steels with nominal compositions (wt.%) of Fe-9Cr-2W-0.3Y_2O-3 and Fe-9Cr-2W-0.3Zr-0.3Y_2O_3 were fabricated by the mechanical alloying (MA) of premixed powders and then consolidated by hot isostatic pressing (HIP) techniques. The experimental results showed that the average grain size decreases with Zr addition. The trigonal δ-phase Y_4Zr_3O_1_2 oxides and body-centered cubic Y_2O_3 oxides are formed in the 9Cr-Zr-ODS steel and 9Cr non-Zr ODS steel, respectively, and the average size of Y_4Zr_3O_1_2 particles is much smaller than that of Y_2O_3. The dispersion morphology of the oxide particles in 9Cr-Zr-ODS steel is significantly improved and the number density is 1.1 x 10"2"3/m"3 with Zr addition. The 9Cr-Zr-ODS steel shows much higher tensile ductility, ultimate tensile strength and Vickers hardness at the same time

  5. Microstructure refinement and strengthening mechanisms of a 9Cr oxide dispersion strengthened steel by zirconium addition

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Hai Jian; Lu, Zheng; Wang, Dong Mei; Liu, Chunming [Key Laboratory for Anisotropy and Texture of Materials (Ministry of Education), School of Materials Science and Engineering, Northeastern University, Shenyang (China)

    2017-02-15

    To study the effects of zirconium (Zr) addition on the microstructure, hardness and the tensile properties of oxide dispersion strengthened (ODS) ferritic-martensitic steels, two kinds of 9Cr-ODS ferritic-martensitic steels with nominal compositions (wt.%) of Fe-9Cr-2W-0.3Y{sub 2}O-3 and Fe-9Cr-2W-0.3Zr-0.3Y{sub 2}O{sub 3} were fabricated by the mechanical alloying (MA) of premixed powders and then consolidated by hot isostatic pressing (HIP) techniques. The experimental results showed that the average grain size decreases with Zr addition. The trigonal δ-phase Y{sub 4}Zr{sub 3}O{sub 12} oxides and body-centered cubic Y{sub 2}O{sub 3} oxides are formed in the 9Cr-Zr-ODS steel and 9Cr non-Zr ODS steel, respectively, and the average size of Y{sub 4}Zr{sub 3}O{sub 12} particles is much smaller than that of Y{sub 2}O{sub 3}. The dispersion morphology of the oxide particles in 9Cr-Zr-ODS steel is significantly improved and the number density is 1.1 x 10{sup 23}/m{sup 3} with Zr addition. The 9Cr-Zr-ODS steel shows much higher tensile ductility, ultimate tensile strength and Vickers hardness at the same time.

  6. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

  7. Estimation of flow stress of radiation induced F/M steels using molecular dynamics and discrete dislocation dynamics approach

    International Nuclear Information System (INIS)

    More, Ameya; Dutta, B.K.; Durgaprasad, P.V.; Arya, A.K.

    2012-01-01

    Fe-Cr based Ferritic/Martensitic (F/M) steels are the candidate structural materials for future fusion reactors. In this work, a multi-scale approach comprising atomistic Molecular Dynamics (MD) simulations and Discrete Dislocation Dynamics (DDD) simulations are used to model the effect of irradiation dose on the flow stress of F/M steels. At the atomic scale, molecular dynamics simulations are used to study the dislocation interaction with irradiation induced defects, i.e. voids and He bubbles. Whereas, the DDD simulations are used to estimate the change in flow stress of the material as a result of irradiation hardening. (author)

  8. New ferritic steels for advanced steam plants

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, K.H; Koenig, H. [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

    1998-12-31

    During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

  9. New ferritic steels for advanced steam plants

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, K H; Koenig, H [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

    1999-12-31

    During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

  10. Development of ferritic steels for reduced activation: the US program

    International Nuclear Information System (INIS)

    Klueh, R.L.; Gelles, D.S.; Lechtenberg, T.A.

    1986-01-01

    The Cr-Mo ferritic (martensitic) steels are candidates for the structural components of fusion reactors. Irradiation of such steels in a fusion environment produces long-lived radioactive isotopes, which lead to difficult radioactive waste disposal problems once the structure is removed from service. Such problems could be reduced by using steels that contain only elements that produce radioactive isotopes that decay to low levels in a reasonable time (tens of years instead of hundreds or thousands of years). The US Department of Energy has a program to develop steels to meet the criteria for shallow land burial as opposed to deep geologic storage. A review of the alloy development programs indicates that ferritic steels that meet these criteria can be developed

  11. Fatigue crack growth behavior of RAFM steel in Paris and threshold regimes at different temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Babu, M. Nani; Sasikala, G., E-mail: gsasi@igcar.gov.in; Dutt, B. Shashank; Venugopal, S.; Bhaduri, A.K.; Jayakumar, T.

    2014-04-01

    Fatigue crack growth (FCG) behavior of a reduced activation ferritic martensitic (indigenous RAFM) steel has been evaluated at 300, 653 and 823 K in Paris and threshold regimes. The effect of temperature on threshold stress intensity factor range and associated crack closure mechanisms is highlighted. The FCG results were compared with those for EUROFER 97. Further, crack tip effective stress intensity factor ranges (ΔK{sub tip,eff}) have been evaluated by taking crack tip shielding into account in order to examine the effect of temperature on true intrinsic FCG behavior.

  12. Presentation of the results for deuterium retention and thermal release in a new type of low activation ferritic-martensitic steel EUROFER / Результаты исследования по удержанию дейтерия и термической десорбции в условиях низкой активации ферритно-мартенситной стали EUROFER / Rezultati zadržavanja i termalne desorpcije deuterijuma u EUROFER-u, novoj vrsti feritno-martenzitnog čelika niske aktivacije

    Directory of Open Access Journals (Sweden)

    Sanja Lj. Korica

    2016-04-01

    Full Text Available This work presents the results of the investigation into deuterium retention and thermal release in low activation ferritic-martensitic steel EUROFER and its main components, Fe and Cr, using ion implantation, nuclear reaction depth profiling and the thermal desorption spectrometry technique. The samples were exposed to deuterium ECR plasma of 6.5*1024 Dm-2 fluence. Our NRA results show that, at 300 K, most of deuterium is trapped in the near-surface region. A part of the implanted deuterium diffuses into the bulk and is trapped by defects beyond the implantation range. The results of the TDS measurement of the Cr sample confirm the formation of an ordered Cr-hydride phase. / В данной статье приведены результаты исследований по удержанию и термической десорбции дейтерия в железе, хроме и сплаве EOROFER – представляющем новейший материал для фузионного (термоядерного реактора. Исследования показали следующие результаты: *Удержание дейтерия в хроме намного выше, чем в железе (из-за образования гибридов хрома *Удержание дейтерия в сплаве EUROFER выше на фактор 2, чем в железе *Выявлена специфиеская структура в концентрационном профиле железа и сплава EUROFER на глубине ~ 4 μm *Высокий уровень диффузии и удержания дейтерия свидетельствует о потенциальном применении Au в качестве диффузионного барьера в фузионном (термоядерном реакторе. / U ovom radu ispitivano je zadržavanje i termalna desorpcija deuterijuma u gvo

  13. Dislocation density evolution in the process of high-temperature treatment and creep of EK-181 steel

    Energy Technology Data Exchange (ETDEWEB)

    Vershinina, Tatyana, E-mail: vershinina@bsu.edu.ru [Belgorod State National Research University, Pobedy street 85, Belgorod 308015 (Russian Federation); Leont' eva-Smirnova, Maria, E-mail: smirnova@bochvar.ru [Bochvar High-Technology Research Institute of Inorganic Materials, ul. Rogova 5, Moscow 123098 (Russian Federation)

    2017-03-15

    X-ray diffraction has been used to study the dislocation structure in ferrite-martensite high-chromium steel EK-181 in the states after heat treatment and high-temperature creep. The influence of heat treatment and stress on evolution of lath martensite structure was investigated by and electron back-scattered diffraction. The effect of nitrogen content on the total dislocation density, fraction of edge and screw dislocation segments are analyzed. - Highlights: •Fraction of edge dislocation in quenched state depends on nitrogen concentration. •Nitrogen affects the character of dislocation structure evolution during annealing. •Edge dislocations fraction influences on dislocation density after aging and creep.

  14. High-throughput design of low-activation, high-strength creep-resistant steels for nuclear-reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Qi; Zwaag, Sybrand van der [Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS, Delft (Netherlands); Xu, Wei, E-mail: xuwei@ral.neu.edu.cn [State Key Laboratory of Rolling and Automation, Northeastern University, 110819, Shenyang (China); Novel Aerospace Materials Group, Faculty of Aerospace Engineering, Delft University of Technology, Kluyverweg 1, 2629 HS, Delft (Netherlands)

    2016-02-15

    Reduced-activation ferritic/martensitic steels are prime candidate materials for structural applications in nuclear power reactors. However, their creep strength is much lower than that of creep-resistant steel developed for conventional fossil-fired power plants as alloying elements with a high neutron activation cannot be used. To improve the creep strength and to maintain a low activation, a high-throughput computational alloy design model coupling thermodynamics, precipitate-coarsening kinetics and an optimization genetic algorithm, is developed. Twelve relevant alloying elements with either low or high activation are considered simultaneously. The activity levels at 0–10 year after the end of irradiation are taken as optimization parameter. The creep-strength values (after exposure for 10 years at 650 °C) are estimated on the basis of the solid-solution strengthening and the precipitation hardening (taking into account precipitate coarsening). Potential alloy compositions leading to a high austenite fraction or a high percentage of undesirable second phase particles are rejected automatically in the optimization cycle. The newly identified alloys have a much higher precipitation hardening and solid-solution strengthening at the same activity level as existing reduced-activation ferritic/martensitic steels.

  15. Development of novel protective high temperature coatings on heat exchanger steels and their corrosion resistance in simulated coal firing environment; Developpement de revetements pour les aciers d'echangeurs thermiques et amelioration de leur resistance a la corrosion en environnement simulant les fumees de combustion et de charbon

    Energy Technology Data Exchange (ETDEWEB)

    Rohr, V.

    2005-10-15

    Improving the efficiencies of thermal power plants requires an increase of the operating temperatures and thus of the corrosion resistance of heat exchanger materials. Therefore, the present study aimed at developing protective coatings using the pack cementation process. Two types of heat exchanger steels were investigated: a 17% Cr-13% Ni austenitic steel and three ferritic-martensitic steels with 9 (P91 and P92) and 12% Cr (HCM12A). The austenitic steel was successfully aluminized at 950 C. For the ferritic-martensitic steels, the pack cementation temperature was decreased down to 650 C, in order to maintain their initial microstructure. Two types of aluminides, made of Fe{sub 2}Al{sub 5} and FeAl, were developed. A mechanism of the coating formation at low temperature is proposed. Furthermore, combining the pack cementation with the conventional heat treatment of P91 allowed to take benefit of higher temperatures for the deposition of a two-step Cr+Al coating. The corrosion resistance of coated and uncoated steels is compared in simulated coal firing environment for durations up to 2000 h between 650 and 700 C. It is shown that the coatings offer a significant corrosion protection and, thus, an increase of the component lifetime. Finally, the performance of coated 9-12% Cr steels is no longer limited by corrosion but by interdiffusion between the coating and the substrate. (author)

  16. Electron microscopy study of microstructure of the oxide-dispersion-strengthed steel

    International Nuclear Information System (INIS)

    Xing, H.; Sun, J.; Zhou, Z.J.

    2010-01-01

    The microstructure of the ODS ferritic-martensitic steel with chemical composition of Fe-12Cr-2W-0.5Ti-0.2V-0.2Si-0.13C-0.35Y 2 O 3 wt% fabricated by MA and HIP has been investigated by TEM. The emphasis is focused on the structure and chemical composition of the fine ODS particles and inclusions. The results showed that two types of complex ODS particles such as Y-Ti-O and Y-Si-O with nanometer size distribute homogeneously and incoherently in the matrix of the steel. Additionally, large (Ti,V)N inclusions were observed in the steel. The results of microstructural characterization are discussed to correlate with the processing and mechanical properties of the ODS steel. (author)

  17. A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

    Science.gov (United States)

    Şahin, Sümer; Übeyli, Mustafa

    2008-12-01

    Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can't withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.

  18. Effects of irradiation conditions and environment on the reflectivity of different steel mirrors for ITER diagnostics systems

    International Nuclear Information System (INIS)

    Hernandez, Teresa; Martin, Piedad; Fernandez, Pilar; Hodgson, Eric R.

    2009-01-01

    In this work possible degradation of the reflectivity for mirrors made from various steels subjected to ionizing radiation, at moderate temperature and in different environments (vacuum, air, or nitrogen) up to a total dose of 9 MGy, has been examined. Mirrors were prepared from conventional austenitic stainless steel (316L) and also reduced activation ferritic/martensitic (RAFM) steels (Eurofer, ODS-Eurofer, F82H), and the reflectivity studied from ultraviolet to near infrared, before and after different treatments. Under all conditions the austenitic steel mirrors only degrade slightly (<10%), however for the reduced activation steels important reflectivity degradation for wavelengths below 1000 nm are observed for the different conditions. Surface morphology and microstructure has been also investigated using scanning electron microscopy (SEM). The production of near surface nitrides is considered to be the possible cause of the optical degradation.

  19. R and D status of China low activation martensitic steel

    International Nuclear Information System (INIS)

    Huang Qunying; Li Chunjing; Li Yanfen; Liu Shaojun; Wu Yican; Li Jiangang; Shan Yiyin; Yu Jinnan; Zhu Shengyun; Zhang Pinyuan; Yang Jianfeng; Han Fusheng; Kong Mingguang; Li Heqin; Muroga, T.; Nagasaka, T.

    2007-01-01

    The Reduced Activation Ferritic/Martensitic (RAFM) steel is considered as the primary candidate structural material for DEMO and the first fusion plant, and widely studied in the world. China low activation martensitic steel (CLAM) is being developed in Institute of Plasma Physics, Chinese Academy of Sciences, under wide collaboration with many other domestic and foreign institutes and universities. This paper summarized the main R and D progress on CLAM, which covered composition optimization of the CLAM, smelting and processing techniques, physical and mechanical property test and evaluation before and after irradiation, compatibility with liquid LiPb, welding techniques etc. Finally, further research and development, and the prospects on its application were stated. (authors)

  20. Dynamic mechanical properties of reduced activation ferritic steels

    International Nuclear Information System (INIS)

    Hirose, T.; Kohyama, A.; Tanigawa, H.; Ando, M.; Jitsukawa, S.

    2003-01-01

    A fatigue test method by a miniaturized hourglass-shaped fatigue specimen has been developed for International Fusion Materials Irradiation Facility (IFMIF) and sufficient potential as the alternative to a conventional large specimen was presented. Furthermore, focused ion beam micro- sampling method was successfully applied to microstructural analysis on fracture process. Where, the effects of displacement damage and transmutation helium on the fatigue properties of Reduced Activation Ferritic/Martensitic Steels, RAFs, were investigated. Neutron irradiation and helium-ion-implantation at ambient temperature caused radiation hardening to degrade fatigue lifetime of F82H steel. Microstructural analysis revealed that local brittle fractures occurred at early stage of fatigue tests was the origin of the degradation.. No significant difference in fatigue life degradation was detected with and without implanted helium. This result suggests that 100 appm helium implanted has no impact on fracture life time under neutron irradiation. (author)

  1. Influence of LBE long term exposure and simultaneous fast neutron irradiation on the mechanical properties of T91 and 316L

    Energy Technology Data Exchange (ETDEWEB)

    Stergar, E., E-mail: estergar@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Eremin, S.G. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation); Gavrilov, S.; Lambrecht, M. [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Makarov, O.; Iakovlev, V. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

    2016-05-15

    The LEXUR–II–LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-03-28

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

  3. Compatibility study between U-UO{sub 2} cermet fuel and T91 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kaity, Santu; Khan, K.B. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sengupta, Pranesh; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2016-12-01

    Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO{sub 2} cermet fuel and T91 cladding material. These diffusion couples were annealed at 923–1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope (SEM), X-ray energy dispersive spectroscope (EDS) and electron probe microanalyser (EPMA). It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U{sub 6}(Fe,Cr) and U(Fe,Cr){sub 2} intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.

  4. Aumento en la resistencia a la corrosión de un acero t91 aluminizado

    OpenAIRE

    QUINTERO, LUZ S.; LAVERDE, DIONISIO; ESCALANTE, HUMBERTO

    2011-01-01

    El acero ferrítico T91 ha sido ampliamente utilizado en plantas de generación de energía eléctrica, calderas y plantas nucleares. Sin embargo, su resistencia a la oxidación se ve gravemente afectada a temperaturas superiores a los 600ºC. En esta investigación se empleó la técnica de cementación empacada para aluminizar un acero T91 y se encontraron las condiciones óptimas para la formación de un recubrimiento libre de porosidad y con espesor uniforme (300 micras). Se evaluó la capa de óxido f...

  5. AUMENTO EN LA RESISTENCIA A LA CORROSIÓN DE UN ACERO T91 ALUMINIZADO

    Directory of Open Access Journals (Sweden)

    LUZ S. QUINTERO

    2011-01-01

    Full Text Available El acero ferrítico T91 ha sido ampliamente utilizado en plantas de generación de energía eléctrica, calderas y plantas nucleares. Sin embargo, su resistencia a la oxidación se ve gravemente afectada a temperaturas superiores a los 600ºC. En esta investigación se empleó la técnica de cementación empacada para aluminizar un acero T91 y se encontraron las condiciones óptimas para la formación de un recubrimiento libre de porosidad y con espesor uniforme (300 micras. Se evaluó la capa de óxido formada luego de una exposición a 800ºC en atmósfera de aire. La caracterización del recubrimiento y de la capa de óxido (Al2O3 se llevó a cabo empleando un microscopio electrónico de barrido (SEM, Espectroscopia de Rayos X (EDX y mapping. Se comprobó una buena resistencia a la oxidación del acero T91 aluminizado, debido a la formación de una fina capa continua de óxido de aluminio de aproximadamente 6 micras de espesor luego de una exposición de 7 días.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

    2011-10-01

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

  7. The development of EUROFER reduced activation steel

    Energy Technology Data Exchange (ETDEWEB)

    Schaaf, B. van der E-mail: vanderschaaf@nrg-nl.com; Tavassoli, F.; Fazio, C.; Rigal, E.; Diegele, E.; Lindau, R.; LeMarois, G

    2003-09-01

    Ferritic martensitic steels show limited swelling and susceptibility to helium effects and can be made with low activation chemical compositions. These properties make them the reference steel for the development of breeding blankets in fusion power plants. EUROFER97 is the European implementation of such a steel, where experience gained from an IEA co-operation with Japan and the US is also implemented. Results obtained so far show that EUROFER steel has attractive mechanical properties even after long ageing times. Compatibility tests in water and PbLi17 are in progress. Oxidised aluminium is the most effective protective layer in PbLi17. The displacement damage and helium formation strongly influence the hydrogen transport in the steel. Present experiments should be backed by tests in a more fusion relevant environment, e.g. IFMIF. The 2.5 dpa neutron irradiations at low temperatures result in a higher DBTT. High dose irradiations, up to 80 dpa, are underway. The early results of ODS grades with EUROFER steel composition show potential of these grades for increasing the operating temperature with 100-150 K.

  8. Corrosion Resistance of Steels and Armco-Fe in Lead Melt Saturated by Oxygen at 550 degree C

    International Nuclear Information System (INIS)

    Tsisar, V.P.; Fedirko, V.N.; Eliseeva, O.I.

    2007-01-01

    Corrosion resistance of stainless steels and Armco-Fe in static lead melt saturated by oxygen at 550 degree C for 2000 h was investigated. It was determined that double oxide layer was formed on the surface of investigated materials. Outer part of double oxide growths from the initial interface 'solid metal/liquid lead' towards the melt and consists of Fe 3 O 4 . Inner part of double oxide based on the matrix is composed of Fe 3 O 4 for Armco-Fe, Fe 1+x Cr 2-x O 4 for martensitic 0.2 C-13 Cr and ferritic-martensitic EP823 steels and Fe 1+x Cr 2- xO 4 +Ni for austenitic 18Cr-10Ni-1Ti. Lead did not penetrate into the matrix of tested materials and was detected only in the scale formed on austenitic steel

  9. Studies on oxidation and deuterium permeation behavior of a low temperature α-Al_2O_3-forming Fe−Cr−Al ferritic steel

    International Nuclear Information System (INIS)

    Xu, Yu-Ping; Zhao, Si-Xiang; Liu, Feng; Li, Xiao-Chun; Zhao, Ming-Zhong; Wang, Jing; Lu, Tao; Hong, Suk-Ho; Zhou, Hai-Shan; Luo, Guang-Nan

    2016-01-01

    To evaluate the capability of Fe−Cr−Al ferritic steels as tritium permeation barrier in fusion systems, the oxidation behavior together with the permeation behavior of a Fe−Cr−Al steel was investigated. Gas driven permeation experiments were performed. The permeability of the oxidized Fe−Cr−Al steel was obtained and a reduced activation ferritic/martensitic steel CLF-1 was used as a comparison. In order to characterize the oxide layer, SEM, XPS, TEM, HRTEM were used. Al_2O_3 was detected in the oxide film by XPS, and HRTEM showed that Al_2O_3 in the α phase was found. The formation of α-Al_2O_3 layer at a relatively low temperature may result from the formation of Cr_2O_3 nuclei.

  10. Water corrosion test of oxide dispersion strengthened (ODS) steel claddings

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

    2006-07-01

    As a part of feasibility study of ODS steel cladding, its water corrosion resistance was examined under water pool condition. Although addition of Cr is effective for preventing water corrosion, excessive Cr addition leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, water corrosion test was conducted for these ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) Corrosion rate of 9Cr-ODS martensitic and 12Cr-ODS ferritic steel are significantly small and no pitting was observed. Thus, these ODS steels have superior resistance for water corrosion under the condition of 60degC and pH8-12. (2) It was showed that 9Cr-ODS martensitic steel and 12Cr-ODS ferritic steel have comparable water corrosion resistance to that of PNC316 and PNC-FMS at 60degC for 1,000h under varying pH of 8, 10. Water corrosion resistance of these alloys is slightly larger than that of PNC316 and PNC-FMS at pH12 without significant difference of appearance and uneven condition. (author)

  11. Corrosion resistance improvement of ferritic steels through hydrogen additions to the BWR coolant

    International Nuclear Information System (INIS)

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Indig, M.E.

    1984-01-01

    Motivated by the success of oxygen suppression for mitigation of intergranular stress corrosion cracking (IGSCC) in weld sensitized austenitic materials used in Boiling Water Reactors (BWRs), oxygen suppression, through hydrogen additions to the feedwater was investigated to determine its affect on the corrosion resistance of ferritic and martensitic BWR structural materials. The results of these investigations are presented in this paper, where particular emphasis is placed on the corrosion performance of BWR pressure vessel low alloy steels, carbon steel piping materials and martensitic pump materials. It is important to note that the corrosion resistance of these materials in the BWR environment is excellent. Consequently this investigation was also motivated to determine whether there were any detrimental effects of hydrogen additions, as well as to identify any additional margin in ferritic/martensitic materials corrosion performance

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

    Science.gov (United States)

    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.

  13. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    International Nuclear Information System (INIS)

    Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

    2012-01-01

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different 10 B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  14. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-15

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  15. Mechanical and microstructural stability of P92 steel under uniaxial tension at high temperature

    International Nuclear Information System (INIS)

    Giroux, P.F.; Dalle, F.; Sauzay, M.; Malaplate, J.; Fournier, B.; Gourgues-Lorenzon, A.F.

    2010-01-01

    9-12%Cr creep-resistant ferritic-martensitic steels are candidates for structural components of Generation IV nuclear power plants. However, they are sensitive to softening during low-cycle fatigue, creep and creep-fatigue tests, due to the destabilisation of the tempered martensite microstructure, possibly inducing a decrease in further creep resistance. To better identify the softening mechanisms in P92 steel during uniaxial deformation, tensile tests were carried out at 823 K, showing an extended and stable softening stage on true stress-strain curves after some work-hardening. Three phenomena were studied in order to understand this behaviour: mechanical instability (necking), damage and grain size evolution. Examination of fractured and non-fractured tensile specimens (light optical and electron microscopy, macrohardness) suggested that the physical mechanisms responsible for softening are mainly (sub)grain size evolution and diffuse necking. Models were proposed to predict grain growth and beginning of the mechanical instability during homogeneous deformation.

  16. TEM characterization on new 9% Cr advanced steels thermomechanical treated after tempering

    Science.gov (United States)

    Fernández, P.; Hoffmann, J.; Rieth, M.; Roldán, M.; Gómez-Herrero, A.

    2018-03-01

    Phase transformation on new six reduced activation ferritic/martensitic steels (RAFMs) was investigated to provide the basis for the design and development of advanced steels to maintain adequate strength and creep resistance above 500 °C. The new alloys are designed to increase the amount of fine MX precipitates and reduce coarse M23C6 carbides through alloy composition refinement and the application of thermomechanical treatments. The microstructural investigations by TEM have shown M23C6, M2X, and MX precipitation after tempering at 700 °C/2h with low dislocation recovery, while at 825 °C/2h the martensite developed to subgrain formation and growth. At this stage, only M23C6 and MX were detected. Preliminary results demonstrate that it is feasible to produce fine MX strengthened particles dispersed in the matrix with further optimization of tempering treatments.

  17. Corrigendum to 'On the influence of microstructure on the fracture behaviour of hot extruded ferritic ODS steels' [J. Nucl. Mater. 497 (2017) 60-75

    Science.gov (United States)

    Das, A.; Viehrig, H. W.; Altstadt, E.; Heintze, C.; Hoffmann, J.

    2018-02-01

    ODS steels are known to show inferior fracture properties as compared to ferritic martensitic non-ODS steels. Hot extruded 13Cr ODS steel however, showed excellent fracture toughness at a temperature range from room temperature to 400 °C. In this work, the factors which resulted in superior and anisotropic fracture behaviour were investigated by comparing different orientations of two hot extruded materials using scanning electron, electron backscatter and transmission electron microscopy. Fracture behaviour of the two materials was compared using unloading compliance fracture toughness tests. Anisotropic fracture toughness was predominantly influenced by grain morphology. Superior fracture toughness in 13Cr ODS-KIT was predominantly influenced by factors such as smaller void inducing particle size and higher sub-micron particle-matrix interfacial strength.

  18. MICROALLOYED STEELS FOR THE AUTOMOTIVE INDUSTRY

    Directory of Open Access Journals (Sweden)

    Debanshu Bhattacharya

    2014-12-01

    Full Text Available Two major drivers for the use of newer steels in the automotive industry are fuel efficiency and increased safety performance. Fuel efficiency is mainly a function of weight of steel parts, which in turn, is controlled by gauge and design. Safety is determined by the energy absorbing capacity of the steel used to make the part. All of these factors are incentives for the U.S. automakers to use both Highly Formable and Advanced High Strength Steels (AHSS to replace the conventional steels used to manufacture automotive parts in the past. AHSS is a general term used to describe various families of steels. The most common AHSS is the dual-phase steel that consists of a ferrite-martensite microstructure. These steels are characterized by high strength, good ductility, low tensile to yield strength ratio and high bake hardenability. Another class of AHSS is the complex-phase or multi-phase steel which has a complex microstructure consisting of various phase constituents and a high yield to tensile strength ratio. Transformation Induced Plasticity (TRIP steels is another class of AHSS steels finding interest among the U.S. automakers. These steels consist of a ferrite-bainite microstructure with significant amount of retained austenite phase and show the highest combination of strength and elongation, so far, among the AHSS in use. High level of energy absorbing capacity combined with a sustained level of high n value up to the limit of uniform elongation as well as high bake hardenability make these steels particularly attractive for safety critical parts and parts needing complex forming. A relatively new class of AHSS is the Quenching and Partitioning (Q&P steels. These steels seem to offer higher ductility than the dual-phase steels of similar strengths or similar ductility as the TRIP steels at higher strengths. Finally, martensitic steels with very high strengths are also in use for certain parts. The most recent initiative in the area of AHSS

  19. Effects of temperature and strain rate on the tensile behaviors of SIMP steel in static lead bismuth eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Jian, E-mail: jliu12b@imr.ac.cn [Institute of Metal Research, Chinese Academy of Sciences, Shenyang, 110016 (China); University of Chinese Academy of Sciences, Beijing, 100049 (China); Yan, Wei [Institute of Metal Research, Chinese Academy of Sciences, Shenyang, 110016 (China); Sha, Wei [School of Planning, Architecture and Civil Engineering, Queen' s University Belfast, Belfast, BT9 5AG (United Kingdom); Wang, Wei; Shan, Yiyin [Institute of Metal Research, Chinese Academy of Sciences, Shenyang, 110016 (China); Yang, Ke, E-mail: kyang@imr.ac.cn [Institute of Metal Research, Chinese Academy of Sciences, Shenyang, 110016 (China)

    2016-05-15

    In order to assess the susceptibility of candidate structural materials to liquid metal embrittlement, this work investigated the tensile behaviors of ferritic-martensitic steel in static lead bismuth eutectic (LBE). The tensile tests were carried out in static lead bismuth eutectic under different temperatures and strain rates. Pronounced liquid metal embrittlement phenomenon is observed between 200 °C and 450 °C. Total elongation is reduced greatly due to the liquid metal embrittlement in LBE environment. The range of ductility trough is larger under slow strain rate tensile (SSRT) test. - Highlights: • The tensile behaviors of SIMP steel in LBE are investigated for the first time. • The SIMP is susceptible to LME at different strain rates and temperatures. • The total elongation is reduced greatly. • The ductility trough is wider under SSRT. • The tensile specimens rupture in brittle manner without obvious necking.

  20. Effects of temperature and strain rate on the tensile behaviors of SIMP steel in static lead bismuth eutectic

    International Nuclear Information System (INIS)

    Liu, Jian; Yan, Wei; Sha, Wei; Wang, Wei; Shan, Yiyin; Yang, Ke

    2016-01-01

    In order to assess the susceptibility of candidate structural materials to liquid metal embrittlement, this work investigated the tensile behaviors of ferritic-martensitic steel in static lead bismuth eutectic (LBE). The tensile tests were carried out in static lead bismuth eutectic under different temperatures and strain rates. Pronounced liquid metal embrittlement phenomenon is observed between 200 °C and 450 °C. Total elongation is reduced greatly due to the liquid metal embrittlement in LBE environment. The range of ductility trough is larger under slow strain rate tensile (SSRT) test. - Highlights: • The tensile behaviors of SIMP steel in LBE are investigated for the first time. • The SIMP is susceptible to LME at different strain rates and temperatures. • The total elongation is reduced greatly. • The ductility trough is wider under SSRT. • The tensile specimens rupture in brittle manner without obvious necking.

  1. Correlation of microstructure and strain hardening behavior in the ultrafine-grained Nb-bearing dual phase steels

    Energy Technology Data Exchange (ETDEWEB)

    Ghatei Kalashami, A. [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Kermanpur, A., E-mail: ahmad_k@cc.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Ghassemali, E. [Jönköping University, School of Engineering, Department of Materials and Manufacturing, P.O. Box 1026, SE-551 11 Jönköping (Sweden); Najafizadeh, A. [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Mazaheri, Y. [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Department of Materials Engineering, Bu-Ali Sina University, Hamedan 65178-38695 (Iran, Islamic Republic of)

    2016-12-15

    Ultrafine-grained dual phase (DP) steels with different Nb contents (0.00, 0.06 and 0.12 wt%) were produced by cold-rolling followed by intercritical annealing of ferrite/martensite starting microstructure at 770 °C for different holding times. Scanning electron microscopy, equipped with electron backscattered diffraction (EBSD) detector, nanoindentation and tensile testing were used to characterize microstructural evolutions and their correlations to the strain hardening and fracture behavior. EBSD results confirmed the retardation effect of Nb on recrystallization. It was found that the strains stored in the grains and density of geometrically necessary dislocations (GNDs) were increased with the addition of Nb. Strain hardening analysis showed that plastic deformation of the DP steels occurred in three distinct stages, which based on the EBSD results, nanoindentation and fracture analysis, were controlled by microstructural features such martensite volume fraction and size, density of GNDs and individual ferrite and martensite tensile properties.

  2. Comparison of various 9-12%Cr steels under fatigue and creep-fatigue loadings at high temperature

    International Nuclear Information System (INIS)

    Fournier, B.; Dalle, F.; Sauzay, M.; Longour, J.; Salvi, M.; Caes, C.; Tournie, I.; Giroux, P.F.; Kim, S.H.

    2011-01-01

    The present article compares the cyclic behaviour of various 9-12%Cr steels, both commercial grades and optimized materials (in terms of creep strength). These materials were subjected to high temperature fatigue and creep-fatigue loadings. TEM examinations of the microstructure after cyclic loadings were also carried out. It appears that all the tempered ferritic-martensitic steels suffer from a cyclic softening effect linked to the coarsening of the sub-grains and laths and to the decrease of the dislocation density. These changes of the microstructure lead to a drastic loss in creep strength for all the materials under study. However, due to a better precipitation state, several materials optimized for their creep strength still present a good creep resistance after cyclic softening. These results are discussed and compared to the literature in terms of the physical mechanisms responsible for cyclic and creep deformation at the microstructural scale. (authors)

  3. On the empirical determination of positron trapping coefficient at nano-scale helium bubbles in steels irradiated in spallation target

    Science.gov (United States)

    Krsjak, Vladimir; Kuriplach, Jan; Vieh, Christiane; Peng, Lei; Dai, Yong

    2018-06-01

    In the present work, the specific positron trapping rate of small helium bubbles was empirically derived from positron annihilation lifetime spectroscopy (PALS) and transmission electron microscopy (TEM) studies of Fe9Cr martensitic steels. Both techniques are well known to be sensitive to nanometer-sized helium-filled cavities induced during irradiation in a mixed proton-neutron spectrum of spallation target. Complementary TEM and PALS studies show that positrons are being trapped to these defects at a rate of 1.2 ± 0.8 × 10-14 m3s-1. This suggests that helium bubbles in ferritic/martensitic steels are attractive traps for positrons comparable to mono-vacancies and quantitative analysis of the bubbles by PALS technique is plausible.

  4. First results on T91 claddings with and without modified FeCrAlY coatings exposed in PbBi under varying conditions

    International Nuclear Information System (INIS)

    Weisenburger, A.; Heinzel, A.; Miller, G.; Rousanov, A.

    2008-01-01

    It is well known that at temperatures above 500 deg C low activation austenitic steels suffer from severe corrosion in lead or lead-bismuth. Low activation martensitic steels instead form under similar conditions concerning temperature and oxygen content thick oxide scales that periodically may span off. Both groups of materials are therefore restricted to areas having lower temperature load. For parts that are intended to be used in high-temperature regions, like claddings, surface protection has to be applied. From gas turbines the role of elements forming thin stable oxide scales is well understood. The concept chosen here for thermally high loaded parts, the claddings, is the deposition of a FeCrAlY coating of about 30 vt,m thickness that is afterwards re-melted applying a pulsed electron beam (GESA). The beam energy is adjusted in a way to melt the entire coating together with a few thin region of the bulk to create a perfect intermixing at the boundary. This results in a new surface area of the cladding with an aluminium content of the order of 5 wt.% that will be sufficiently high to grow thin stable oxide scales. This concept is proven for austenitic cladding materials like 1.4970 as well as for martensitic ones like T91. In long-term corrosion tests the compatibility to Pb or PbBi, the resistance against corrosion and severe oxidation, was clearly demonstrated. No negative response of such a modified coating on the mechanical properties and the stability under irradiation has been observed as of yet. This paper will focus on the surface modification process, the corrosion results thus far obtained and on the evaluation of some mechanical properties. For example, the swelling of the fuel by irradiation will lead during operation to an increase of the internal pressure. This is simulated in experiments where an internal pressure of defined value was applied on T91 cladding tubes. The influence of flow velocity between to 3 m/s on the oxidation behaviour of T

  5. High temperature oxidation test of oxide dispersion strengthened (ODS) steel claddings

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

    2006-07-01

    In a feasibility study of ODS steel cladding, its high temperature oxidation resistance was evaluated. Although addition of Cr is effective for preventing high temperature oxidation, excessively higher amount of Cr leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, high temperature oxidation test was conducted for ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) 9Cr-ODS martensitic and 12Cr-ODS ferritic steel have superior high temperature oxidation resistance compared to 11mass%Cr PNC-FMS and even 17mass% SUS430 and equivalent to austenitic PNC316. (2) The superior oxidation resistance of ODS steel was attributed to earlier formation of the protective alpha-Cr 2 O 3 layer at the matrix and inner oxide scale interface. The grain size of ODS steel is finer than that of PNC-FMS, so the superior oxidation resistance of ODS steel can be attributed to the enhanced Cr-supplying rate throughout the accelerated grain boundary diffusion. Finely dispersed Y 2 O 3 oxide particles in the ODS steel matrix may also stabilized the adherence between the protective alpha-Cr 2 O 3 layer and the matrix. (author)

  6. Development of ferritic steels for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Development of ferritic steels for fusion reactor applications

    International Nuclear Information System (INIS)

    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

  8. Accelerated development of advanced steels for nuclear applications

    International Nuclear Information System (INIS)

    Ghoniem, N.; Zinkle, S.

    2009-01-01

    Significant progress has been achieved in the operational performance and radiation resistance of ferritic-martensitic steels during the past few decades. Conventional high temperature steels, such as HT-9 and 2 1/4 Cr-1Mo have evolved into super Oxide Dispersion Strengthened (ODS) steels through successive optimization to meet strict performance and radiation-resistance constraints. Such evolution was possible through a combination of experimentation, modeling and empirical information. Further development and optimization of structural steels in nuclear applications will require full utilization of the available array of sophisticated experimental techniques and multiscale computational modeling, in addition to empirical data. We present here a systematic approach to the process of optimum steel development, by linking material fabrication to thermo-mechanical properties through a physical understanding of microstructure evolution. The optimization process is based on the application of design constraints (e.g. low activation, low DBTT, low swelling, creep resistance, and weldability) to describe the required microstructures, which in turn, can be controlled through material processing techniques. Prospects for future design of optimum structural steels in nuclear applications by utilization of the hierarchy of multiscale experimental and computational strategies will be described. (author)

  9. Fabrication and characterization of a Spanish RAFM steel

    International Nuclear Information System (INIS)

    Rodriguez, D.; Serrano, M.; Moran, A.; Artimez, J. M.

    2009-01-01

    One of the main challenges for the realization of the future fusion reactor is the development and qualification of structural materials for first wall and breeding blanket. The fusion reactor application requires materials resistant to radiation damage, with excellent mechanical properties at high temperatures, good corrosion behaviour and reduced activation potential. Reduced Activation (RAFM) 9Cr Ferritic/Martenistic steels are the main candidates for first wall and blanket of fusion reactors, due to their resistance to swelling and excellent structural and thermal properties. These steels are based on the classical Cr-Mo steel grades but with a chemical composition modified in order to fulfil the low activation requirements, substituting the alloying elements with long decay times due to high activation by neutron irradiation. For this purpose the Mo is replaced by W, the Nb by Ta and Ni is removed. A summary of the activities related to the evaluation of the microstructural and mechanical properties of a reduced activation ferritic/martensitic steel fabricated at a semi-industrial scale in Spain will be presented in this paper. The steel chemical composition fulfils or is very close to the compositional specifications and metallurgical properties of the EUROFER steel. This activity corresponds to the ITMA and CIEMAT participation on Task 4 of the CONSOLIDER TECNO F US INGENIO 2010, financed by the Spanish Ministry of Science and Innovation. (author)

  10. Nanoindentation study of ferrite–martensite dual phase steels developed by a new thermomechanical processing

    Energy Technology Data Exchange (ETDEWEB)

    Mazaheri, Yousef, E-mail: y.mazaheri@ma.iut.ac.ir [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of); Faculty of Engineering, Bu-Ali Sina University, Hamedan 65178-38695 (Iran, Islamic Republic of); Kermanpur, Ahmad; Najafizadeh, Abbas [Department of Materials Engineering, Isfahan University of Technology, Isfahan 84156-83111 (Iran, Islamic Republic of)

    2015-07-15

    Dual phase (DP) steels consisting different volume fractions of ferrite and martensite and different ferrite grain size were produced by a new route utilizing cold-rolling and subsequent intercritical annealing of ferrite/martensite duplex starting structure at 770 °C for different times. Scanning electron microscopy has been supplemented by nanoindentation and tensile test to follow microstructural changes and their correlations to the variation in phase's hardness and mechanical properties. The results showed that longer holding times resulted in coarser and softer ferrite grains in DP microstructures. Martensite nanohardness variation with holding time is related to change in its carbon content. Mechanical properties such as strength, elongation and toughness are well correlated with the martensite/ferrite hardness ratio.

  11. Bulk Nanostructured FCC Steels With Enhanced Radiation Tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xinghang; Hartwig, K. Ted; Allen, Todd; Yang, Yong

    2012-10-27

    The objective of this project is to increase radiation tolerance in austenitic steels through optimization of grain size and grain boundary (GB) characteristics. The focus will be on nanocrystalline austenitic Fe-Cr-Ni alloys with an fcc crystal structure. The long-term goal is to design and develop bulk nanostructured austenitic steels with enhanced void swelling resistance and substantial ductility, and to enhance their creep resistance at elevated temperatures via GB engineering. The combination of grain refinement and grain boundary engineering approaches allows us to tailor the material strength, ductility, and resistance to swelling by 1) changing the sink strength for point defects, 2) by increasing the nucleation barriers for bubble formation at GBs, and 3) by changing the precipitate distributions at boundaries. Compared to ferritic/martensitic steels, austenitic stainless steels (SS) possess good creep and fatigue resistance at elevated temperatures, and better toughness at low temperature. However, a major disadvantage of austenitic SS is that they are vulnerable to significant void swelling in nuclear reactors, especially at the temperatures and doses anticipated in the Advanced Burner Reactor. The lack of resistance to void swelling in austenitic alloys led to the switch to ferritic/martensitic steels as the preferred material for the fast reactor cladding application. Recently a type of austenitic stainless steel, HT-UPS, was developed at ORNL, and is expected to show enhanced void swelling resistance through the trapping of point defects at nanometersized carbides. Reducing the grain size and increasing the fraction of low energy grain boundaries should reduce the available radiation-produced point defects (due to the increased sink area of the grain boundaries), should make bubble nucleation at the boundaries less likely (by reducing the fraction of high-energy boundaries), and improve the strength and ductility under radiation by producing a higher

  12. Mechanical characteristics of heterogeneous structures obtained by high-temperature brazing of corrosion-resistant steels with rapidly quenched non-boron nickel-based alloys

    Science.gov (United States)

    Kalin, B.; Penyaz, M.; Ivannikov, A.; Sevryukov, O.; Bachurina, D.; Fedotov, I.; Voennov, A.; Abramov, E.

    2018-01-01

    Recently, the use rapidly quenched boron-containing nickel filler metals for high temperature brazing corrosion resistance steels different classes is perspective. The use of these alloys leads to the formation of a complex heterogeneous structure in the diffusion zone that contains separations of intermediate phases such as silicides and borides. This structure negatively affects the strength characteristics of the joint, especially under dynamic loads and in corrosive environment. The use of non-boron filler metals based on the Ni-Si-Be system is proposed to eliminate this structure in the brazed seam. Widely used austenitic 12Cr18Ni10Ti and ferrite-martensitic 16Cr12MoSiWNiVNb reactor steels were selected for research and brazing was carried out. The mechanical characteristics of brazed joints were determined using uniaxial tensile and impact toughness tests, and fractography was investigated by electron microscopy.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Tempering of Mn and Mn-Si-V dual-phase steels

    Science.gov (United States)

    Speich, G. R.; Schwoeble, A. J.; Huffman, G. P.

    1983-06-01

    Changes in the yield behavior, strength, and ductility of a Mn and a Mn-Si-V d11Al-phase (ferrite-martensite) steel were investigated after tempering one hour at 200 to 600 °C. The change in yield behavior was complex in both steels with the yield strength first increasing and then decreasing as the tempering temperature was increased. This complex behavior is attributed to a combination of factors including carbon segregation to dislocations, a return of discontinuous yielding, and the relief of resid11Al stresses. In contrast, the tensile strength decreased continuously as the tempering temperature was increased in a manner that could be predicted from the change in hardness of the martensite phase using a simple composite strengthening model. The initial tensile ductility (total elongation) of the Mn-Si-V steel was much greater than that of the Mn steel. However, upon tempering up to 400 °C, the ductility of the Mn-Si-V decreased whereas that of the Mn steel increased. As a result, both steels had similar ductilities after tempering at 400 °C or higher temperatures. These results are attributed to the larger amounts of retained austenite in the Mn-Si-V steel (9 pct) compared to the Mn steel (3 pct) and its contribution to tensile ductility by transforming to martensite during plastic straining. Upon tempering at 400 °C, the retained austenite decomposes to bainite and its contribution to tensile ductility is eliminated.

  15. Oxidation and stress corrosion cracking of stainless steels in SCWRs

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Castro, L.; Blazquez, F.

    2008-01-01

    SCWRs are high-temperature, high-pressure, water-cooled reactors that operate above the thermodynamic critical point of water (374 deg C, 22.1 MPa). The SCWR offers many advantages compared to state-of- the-art LWRs including the use of a single phase coolant with high enthalpy, the elimination of components such as steam generators and steam separators and dryers, a low coolant mass inventory resulting in smaller components, and a much higher efficiency ∼ 44% vs. 33% in current LWRs). In these systems high pressure (25 MPa) coolant enters the vessel at 280 deg C which is heated to about 500 deg C and delivered to a power conversion cycle. Supercritical water (SCW) exhibits properties significantly different from those of liquid water below the critical point. Supercritical water acting essentially as a non-polar dense gas with solvation properties approaching those of a low-polarity organic. In this conditions, can dissolve gases like oxygen to complete miscibility. Depending upon what species are present and how much oxygen is present in the solution can becomes a very aggressive oxidising environment. Most of the data on corrosion in supercritical water are from fossil plant or oxidation waste disposal systems. However there is very limited data on corrosion in low conductivity de-aerated SCW and less on stress corrosion cracking behaviour under operating conditions foreseen for SCWR. Candidate materials for structural components are materials for high temperatures and include ferritic-martensitic alloys; oxide dispersion strengthened (ODS) ferritic/martensitic steels and strengthened steels by precipitation and for lower temperatures the austenitic stainless steels, such as 304 and 316, used in the LWR. Low swelling austenitic steels are also of high interest for areas with high dpa and high temperature. A review of the available information on corrosion and stress corrosion behaviour of different types of stainless steels in supercritical water at high

  16. Investigations on the growth kinetics of Laves phase precipitates in 12% Cr creep-resistant steels: Experimental and DICTRA calculations

    International Nuclear Information System (INIS)

    Prat, O.; Garcia, J.; Rojas, D.; Carrasco, C.; Inden, G.

    2010-01-01

    The growth kinetics of Laves phase precipitates (type Fe 2 W) in the early stage of creep (650 deg. C for 10,000 h) in two 12% Cr ferrite-martensitic steels has been investigated. In one alloy the Laves phase formed on tempering, while in the second alloy the Laves phase precipitated during creep. Kinetic simulations were performed using the software DICTRA. The particle size of the Laves phase was measured on transmission electron microscopy samples. The equilibrium phase fraction of the Laves phase was reached in the first thousand hours. Simulations of particle growth showed good agreement with the experimental results. Competitive growth between M 23 C 6 and the Laves phase showed that M 23 C 6 carbides reached their equilibrium after 12 days, whereas the Laves phase reached equilibrium after 3 months. Simulations of the influence of the interfacial energy and addition of Co, Cu and Si on Laves phase precipitation are presented.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  18. Effect of hot-dip galvanizing processes on the microstructure and mechanical properties of 600-MPa hot-dip galvanized dual-phase steel

    Science.gov (United States)

    Kuang, Chun-fu; Zheng, Zhi-wang; Wang, Min-li; Xu, Quan; Zhang, Shen-gen

    2017-12-01

    A C-Mn dual-phase steel was soaked at 800°C for 90 s and then either rapidly cooled to 450°C and held for 30 s (process A) or rapidly cooled to 350°C and then reheated to 450°C (process B) to simulate the hot-dip galvanizing process. The influence of the hot-dip galvanizing process on the microstructure and mechanical properties of 600-MPa hot-dip galvanized dual-phase steel (DP600) was investigated using optical microscopy, scanning electron microscopy (SEM), transmission electron microscopy (TEM), and tensile tests. The results showed that, in the case of process A, the microstructure of DP600 was composed of ferrite, martensite, and a small amount of bainite. The granular bainite was formed in the hot-dip galvanizing stage, and martensite islands were formed in the final cooling stage after hot-dip galvanizing. By contrast, in the case of process B, the microstructure of the DP600 was composed of ferrite, martensite, bainite, and cementite. In addition, compared with the yield strength (YS) of the DP600 annealed by process A, that for the DP600 annealed by process B increased by approximately 50 MPa because of the tempering of the martensite formed during rapid cooling. The work-hardening coefficient ( n value) of the DP600 steel annealed by process B clearly decreased because the increase of the YS affected the computation result for the n value. However, the ultimate tensile strength (UTS) and elongation ( A 80) of the DP600 annealed by process B exhibited less variation compared with those of the DP600 annealed by process A. Therefore, DP600 with excellent comprehensive mechanical properties (YS = 362 MPa, UTS = 638 MPa, A 80 = 24.3%, n = 0.17) was obtained via process A.

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

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Yongqiang; Liu, Juan [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); University of Chinese Academy of Science, Beijing 100049 (China); Zhang, Chonghong, E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); Chen, Jiachao [Paul Scherrer Institute, Villigen PSI (Switzerland); Yang, Yitao; Zhang, Liqing; Song, Yin [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China)

    2015-06-15

    Highlights: • Fractal dimensions of fracture surfaces of steels before and after irradiation were calculated. • Fractal dimension can effectively describe change of fracture surfaces induced by irradiation. • Correlation of change of fractal dimension with embrittlement of irradiated steels is discussed. - Abstract: 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.

  20. Comparison of corrosion behavior of EUROFER and CLAM steels in flowing Pb–15.7Li

    Energy Technology Data Exchange (ETDEWEB)

    Konys, J., E-mail: juergen.konys@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Krauss, W. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Zhu, Z.; Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2014-12-15

    Ferritic martensitic steels are envisaged to be applied as structural materials in HCLL blanket systems. Their compatibility with the liquid breeder, which is in direct contact with the structural alloy, will be essential for reliable and safe operation of the designed blankets. Formerly performed corrosion tests of RAFM steels in PICOLO loop of KIT were mainly done at high flow velocities, e.g., 0.22 m/s and delivered severe attack with material loss rates above 400 μm/yr at 823 K. Meanwhile, flow velocities for corrosion testing have been reduced into the “cm range” to be near fusion relevant conditions. Among the international ITER-partners, many varieties of RAFM steels have been developed and manufactured within the last decade, e.g., the so-called Chinese Low Activation Martensitic steel (CLAM). In this paper, the long term corrosion behavior of EUROFER and CLAM steel in flowing Pb–15.7Li will be presented at a flow velocity of about 0.10 m/s and compared with earlier obtained results of RAFM steels exposed at other operation parameters of PICOLO loop. The observed corrosion attack is near 220 μm/yr and fits well to predictions made by MATLIM-modeling for low flow velocities in the turbulent flow regime.

  1. Effect of microstructure on low cycle fatigue properties of ODS steels

    Energy Technology Data Exchange (ETDEWEB)

    Kubena, Ivo, E-mail: kubena@ipm.cz [IPM, Institute of Physics of Materials, Academy of Sciences of the Czech Republic, Zizkova 22, Brno (Czech Republic); Fournier, Benjamin [CEA/DEN/DANS/DMN/SRMA, Bat. 453, 91191 Gif-sur-Yvette Cedex (France); Kruml, Tomas [CEITEC IPM, Institute of Physics of Materials, Academy of Sciences of the Czech Republic, Zizkova 22, Brno (Czech Republic)

    2012-05-15

    Highlights: Black-Right-Pointing-Pointer Three various ODS steels are studied and compared. Black-Right-Pointing-Pointer Low cycle fatigue data at RT, 650 Degree-Sign C and 750 Degree-Sign C are given. Black-Right-Pointing-Pointer Microstructural characterization. Black-Right-Pointing-Pointer Detailed discussion of strengthening mechanisms. - Abstract: Low cycle fatigue properties at room temperature, 650 Degree-Sign C and 750 Degree-Sign C of three high chromium steels (9%Cr ferritic-martensitic and two 14%Cr ferritic steels) strengthened by oxide dispersion were studied and compared. Cyclic softening/hardening curves, cyclic deformation curves, S-N curves and Coffin-Manson curves are presented together with microstructural observations. Differences in cyclic response, stress level and fatigue life are attributed to differences in the matrix microstructure. The oxide particles stabilize the cyclic response, even if cyclic softening is detected for some experimental conditions. The strength of these steels is discussed in terms of strengthening mechanisms such as grain size effect, particle-dislocations interaction and dislocation density. Comparing three different ODS steels offers an opportunity to tests the contribution of individual mechanisms to the cyclic strength. The reduction of fatigue life in one of the ferritic steels is explained by the presence of large grains, facilitating the fatigue crack nucleation and the early growth.

  2. Microstructure and mechanical properties of friction stir welded 9Cr ODS steel

    International Nuclear Information System (INIS)

    Min, Hyoung Kee; Kang, Suk Hoon; Noh, Sanghoon; Lee, Jung Gu; Jang, Jinsung; Kim, Tae Kyu

    2013-01-01

    It is well known that the welding of ODS steel with a conventional melting.solidification process is not adequate to reserve nano-oxide particles in the matrix homogeneously. To reserve nano-oxide particles in the matrix homogeneously, friction stir welding (FSW) is the most promising technique to join ODS alloys. In this study, the effects of FSW on the microstructure and mechanical properties of a ODS steel were studied to apply the FSW process to 9Cr ODS steels. Microstructures were observed by means of optical microscopy, electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). A tensile test and hardness test were carried out to the investigate mechanical properties. FSW could successfully produce defect-free welds on ODS plates. FSW produced a fine grain structure consisting of ferrite and martensite. Tensile strengths and elongations of the SZs were excellent at 298 K, compared to those of the BM. This study suggests that FSW might be an appropriate welding method of ODS steels. Oxide dispersion strengthened (ODS) ferritic-martensitic (FM) steel containing 9 wt%Cr is a promising candidate material for high temperature components operating in aggressive environments such as nuclear fusion and fission systems because of the excellent elevated temperature strength, corrosion and radiation resistance. These characteristics come from microstructures consisting of fine grains and nano-oxide particles dispersed in high number density. However, for more applications of ODS steel in nuclear systems, its weldability is the one of the barrier to be solved

  3. Weld characterization of RAFM steel. EBP structural materials milestone 3

    Energy Technology Data Exchange (ETDEWEB)

    Alamo, A. [Service de Recherches Metallurgiques Appliquees, CEA Centre d' Etudes Nucleaires de Saclay, Saclay (France); Fontes, A. [Service de Techniques Avancees, CEA Centre d' Etudes Nucleaires de Saclay, Saclay (France); Schaefer, L. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Gauthier, A.; Tavassoli, A.A. [CEA Centre d' Etudes Nucleaires de Saclay, Saclay (France); Van Osch, E.V.; Van der Schaaf [ed.] [ECN Netherlands Energy Research Foundation, Petten (Netherlands)

    1999-07-01

    In the long term part of the European Fusion technology programme welding of reduced activation ferritic martensitic (RAFM)steels takes a prominent place. The blanket structures are complex and welding is an important element in manufacturing procedures. In the 95-98 program several Structural Materials tasks of the European Blanket Project are devoted to welding of RAFM steels. In the milestone 3 defined for the program a review of the weld characterization was foreseen in 1998. The present report gives the status of tasks and the major conclusions and recommendations of the welding milestone meeting. The major conclusion is that defect free GTAW (Gas Tungsten Arc Welding), EBW (Electron Beam Welding) and diffusion welds can be accomplished, but further work is needed to assure quantitatively the service boundary conditions. Also for irradiated steel additional work is recommended for the 99-02 period. Development of filler wire material for the European reference RAFM: EUROFER97 is necessary. Establishment of weldability tests must be settled in the next period also. 14 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hari Babu, S., E-mail: shb@igcar.gov.in [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, TN (India); Rajkumar, K.V. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, TN (India); Hussain, S. [UGC-DAE CSR, Kokilamedu 603 104, TN (India); Amarendra, G.; Sundar, C.S. [Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, TN (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, TN (India)

    2013-01-15

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

  5. Reduced-activation steels: present status and future development

    International Nuclear Information System (INIS)

    Klueh, R.L.

    2007-01-01

    Full text of publication follows: Reduced-activation steels for fusion reactor applications were developed in the 1980's to replace the commercial elevated- temperature steels first considered. In the United States, this involved replacing Sandvik HT9 and modified 9Cr-1Mo steels. Reduced-activation steels, which were developed for more rapid radioactivity decay following exposure in a fusion neutron environment, were patterned after the commercial steels they were to replace. The objective for the reduced-activation steels was that they have strengths (yield stress and ultimate tensile strength from room temperature to 600 deg. C) and impact toughness (measured in a Charpy test) comparable to or better than the steels they were replacing. That objective was achieved in reduced-activation steels developed in Japan, Europe, and the United States. Since the reduced-activation steels were developed in the 1980's, reactor designers have been interested designs for increased efficiency of future fusion plants. This means reactors will need to operate at higher temperatures-above 550 deg. C, which is the upper-temperature limit for the reduced-activation steels. Although the tensile and impact toughness of the reduced-activation steels exceed those of the commercial steels they were patterned after, their creep-rupture properties are inferior to some of the commercial steels they replaced. furthermore, they are much inferior to commercial steels that have been developed since the 1980's. Reasons for why the creep-rupture properties for the new commercial ferritic/martensitic steels are superior to the earlier commercial steels and the reduced-activation steels were examined. The reasons involve compositional changes that were made in the earlier commercial steels to give the new commercial steels their superior properties. Computational thermodynamics calculations were carried out to compare the expected equilibrium phases. It appears that similar changes in composition

  6. High Cr ODS steels R and D for high burnup fuel cladding

    International Nuclear Information System (INIS)

    Kimura, A.; Kasada, R.; Kishimoto, H.; Iwata, N.; Cho, H.-S.; Toda, N.; Yutani, K.; Ukai, S.; Fujiwara, M.

    2007-01-01

    High-performance cladding materials is essential to realize highly efficient and high-burnup operation over 150 GWd/t of so called Generation IV nuclear energy systems, such as supercritical-water-cooled reactor (SCWR) and lead-cooled fast reactor (LFR). Oxide dispersion strengthening (ODS) ferritic/ martensitic steels, which contain 9-12%Cr, show rather high resistance to neutron irradiation embrittlement and high strength at elevated temperatures. However, their corrosion resistance is not good enough in SCW and in lead at high temperatures. High-Cr ODS steels have been developed to improve corrosion resistance. An increase in Cr content an addition resulted in a drastic improvement of corrosion resistance in SCW and in lead. On the contrary, high-Cr steels often show an enhancement of aging embrittlement as well as irradiation embrittlement. Anisotropy in tensile properties is another issue. In order to overwhelm these issues, surveillance tests of the material performance have been performed for high Cr-ODS steels produced by new processing technologies. It is demonstrated that the dispersion of nono-sized oxide particles in high density is effective to attain high-performance and high-Cr ODS steels have a high potential as fuel cladding materials for SCWR and LFR with high efficiency and high burnup. (authors)

  7. Austenite grain growth and microstructure control in simulated heat affected zones of microalloyed HSLA steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Lei [Department of Machine Tools and Factory Management, Technical University of Berlin, Pascalstraße 8 – 9, 10587, Berlin (Germany); Federal Institute for Materials Research and Testing, Unter den Eichen 87, 12205, Berlin (Germany); Kannengiesser, Thomas [Federal Institute for Materials Research and Testing, Unter den Eichen 87, 12205, Berlin (Germany); Institute of Materials and Joining Technology, Otto von Guericke University Magdeburg, Universitetsplatz 2, 39106, Magdeburg (Germany)

    2014-09-08

    The roles of microalloying niobium, titanium and vanadium for controlling austenite grain growth, microstructure evolution and hardness were investigated at different simulated heat affected zones (HAZ) for high strength low alloy (HSLA) S690QL steel. High resolution FEG-SEM has been used to characterize fine bainitic ferrite, martensite and nanosized second phases at simulated coarse and fine grain HAZs. It was found that for Ti bearing steel (Ti/N ratio is 2) austenite grain had the slowest growth rate due to the presence of most stable TiN. The fine cuboidal particles promoted intragranular acicular ferrite (IGF) formation. Nb bearing steel exhibited relatively weaker grain growth retardation compared with titanium bearing steels and a mixed microstructure of bainite and martensite was present for all simulated HAZs. IGF existed at coarse grain HAZ of Ti+V bearing steel but it was totally replaced by bainite at fine grain HAZs. Hardness result was closely related to the morphology of bainitic ferrite, intragranular ferrite and second phases within ferrite. The microstructure and hardness results of different simulated HAZs were in good agreement with welded experimental results.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, V.B. [Lorena School of Engineering, University of Sao Paulo, Lorena, SP, 12602-810 (Brazil); Sandim, H.R.Z., E-mail: hsandim@demar.eel.usp.br [Lorena School of Engineering, University of Sao Paulo, Lorena, SP, 12602-810 (Brazil); Raabe, D. [Max-Planck-Institut für Eisenforschung, Düsseldorf, D-40237 (Germany)

    2017-03-15

    Reduced-activation ferritic-martensitic (RAFM) Eurofer-97 steel is a candidate material for structural applications in future fusion reactors. Depending on the amount of prior cold rolling strain and annealing temperature, important solid-state softening reactions such as recovery, recrystallization, and grain growth occur. Eurofer-97 steel was cold rolled up to 70, 80 and 90% reductions in thickness and annealed in the ferrite phase field (below ≈ 800 °C). Changes in microstructure, micro-, and mesotexture were followed by orientation mappings provided by electron backscatter diffraction (EBSD). Eurofer-97 steel undergoes abnormal grain growth above 650 °C and this solid-state reaction seems to be closely related to the high mobility of a few special grain boundaries that overcome pinning effects caused by fine particles. This solid-state reaction promotes important changes in the microstructure and microtexture of this steel. Abnormal grain growth kinetics for each condition was determined by means of quantitative metallography. - Highlights: • Abnormal grain growth (AGG) occurs in Eurofer-97 steel deformed to several strains. • Kinetics of abnormal grain growth has been determined at 750 and 800 °C. • Significant changes in crystallographic texture take place during AGG. • Grain boundaries with misorientations above 45° may explain abnormal grain growth. • Local microstructural instabilities (coarsening of M23C6 carbides) also explain AGG.

  10. Behaviour and microstructure of stainless steels irradiated in the french fast breeder reactors

    International Nuclear Information System (INIS)

    Dubuisson, P.; Gilbon, D.

    1991-01-01

    The burn-up of Fast Breeder Reactors is limited by the irradiation induced dimensional changes and mechanical properties of structural materials used for replaceable in-core components. This paper describes the behaviour improvements and also the radiation-induced microstructures of the different steels used for fuel pin cladding and wrapper tubes in French reactors. Materials of fuel pin cladding are austenitic steels whose main problem is swelling. Improvements in swelling resistance by cold-working, titanium additions and modifications of matrix (Fe-Cr-Ni) from SA 316 to CW 15-15 Ti are shown. These improvements are correlated with a best stability of microstructure under irradiation. Beneficial effects of phosphorus addition and multistabilisation (NbVTi) on radiation induced microstructure and swelling resistance are also shown. Austenitic steels used for wrapper tubes are limited both by swelling and by void embrittlement. The ferritic F17 (17Cr), ferritic-martensitic EM12 (9Cr-2MoNbV) and martensitic EM10 (9Cr-1Mo) steels present high swelling resistance. Nevertheless radiation-induced embrittlement is observed in EM12 and especially in F17. This embrittlement results from a fine and uniform radiation enhanced precipitation in ferrite grains. By contrast, the microstructure of fully martensitic EM10 steel is mush more stable and its ductile-brizzle transition temperature stays below 0 0 C. 12 figs

  11. Erosion and deuterium retention of CLF-1 steel exposed to deuterium plasma

    Science.gov (United States)

    Qiao, L.; Wang, P.; Hu, M.; Gao, L.; Jacob, W.; Fu, E. G.; Luo, G. N.

    2017-12-01

    In recent years reduced activation ferritic martensitic steel has been proposed as the plasma-facing material in remote regions of the first wall. This study reports the erosion and deuterium retention behaviours in CLF-1 steel exposed to deuterium (D) plasma in a linear experimental plasma system as function of incident ion energy and fluence. The incident D ion energy ranges from 30 to 180 eV at a flux of 4 × 1021 D m-2 s-1 up to a fluence of 1025 D m-2. SEM images revealed a clear change of the surface morphology as functions of incident fluence and impinging energy. The mass loss results showed a decrease of the total sputtering yield of CLF-1 steel with increasing incident fluence by up to one order of magnitude. The total sputtering yield of CLF-1 steel after 7.2 × 1024 D m-2 deuterium plasma exposure reduced by a factor of 4 compared with that of pure iron, which can be attributed to the enrichment of W at the surface due to preferential sputtering of iron and chromium. After D plasma exposure, the total deuterium retention in CLF-1 steel samples measured by TDS decreased with increasing incident fluence and energy, and a clear saturation tendency as function of incident fluence or energy was also observed.

  12. Size-dependent characteristics of ultra-fine oxygen-enriched nanoparticles in austenitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin, E-mail: ymiao@anl.gov [Argonne National Laboratory, Lemont, IL 60439 (United States); University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Mo, Kun [Argonne National Laboratory, Lemont, IL 60439 (United States); Zhou, Zhangjian [University of Science and Technology Beijing, Beijing 100082 (China); Liu, Xiang; Lan, Kuan-Che [University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Zhang, Guangming [University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); University of Science and Technology Beijing, Beijing 100082 (China); Miller, Michael K.; Powers, Kathy A. [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States); Stubbins, James F. [University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); International Institute for Carbon-Neutral Energy Research (I2CNER), Kyushu University, Fukuoka 819-0395 (Japan)

    2016-11-15

    Here, a coordinated investigation of the elemental composition and morphology of ultra-fine-scale nanoparticles as a function of size within a variety of austenitic oxide dispersion-strengthened (ODS) steels is reported. Atom probe tomography was utilized to evaluate the elemental composition of these nanoparticles. Meanwhile, the crystal structures and orientation relationships were determined by high-resolution transmission electron microscopy. The nanoparticles with sufficient size (>4 nm) to maintain a Y{sub 2}Ti{sub 2−x}O{sub 7−2x} stoichiometry were found to have a pyrochlore structure, whereas smaller Y{sub x}Ti{sub y}O{sub z} nanoparticles lacked a well-defined structure. The size-dependent characteristics of the nanoparticles in austenitic ODS steels differ from those in ferritic/martensitic ODS steels. - Highlights: • The structural and chemical characteristics of nanoparticles are revealed. • Nanoparticles' crystal structure and elemental composition are size-dependent. • Characteristics of austenitic ODS steels are compared to that of an F/M ODS steel. • Hypothesis about the formation mechanism of nanoparticles is proposed accordingly.

  13. Peculiar Features of Thermal Aging and Degradation of Rapidly Quenched Stainless Steels under High-Temperature Exposures

    Science.gov (United States)

    Shulga, A. V.

    2017-12-01

    This article presents the results of comparative studies of mechanical properties and microstructure of nuclear fuel tubes and semifinished stainless steel items fabricated by consolidation of rapidly quenched powders and by conventional technology after high-temperature exposures at 600 and 700°C. Tensile tests of nuclear fuel tube ring specimens of stainless austenitic steel of grade AISI 316 and ferritic-martensitic steel are performed at room temperature. The microstructure and distribution of carbon and boron are analyzed by metallography and autoradiography in nuclear fuel tubes and semifinished items. Rapidly quenched powders of the considered steels are obtained by the plasma rotating electrode process. Positive influence of consolidation of rapidly quenched powders on mechanical properties after high-temperature aging is confirmed. The correlation between homogeneous distribution of carbon and boron and mechanical properties of the considered steel is determined. The effects of thermal aging and degradation of the considered steels are determined at 600°C and 700°C, respectively.

  14. Steel

    International Nuclear Information System (INIS)

    Zorev, N.N.; Astafiev, A.A.; Loboda, A.S.; Savukov, V.P.; Runov, A.E.; Belov, V.A.; Sobolev, J.V.; Sobolev, V.V.; Pavlov, N.M.; Paton, B.E.

    1977-01-01

    Steels also containing Al, N and arsenic, are suitable for the construction of large components for high-power nuclear reactors due to their good mechanical properties such as good through-hardening, sufficiently low brittleness conversion temperature and slight displacement of the latter with neutron irradiation. Defined steels and their properties are described. (IHOE) [de

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

    Science.gov (United States)

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

    1988-07-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  17. Prediction of long-term precipitate evolution in austenitic heat-resistant steels

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jae-Hyeok; Jung, Woo-Sang; Cho, Young Whan [Korea Institute of Science and Technology, Seoul (Korea, Republic of). Materials/Devices Div.; Kozeschnik, Ernst [Vienna Univ. of Technology (Austria). Inst. of Materials Science and Technology

    2010-07-01

    Numerical prediction of the long-term precipitate evolution in five different austenitic heat-resistant stainless steels, NF709, Super304H, Sanicro25, CF8C-PLUS and HTUPS has been carried out. MX and M{sub 23}C{sub 6} are predicted to remain as major precipitates during long-term aging in these steels. The addition of 3 wt% Cu produces very fine Cu-rich precipitates during aging in Super304H and Sanicro25. It is found that the amount of Z phase start to increase remarkably between 1,000 and 10,000 hours of aging at the expense of MX precipitates in the steels containing a high nitrogen content. However, the growth rate of Z phase is relatively slow and its average size reaches at most a few tens of nanometers after 100,000 hours of aging at 700 C, compared with 9-12% Cr ferritic/martensitic heat-resistant steels. The predicted precipitation sequence and precipitate size during aging are in general agreement with experimental observations. (orig.)

  18. Microstructure evolution of the oxide dispersion strengthened CLAM steel during mechanical alloying process

    Energy Technology Data Exchange (ETDEWEB)

    Song, Liangliang [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Liu, Shaojun, E-mail: shaojun.liu@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China); Mao, Xiaodong [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Science, Hefei, Anhui, 230031 (China)

    2016-11-15

    Highlights: • A nano-sized oxides dispersed ODS-CLAM steel was obtained by MA and HIP. • A minimum saturated grain size of down to 30 nm was achieved by varying the milling time from 0 to 100 h. • Solution of W in the MA powder could be significantly improved by increasing MA rotation speed. - Abstracts: Oxide dispersion strengthened Ferritic/Martensitic steel is considered as one of the most potential structural material for future fusion reactor, owing to its high mechanical properties and good irradiation resistance. The oxide dispersion strengthened China Low Activation Martensitic (ODS-CLAM) steel was fabricated by mechanical alloying (MA) and hot isostatic pressing (HIP). The microstructural evolutions during the process of ball milling and subsequent consolidation were investigated by SEM, XRD and TEM. The results showed that increasing the milling time during the first 36 h milling could effectively decrease the grain size to a value of around 30 nm, over which grain sized remained nearly constant. Increasing the rotation speed promoted the solution of tungsten (W) element obviously and decreased the grain size to a certain degree. Observation on the consolidated and further heat-treated ODS-CLAM steel samples indicated that a martensite microstructure with a high density of nano-particles was achieved.

  19. Friction stir welding of F82H steel for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon, E-mail: shnoh@kaeri.re.kr [Fusion Structural Materials Division, Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Nuclear Materials Division, Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon (Korea, Republic of); Ando, Masami; Tanigawa, Hiroyasu [Fusion Structural Materials Division, Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Fujii, Hidetoshi [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka (Japan); Kimura, Akihiko [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto (Japan)

    2016-09-15

    In the present study, friction stir welding was employed to join F82H steels and develop a potential joining technique for a reduced activation ferritic/martensitic steel. The microstructures and mechanical properties on the joint region were investigated to evaluate the applicability of friction stir welding. F82H steel sheets were successfully butt-joined with various welding parameters. In welding conditions, 100 rpm and 100 mm/min, the stirred zone represented a comparable hardness distribution with a base metal. Stirred zone induced by 100 rpm reserved uniformly distributed precipitates and very fine ferritic grains, whereas the base metal showed a typical tempered martensite with precipitates on the prior austenite grain boundary and lath boundary. Although the tensile strength was decreased at 550 °C, the stirred zone treated at 100 rpm showed comparable tensile behavior with base metal up to 500 °C. Therefore, friction stir welding is considered a potential welding method to preserve the precipitates of F82H steel.

  20. Friction stir welding of F82H steel for fusion applications

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Ando, Masami; Tanigawa, Hiroyasu; Fujii, Hidetoshi; Kimura, Akihiko

    2016-01-01

    In the present study, friction stir welding was employed to join F82H steels and develop a potential joining technique for a reduced activation ferritic/martensitic steel. The microstructures and mechanical properties on the joint region were investigated to evaluate the applicability of friction stir welding. F82H steel sheets were successfully butt-joined with various welding parameters. In welding conditions, 100 rpm and 100 mm/min, the stirred zone represented a comparable hardness distribution with a base metal. Stirred zone induced by 100 rpm reserved uniformly distributed precipitates and very fine ferritic grains, whereas the base metal showed a typical tempered martensite with precipitates on the prior austenite grain boundary and lath boundary. Although the tensile strength was decreased at 550 °C, the stirred zone treated at 100 rpm showed comparable tensile behavior with base metal up to 500 °C. Therefore, friction stir welding is considered a potential welding method to preserve the precipitates of F82H steel.

  1. Failure evaluation on a high-strength alloy SA213-T91 super heater tube of a power generation

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, J.; Purbolaksono, J.; Beng, L.C.; Ahmad, A. [University of Tenaga Nas, Kajang (Malaysia). Dept. of Mechanical Engineering

    2010-07-01

    This article presents failure investigation on a high-strength alloy SA213-T91 superheater tube. This failure is the first occurrence involving the material in Kapar Power Station Malaysia. The investigation includes visual inspections, hardness measurements, and microscopic examinations. The failed super-heater tube shows a wide open rupture with thin and blunt edges. Hardness readings on all the as-received tubes are used for estimating the operating metal temperature of the super-heater tubes. Microstructures of the failed tube show numerous creep cavities consisting of individual pores and chain of pores which form micro-and macro-cracks. The findings confirmed that the super-heater tube is failed by short-term overheating. Higher temperatures of the flue gas due to the inconsistent feeding of pulverized fuels into the burner is identified to cause overheating of the failed tube.

  2. The mechanical properties and microstructures of vanadium bearing high strength dual phase steels processed with continuous galvanizing line simulations

    Science.gov (United States)

    Gong, Yu

    For galvanized or galvannealed steels to be commercially successful, they must exhibit several attributes: (i) easy and inexpensive processing in the hot mill, cold mill and on the coating line, (ii) high strength with good formability and spot weldability, and (iii) good corrosion resistance. At the beginning of this thesis, compositions with a common base but containing various additions of V or Nb with or without high N were designed and subjected to Gleeble simulations of different galvanizing(GI), galvannealing(GA) and supercooling processing. The results revealed the phase balance was strongly influenced by the different microalloying additions, while the strengths of each phase were somewhat less affected. Our research revealed that the amount of austenite formed during intercritical annealing can be strongly influenced by the annealing temperature and the pre-annealing conditions of the hot band (coiling temperature) and cold band (% cold reduction). In the late part of this thesis, the base composition was a low carbon steel which would exhibit good spot weldability. To this steel were added two levels of Cr and Mo for strengthening the ferrite and increasing the hardenability of intercritically formed austenite. Also, these steels were produced with and without the addition of vanadium in an effort to further increase the strength. Since earlier studies revealed a relationship between the nature of the starting cold rolled microstructure and the response to CGL processing, the variables of hot band coiling temperature and level of cold reduction prior to annealing were also studied. Finally, in an effort to increase strength and ductility of both the final sheet (general formability) and the sheared edges of cold punched holes (local formability), a new thermal path was developed that replaced the conventional GI ferrite-martensite microstructure with a new ferrite-martensite-tempered martensite and retained austenite microstructure. The new

  3. Damage Evolution in Complex-Phase and Dual-Phase Steels during Edge Stretching

    Directory of Open Access Journals (Sweden)

    Nikky Pathak

    2017-03-01

    Full Text Available The role of microstructural damage in controlling the edge stretchability of Complex-Phase (CP and Dual-Phase (DP steels was evaluated using hole tension experiments. The experiments considered a tensile specimen with a hole at the center of specimen that is either sheared (sheared edge condition or drilled and then reamed (reamed edge condition. The damage mechanism and accumulation in the CP and DP steels were systematically characterized by interrupting the hole tension tests at different strain levels using scanning electron microscope (SEM analysis and optical microscopy. Martensite cracking and decohesion of ferrite-martensite interfaces are the dominant nucleation mechanisms in the DP780. The primary source of void nucleation in the CP800 is nucleation at TiN particles, with secondary void formation at martensite/bainite interfaces near the failure strain. The rate of damage evolution is considerably higher for the sheared edge in contrast with the reamed edge since the shearing process alters the microstructure in the shear affected zone (SAZ by introducing work-hardening and initial damage behind the sheared edge. The CP microstructures were shown to be less prone to shear-induced damage than the DP materials resulting in much higher sheared edge formability. Microstructural damage in the CP and DP steels was characterized to understand the interaction between microstructure, damage evolution and edge formability during edge stretching. An analytical model for void evolution and coalescence was developed and applied to predict the damage rate in these rather diverse microstructures.

  4. Damage Evolution in Complex-Phase and Dual-Phase Steels during Edge Stretching.

    Science.gov (United States)

    Pathak, Nikky; Butcher, Cliff; Worswick, Michael James; Bellhouse, Erika; Gao, Jeff

    2017-03-27

    The role of microstructural damage in controlling the edge stretchability of Complex-Phase (CP) and Dual-Phase (DP) steels was evaluated using hole tension experiments. The experiments considered a tensile specimen with a hole at the center of specimen that is either sheared (sheared edge condition) or drilled and then reamed (reamed edge condition). The damage mechanism and accumulation in the CP and DP steels were systematically characterized by interrupting the hole tension tests at different strain levels using scanning electron microscope (SEM) analysis and optical microscopy. Martensite cracking and decohesion of ferrite-martensite interfaces are the dominant nucleation mechanisms in the DP780. The primary source of void nucleation in the CP800 is nucleation at TiN particles, with secondary void formation at martensite/bainite interfaces near the failure strain. The rate of damage evolution is considerably higher for the sheared edge in contrast with the reamed edge since the shearing process alters the microstructure in the shear affected zone (SAZ) by introducing work-hardening and initial damage behind the sheared edge. The CP microstructures were shown to be less prone to shear-induced damage than the DP materials resulting in much higher sheared edge formability. Microstructural damage in the CP and DP steels was characterized to understand the interaction between microstructure, damage evolution and edge formability during edge stretching. An analytical model for void evolution and coalescence was developed and applied to predict the damage rate in these rather diverse microstructures.

  5. Irradiation effects of 11 MeV protons on ferritic steels

    International Nuclear Information System (INIS)

    Hamaguchi, Yoshikazu; Kuwano, Hisashi; Misawa, Toshihei

    1985-01-01

    It is considered that ferritic/martensitic steels are the candidate of the first wall materials for future fusion reactors. The most serious problem in the candidate materials is the loss of ductility due to the elevation of ductile-brittle transition temperature by the high dpa irradiation of neutrons. 14 MeV neutrons produced by D-T reaction cause high dpa damage and also produce large quantity of helium and hydrogen atoms in first wall materials. Those gas atoms also play an important role in the embrittlement of steels. The main purpose of this work was to simulate the behavior of hydrogen produced by the transmutation in the mechanical properties of ferritic steels when they were irradiated with 11 MeV protons. The experimental procedure and the results of hardness, the broadening of x-ray diffraction lines, Moessbauer spectroscopy and small punch test are reported. High energy protons of 10 - 20 MeV are suitable to the simulation experiment of 14 MeV neutron radiation damage. But the production of the active nuclei emitting high energy gamma ray and having long life, Co-56, is the most serious problem. Another difficulty is the control of irradiation temperature. A small irradiation chamber must be developed. (Kako, I.)

  6. Microstructural investigations of the trimmed edge of DP980 steel sheets

    Science.gov (United States)

    Bhattacharya, S.; Green, D. E.; Sohmshetty, R.; Alpas, A. T.

    2017-10-01

    In order to reduce vehicle weight while maintaining crashworthiness, advanced high strength steels (AHSSs), such as DP980, are extensively used for manufacturing automotive body components. During trimming operations, the high tensile strength of DP980 sheets tends to cause damage of the trim edge of D2 die inserts, which result in deterioration of the edge quality. The objective of this work is to study the damage microstructures at the trimmed edge of DP980 steel sheets as a function of the number of trimming cycles. A mechanical press equipped with AISI D2 tool steel inserts was used to continuously trim 1.4 mm thick sheets of DP980 at a rate of 30 strokes/min. Cross-sectional SEM images of the trimmed edges revealed that the sheared edge quality of the DP980 sheets decreased, indicated by an increase in the burr width, with an increase in the number of trims from 40,000 to 70,000. Plastic strains were estimated using the displacements of the martensite plates within plastic flow fields of ferrite. Site-specific cross-sectional TEM samples, excised from the trimmed edge using the in-situ `lift-out' technique by focused ion-beam (FIB)-milling, revealed cracking at the ferrite/martensite interfaces after 70,000 cycles indicating an increase in the depth of deformation zone possibly due to trimming with a chipped and blunted die edge.

  7. Microstructure, Mechanical and Corrosion Properties of Friction Stir-Processed AISI D2 Tool Steel

    Science.gov (United States)

    Yasavol, Noushin; Jafari, Hassan

    2015-05-01

    In this study, AISI D2 tool steel underwent friction stir processing (FSP). The microstructure, mechanical properties, and corrosion resistance of the FSPed materials were then evaluated. A flat WC-Co tool was used; the rotation rate of the tool varied from 400 to 800 rpm, and the travel speed was maintained constant at 385 mm/s during the process. FSP improved mechanical properties and produced ultrafine-grained surface layers in the tool steel. Mechanical properties improvement is attributed to the homogenous distribution of two types of fine (0.2-0.3 μm) and coarse (1.6 μm) carbides in duplex ferrite-martensite matrix. In addition to the refinement of the carbides, the homogenous dispersion of the particles was found to be more effective in enhancing mechanical properties at 500 rpm tool rotation rate. The improved corrosion resistance was observed and is attributed to the volume fraction of low-angle grain boundaries produced after friction stir process of the AISI D2 steel.

  8. Thermal Aging Evaluation of Mod. 9Cr-1Mo Steel using Nonlinear Rayleigh Waves

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young-Sang; Kim, Hoe-Woong; Kim, Jong-Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Marino, Daniel; Kim, Jin-Yeon; Jacobs, L.J [Georgia Institute of Technology, Atlanta (United States); Ruiz, Alberto [UMSNH, Morelia (Mexico)

    2014-10-15

    Thermal aging can pose a high risk to decreases in the mechanical properties such as strength or creep resistance. This can lead to an unexpected failure during long term operation. Nonlinear NDE techniques are preferred over conventional NDE techniques (linear ultrasonic measurements) because nonlinear ultrasonic techniques have shown their capability to detect a microstructural damage in the structures undergoing fatigue and creep. These nonlinear ultrasonic techniques make use of the fact that the dislocation density increases, which will create a nonlinear distortion of an ultrasonic wave; this damage causes the generation of measurable higher harmonic components in an initially mono-chromatic ultrasonic signal. This study investigates the recently developed non-contact nonlinear ultrasonic technique to detect the microstructural damage of mod. 9Cr-1Mo steel based on nonlinear Rayleigh wave with varying propagation distances. Nonlinear Rayleigh surface wave measurements using a non-contact, air-coupled ultrasonic transducer have been applied for the thermal aging evaluation of modified 9Cr-1Mo ferritic-martensitic steel. Thermal aging for various heat treatment times of mod.. 9Cr-1Mo steel specimens is performed to obtain the nucleation and growth of precipitated particles in specimens. The amplitudes of the first and second harmonics are measured along the propagation distance and the relative nonlinearity parameter is obtained from these amplitudes. The relative nonlinearity parameter shows a similar trend with the Rockwell C hardness.

  9. Microstructure and Tensile-Shear Properties of Resistance Spot-Welded Medium Mn Steel

    Directory of Open Access Journals (Sweden)

    Qiang Jia

    2018-01-01

    Full Text Available The medium Mn steels are gaining increasing attention due to their excellent combination of mechanical properties and material cost. A cold-rolled 0.1C5Mn medium Mn steel with a ferrite matrix plus metastable austenite duplex microstructure was resistance spot-welded with various welding currents and times. The nugget size rose with the increase of heat input, but when the welding current exceeded the critical value, the tensile-shear load increased slowly and became unstable due to metal expulsion. The fusion zone exhibited a lath martensite microstructure, and the heat-affected zone was composed of a ferrite/martensite matrix with retained austenite. The volume fraction of retained austenite decreased gradually from the base metal to the fusion zone, while the microhardness presented a reverse varying trend. Interfacial failure occurred along the interface of the steel sheets with lower loading capacity. Sufficient heat input along with serious expulsion brought about high stress concentration around the weld nugget, and the joint failed in partial interfacial mode. Pull-out failure was absent in this study.

  10. Materials design data for reduced activation martensitic steel type EUROFER

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. E-mail: tavassoli@cea.fr; Alamo, A.; Bedel, L.; Forest, L.; Gentzbittel, J.-M.; Rensman, J.-W.; Diegele, E.; Lindau, R.; Schirra, M.; Schmitt, R.; Schneider, H.C.; Petersen, C.; Lancha, A.-M.; Fernandez, P.; Filacchioni, G.; Maday, M.F.; Mergia, K.; Boukos, N.; Baluc,; Spaetig, P.; Alves, E.; Lucon, E

    2004-08-01

    Materials design limits derived so far from the data generated in Europe for the reduced activation ferritic/martensitic (RAFM) steel type Eurofer are presented. These data address the short-term needs of the ITER Test Blanket Modules and a DEMOnstration fusion reactor. Products tested include plates, bars, tubes, TIG and EB welds, as well as powder consolidated blocks and solid-solid HIP joints. Effects of thermal ageing and low dose neutron irradiation are also included. Results are sorted and screened according to design code requirements before being introduced in reference databases. From the physical properties databases, variations of magnetic properties, modulus of elasticity, density, thermal conductivity, thermal diffusivity, specific heat, mean and instantaneous linear coefficients of thermal expansion versus temperature are derived. From the tensile and creep properties databases design allowable stresses are derived. From the instrumented Charpy impact and fracture toughness databases, ductile to brittle transition temperature, toughness and behavior of materials in different fracture modes are evaluated. From the fatigue database, total strain range versus number of cycles to failure curves are plotted and used to derive fatigue design curves. Cyclic curves are also derived and compared with monotonic hardening curves. Finally, irradiated and aged materials data are compared to ensure that the safety margins incorporated in unirradiated design limits are not exceeded.

  11. Positron annihilation lifetime study of oxide dispersion strengthened steels

    International Nuclear Information System (INIS)

    Krsjak, V.; Szaraz, Z.; Hähner, P.

    2012-01-01

    A comparative positron annihilation lifetime study has been performed on various commercial ferritic and ferritic/martensitic oxide dispersion strengthened (ODS) steels. Both as-extruded and recrystallized materials were investigated. In the materials with recrystallized coarse-grained microstructures, only the positron trapping at small vacancy clusters and yttria nanofeatures was observed. Materials which had not undergone recrystallization treatment clearly showed additional positron trapping which is associated with dislocations. Dislocation densities were calculated from a two-component decomposition of the positron lifetime spectra by assuming the first component to be a superposition of the bulk controlled annihilation rate and the dislocation controlled trapping rate. The second component (which translates into lifetimes of 240–260 ps) was found to be well separated in all those ODS materials. This paper presents the potentialities and limitations of the positron annihilation lifetime spectroscopy, and discusses the results of the experimental determination of the defect concentrations and sensitivity of this technique to the material degradation due to thermally induced precipitation of chromium-rich α′ phases.

  12. Fracture Resistances of Y_2O_3 Particle Dispersion Strengthened 9Cr Steel at Room Temperature and High Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji Hyun; Kang, Suk Hoon; Lee, Yongbok; Kim, Sung Soo

    2012-01-01

    The fracture resistance and tensile properties of Y_2O_3 oxide dispersion strengthened steel containing 9 wt% Cr (9Cr-ODS) were measured at various temperatures up to 700°C. The fracture characteristics were compared with those of commercial E911 ferritic/martensitic steel. The strength of 9Cr-ODS was at least 30% higher than that of E911 steel at the test temperatures below 500°C. The strength difference between the two materials was almost diminished at 700°C. 9Cr-ODS showed cleavage fracture behavior at room temperature and unstable crack growth behaviors at 300°C and 500°C. The J-R fracture resistance of 9Cr-ODS was much lower than that of E911 steel at all temperatures. It was deduced that the coarse Cr_2O_3 particles that were formed during the alloying process provided the crack initiation sites of cleavage fracture in 9Cr-ODS.

  13. Al-Mn CVD-FBR coating on P92 steel as protection against steam oxidation at 650 °C: TGA-MS study

    Science.gov (United States)

    Castañeda, S. I.; Pérez, F. J.

    2018-02-01

    The initial stages oxidation of the P92 ferritic/martensitic steel with and without Al-Mn coating at 650 °C in Ar+40%H2O for 240 h were investigated by mass spectrometry (MS) and thermogravimetric analysis (TGA). TGA-MS measurements were conducted in a closed steam loop. An Al-Mn coating was deposited on P92 steel at 580 °C for 2 h by chemical vapour deposition in a fluidized bed reactor (CVD-FBR). The coating as-deposited was treated in the same reactor at 700 °C in Ar for 2h, in order to produce aluminide phases that form the protective alumina layer (Al2O3) during oxidation. MS measurements at 650 °C of the Al-Mn/P92 sample for 200 h indicated the presence of (Al-Mn-Cr-Fe-O) volatile species of small intensity. Uncoated P92 steel oxidized under the same steam oxidation conditions emitted greater intensities of volatile species of Cr, Fe and Mo in comparison with intensities from coated steel. TGA measurements verified that the mass gained by the coated sample was up to 300 times lower than for uncoated P92 steel. The morphology, composition and structure of samples by Scanning Electron Microscopy SEM, Backscattered Electron (BSE) detection, X-ray Energy Dispersive Spectrometry (EDAX) and X-ray Diffraction (XRD) are described.

  14. Corrosion of steels in molten gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li)

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Masatoshi, E-mail: kondo.masatoshi@nr.titech.ac.jp [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Ishii, Masaomi [Department of Nuclear Engineering, School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Muroga, Takeo [Department of Helical Plasma Research, National Institute for Fusion Science, Toki, Gifu 502-5292 (Japan)

    2015-10-15

    Graphical abstract: Corrosion of RAFM steel, JLF-1, in liquid Sn–20Li was caused by the formation of Fe-Sn alloyed layer. - Highlights: • The corrosion tests were performed for the reduced activation ferritic martensitic steel JLF-1 and the austenitic steel SUS316 in liquid Ga, Sn and Sn-20Li at 873 K up to 750 h. • The weight loss of the specimens exposed to liquid Ga, Sn and Sn-20Li was evaluated. • The corrosion of the steels in liquid Ga was caused by the alloying reaction between Ga and Fe on the steel surface. • The corrosion of the steels in liquid Sn was caused by the alloying reaction between Sn and Fe on the steel surface. • The corrosion of the steels in liquid Sn-20Li was caused by the formation of the Fe-Sn alloyed layer and the diffusion of Sn and Li into the steel matrix. - Abstract: The compatibility of steels in liquid gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li) was investigated by means of static corrosion tests. The corrosion tests were performed for reduced activation ferritic martensitic steel JLF-1 (JOYO-HEAT, Fe–9Cr–2W–0.1C) and austenitic steel SUS316 (Fe–18Cr–12Ni–2Mo). The test temperature was 873 K, and the exposure time was 250 and 750 h. The corrosion of these steels in liquid Ga, Sn and Sn–20Li alloy was commonly caused by the formation of a reaction layer and the dissolution of the steel elements into the melts. The reaction layer formed in liquid Ga was identified as Fe{sub 3}Ga from the results of metallurgical analysis and the phase diagram. The growth rate of the reaction layer on the JLF-1 steel showed a parabolic rate law, and this trend indicated that the corrosion could be controlled by the diffusion process through the layer. The reaction layer formed in liquid Sn and Sn–20Li was identified as FeSn. The growth rate had a linear function with exposure time. The corrosion in Sn and Sn–20Li could be controlled by the interface reaction on the layer. The growth rate of the layer formed

  15. Corrosion of steels in molten gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li)

    International Nuclear Information System (INIS)

    Kondo, Masatoshi; Ishii, Masaomi; Muroga, Takeo

    2015-01-01

    Graphical abstract: Corrosion of RAFM steel, JLF-1, in liquid Sn–20Li was caused by the formation of Fe-Sn alloyed layer. - Highlights: • The corrosion tests were performed for the reduced activation ferritic martensitic steel JLF-1 and the austenitic steel SUS316 in liquid Ga, Sn and Sn-20Li at 873 K up to 750 h. • The weight loss of the specimens exposed to liquid Ga, Sn and Sn-20Li was evaluated. • The corrosion of the steels in liquid Ga was caused by the alloying reaction between Ga and Fe on the steel surface. • The corrosion of the steels in liquid Sn was caused by the alloying reaction between Sn and Fe on the steel surface. • The corrosion of the steels in liquid Sn-20Li was caused by the formation of the Fe-Sn alloyed layer and the diffusion of Sn and Li into the steel matrix. - Abstract: The compatibility of steels in liquid gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li) was investigated by means of static corrosion tests. The corrosion tests were performed for reduced activation ferritic martensitic steel JLF-1 (JOYO-HEAT, Fe–9Cr–2W–0.1C) and austenitic steel SUS316 (Fe–18Cr–12Ni–2Mo). The test temperature was 873 K, and the exposure time was 250 and 750 h. The corrosion of these steels in liquid Ga, Sn and Sn–20Li alloy was commonly caused by the formation of a reaction layer and the dissolution of the steel elements into the melts. The reaction layer formed in liquid Ga was identified as Fe 3 Ga from the results of metallurgical analysis and the phase diagram. The growth rate of the reaction layer on the JLF-1 steel showed a parabolic rate law, and this trend indicated that the corrosion could be controlled by the diffusion process through the layer. The reaction layer formed in liquid Sn and Sn–20Li was identified as FeSn. The growth rate had a linear function with exposure time. The corrosion in Sn and Sn–20Li could be controlled by the interface reaction on the layer. The growth rate of the layer formed in

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

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tyburska-Puschel, Beata [Univ. of Wisconsin, Madison, WI (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    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. Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

    International Nuclear Information System (INIS)

    Aubert, P.; Tavassoli, F.; Rieth, M.; Diegele, E.; Poitevin, Y.

    2011-01-01

    EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 o C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.

  18. A study of local deformation and damage of dual phase steel

    International Nuclear Information System (INIS)

    Sirinakorn, T.; Wongwises, S.; Uthaisangsuk, V.

    2014-01-01

    Highlights: • Crack initiation in DP microstructure was investigated using 2D RVE simulation. • The GTN damage model was used to describe void evolution in ferrite and interphase. • Predicted damage at triple junctions agreed with observed crack initiation sites. • RVE with GNDs zone showed damages took place at earlier deformation state. • Overall loading condition exhibited significant effect on damage evolution rate. - Abstract: Deformation and fracture behavior of Dual Phase (DP) high strength steel were investigated by means of a microstructure based Finite Element (FE) modeling. Representative Volume Elements (RVEs) were applied to consider effects of various microstructure constituents and characteristics. Individual stress–strain curves were provided for ferrite, martensite as well as transformation induced Geometrically Necessary Dislocations (GNDs) taking into account in the RVEs. Principally, the GNDs occurred around phase boundaries during quenching process due to the austenite–martensite transformation. Flow behaviors of individual phases were defined on the basis of dislocation theory and partitioning of local chemical composition. Then, flow curves of the examined DP steel were predicted. Furthermore, the Gurson–Tvergaard–Needleman (GTN) model was used to represent ductile damage evolution in the microstructure. Occurrences of void initiation were characterized and damage parameters for RVE simulations were hence identified. Finally, influences of the GNDs, local stress and strain distributions and interactions between phases on predicted crack initiation in the DP microstructure were discussed and correlated with experimental results

  19. Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

    Science.gov (United States)

    Aubert, P.; Tavassoli, F.; Rieth, M.; Diegele, E.; Poitevin, Y.

    2011-02-01

    EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 °C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.

  20. Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, P., E-mail: Philippe.aubert@cea.fr [CEA Saclay, DEN/DM2S, F-91191 Gif sur Yvette (France); Tavassoli, F. [CEA Saclay, DEN/DMN, F-91191 Gif sur Yvette (France); Rieth, M. [KIT, IMF I, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Diegele, E.; Poitevin, Y. [Fusion for Energy (F4E), C/Josep Pla 2 - Ed. B3, 08019 Barcelona (Spain)

    2011-02-15

    EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 {sup o}C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.

  1. Review of candidate welding processes of RAFM steels for ITER test blanket modules and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, P., E-mail: philippe.aubert@cea.fr [CEA, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Tavassoli, F. [CEA, DEN, DMN, F-91191 Gif-sur-Yvette (France); Rieth, M. [FZK, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Diegele, E.; Poitevin, Y. [Fusion for Energy (F4E), C/Josep Pla 2 - Ed. B3, 08019 Barcelona (Spain)

    2011-10-01

    EUROFER weldability is investigated in support of the European TBM manufacturing. Electron beam, hybrid, laser and NGTIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel. It is shown that the most promising processes are laser, electron beam and hybrid welding, depending on the section size and accessibility. They produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The FZ are typically composed of martensite laths, with small grain sizes. In the HAZ, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. A one step PWHT (750 deg. C/3 h) is successfully applied to joints restoring good material performance. Distortion levels, with and without PWHT, are controlled through adaptation of manufacturing steps and clamping devices, obtaining levels not exceeding 120 {mu}m (+/-60 {mu}m) on a full 'one cell mock-up'.

  2. Mechanism of Secondary Hardening in Rapid Tempering of Dual-Phase Steel

    Science.gov (United States)

    Saha, Dulal Chandra; Nayak, Sashank S.; Biro, Elliot; Gerlich, Adrian P.; Zhou, Y.

    2014-12-01

    Dual-phase steel with ferrite-martensite-bainite microstructure exhibited secondary hardening in the subcritical heat affected zone during fiber laser welding. Rapid isothermal tempering conducted in a Gleeble simulator also indicated occurrence of secondary hardening at 773 K (500 °C), as confirmed by plotting the tempered hardness against the Holloman-Jaffe parameter. Isothermally tempered specimens were characterized by analytic transmission electron microscopy and high-angle annular dark-field imaging. The cementite (Fe3C) and TiC located in the bainite phase of DP steel decomposed upon rapid tempering to form needle-shaped Mo2C (aspect ratio ranging from 10 to 25) and plate-shaped M4C3 carbides giving rise to secondary hardening. Precipitation of these thermodynamically stable and coherent carbides promoted the hardening phenomenon. However, complex carbides were only seen in the tempered bainite and were not detected in the tempered martensite. The martensite phase decomposed into ferrite and spherical Fe3C, and interlath-retained austenite decomposed into ferrite and elongated carbide.

  3. Investigation of the relationships between mechanical properties and microstructure in a Fe-9%Cr ODS steel

    Directory of Open Access Journals (Sweden)

    Hary Benjamin

    2016-01-01

    Full Text Available Ferritic-martensitic Oxide Dispersion Strengthened (ODS steels are potential materials for fuel pin cladding in Sodium Fast Reactor (SFR and their optimisation is essential for future industrial applications. In this paper, a feasibility study concerning the generation of tensile specimens using a quenching dilatometer is presented. The ODS steel investigated contains 9%Cr and exhibits a phase transformation between ferrite and austenite around 870 °C. The purpose was to generate different microstructures and to evaluate their tensile properties. Specimens were machined from a cladding tube and underwent controlled heat treatments inside the dilatometer. The microstructures were observed using Electron Backscatter Diffraction (EBSD and tensile tests were performed at room temperature and at 650 °C. Results show that a tempered martensitic structure is the optimum state for tensile loading at room temperature. At 650 °C, the strengthening mechanisms that are involved differ and the microstructures exhibit more similar yield strengths. It also appeared that decarburisation during heat treatment in the dilatometer induces a decrease in the mechanical properties and heterogeneities in the dual-phase microstructure. This has been addressed by proposing a treatment with a much shorter time in the austenitic domain. Thereafter, the relaxation of macroscopic residual stresses inside the tube during the heat treatment was evaluated. They appear to decrease linearly with increasing temperature and the phase transformation has a limited effect on the relaxation.

  4. Characterization of ASTM A335 P92 steel in continuous cooling cycles

    International Nuclear Information System (INIS)

    Xaubet, M. N.; Danón, C. A.; Ramos, C. P.

    2013-01-01

    The operating conditions demanded by Generation IV reactors include high temperatures, higher radiation doses and highly corrosive environments. Among the structural materials proposed both for in-core and out-of-core applications, elevated-temperature P91/P92 ferritic/martensitic steels have been considered. This work studies the transformation behavior and microstructural evolution of ASTM A335 P92 steel in continuous cooling cycles (CCT). The material was austenized at 1050 ºC and afterwards cooled down at controlled rates (300, 50 and 15 ºC/h). The determination and characterization of the phases present in the samples was performed by optical microscopy, field emission scanning electron microscopy, Mössbauer spectroscopy and X-ray diffraction. Samples whose cooling rates were on the limits of the studied range (300 and 15 ºC/h) presented completely martensitic and completely ferritic structures, respectively. The sample cooled down at intermediate rate exhibited, though, a mixed structure of martensite and ferrite. Second-phase precipitation has also been observed in all the samples. (author)

  5. Development of ultrafine ferritic sheaves/plates in SAE 52100 steel for enhancement of strength by controlled thermomechanical processing

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, J. [Department of Metallurgical and Materials Engineering, Indian Institute of Technology, Kharagpur, W.B. 721 302 (India); Scientific Services and Research and Development, Tata Steel, Jamshedpur 831 001, Jharkhand (India); Manna, I., E-mail: imanna@metal.iitkgp.ernet.in [Department of Metallurgical and Materials Engineering, Indian Institute of Technology, Kharagpur, W.B. 721 302 (India); Central Glass and Ceramic Research Institute (CGCRI), Council of Scientific and Industrial Research (CSIR), 196 Raja S C Mullick Road, Jadavpur, Kolkata 700032 (India)

    2012-06-30

    Highlights: Black-Right-Pointing-Pointer Ultrafine bainite + martensite duplex microstructure developed in SAE 52100 steel. Black-Right-Pointing-Pointer Thermomechanical processing modifies size + morphology of bainitic ferrite. Black-Right-Pointing-Pointer Processing involves hot deformation prior to/during/after austenitizing. Black-Right-Pointing-Pointer Significant improvement in mechanical strength achieved. Black-Right-Pointing-Pointer Similar study on high carbon, low alloy steel not reported in the literature. - Abstract: The present study attempts to tailor the size, morphology and distribution of the ferrite needles/sheaves by thermomechanical processing and develop an ultrafine ferrite + martensite duplex microstructure for enhancement of strength and toughness in SAE 52100 steel. The thermo-mechanical routine included 5% hot deformation before, during or after austenitizing at 950 Degree-Sign C for 15 min followed by austempering at 270 Degree-Sign C for 30 min and subsequent water quenching to room temperature. Optical/electron microscopy along with X-ray diffraction was used to quantitatively monitor the size, morphology and distribution of the phase or phase aggregate. Significant improvement in nanohardness, wear resistance and elastic modulus and was observed in samples subjected to thermomechanical processing, as compared to that following the same austenitizing and austempering routine without hot deformation at any stage. However, improvement in the bulk mechanical property due to the present thermo-mechanical is lower than that obtained in our earlier study comprising cold deformation prior to austenitizing and austempering.

  6. Influence of the fabrication process parameters on microstructures and mechanical properties of 10Cr-1Mo ODS steel

    International Nuclear Information System (INIS)

    Jin, Hyun Ju; Kim, Ki Baik; Choi, Byoung Kwon; Kang, Suk Hoon; Noh, Sang Hoon; Kim, Ga Eon; Kim, Tae Kyu

    2016-01-01

    Oxide dispersion strengthened (ODS) FM steels have been developed as the most promising core structural material for high- temperature components operating in severe environments such as nuclear fusion and fission systems owing to its excellent elevated temperature strength and radiation resistance stemming from the addition of extremely thermally stable oxide particles dispersed in a ferritic/martensitic matrix. To realize the structural components such as plates, sheets and tubes in SFR, the development of manufacturing processes is an essential issue for the ODS FM steel. While the ODS steel has superior radiation resistance and high temperature strength, in comparison with the existing commercial steels, it is difficult for the ODS steel to obtain sufficient workability for the fabrication due to high hardness and low ductility at room temperature, meaning that the manufacturing of the ODS plate including cladding tube can be complicated by the low cold workability. In order to prevent the ODS steel from any damage during the manufacturing process, thus, the introduction of intermediate heat treatments between cold rolling processes is necessary. This study investigates effects of the fabrication process parameters such as the cold working ratio, the intermediate and final heat treatments on the microstructure and mechanical properties of 10Cr-1Mo ODS steel. In an effort to optimize the manufacturing route of the ODS FM steel, the microstructural and mechanical evolutions for the ODS plate manufactured by a control of the fabrication process parameters were evaluated in the present study. In the present study, the effect of a cold rolling and intermediate heat treatments on microstructures and mechanical properties of 10Cr-1Mo FM ODS steel were investigated. During the manufacturing route the hardness measurements remained below the critical value of 400 Hv. Intermediate heat treatment with slow cooling led to a softened ferritic structures which can be further

  7. Atom probe characterization of yttria particles in ODS Eurofer steel

    International Nuclear Information System (INIS)

    Aleev, A.A.; Zaluzhny, A.G.; Nikitin, A.A.; Rogozhkin, S.V.; Iskandarov, N.A.; Vladimirov, P.; Moeslang, A.; Lindau, R.; Klimenkov, M.

    2009-01-01

    Oxide dispersion strengthened steels exhibit higher temperature and radiation resistance than conventionally produced ferritic/martensitic steels. Such behaviour, as believed, is mainly caused by presence of highly dispersed and extremely stable oxide particles with sizes of few nanometers. It was shown that the most promising oxide additive was yttria (Y 2 O 3 ) and as mechanical parameters were strongly depended on size and number density of formed peculiarities it is required to reduce their dimensions to few nanometers and drastically increase their number. At present, considerable effort is focused on investigation of behaviour and properties of such particles. Recent studies of Eurofer ODS steel (9%-CrWVTa) by SANS and PoAS revealed the presence of high number density structural peculiarities with size approximately one nanometer. At the same time, previous studies by TEM identified only high number of small (5-10 nm) Y 2 O 3 particles. So, the purpose of this work was to look into this material by means of tomographic atom probe and find out the chemistry and origin of peculiarities with sizes less than 5 nm. These investigations revealed fine (∼ 2 nm) particles that were enriched not only in yttrium and oxygen but also in vanadium and nitrogen. Concentration of vanadium in them is approximately at the same level as yttrium. Moreover, some particles were found to be enriched in only three or even two elements mentioned above. However, total concentration of chemical elements in these particles is considerably less than that of iron. Estimated number density for detected particles is (1 / 5) x 10 23 m -3 . (author)

  8. Effect of HIP temperature and cooling rate on microstructure and hardness of joints for ODS-RAFM steels and JLF-1 steel

    International Nuclear Information System (INIS)

    Fu, Haiying; Nagasaka, Takuya; Muroga, Takeo; Kimura, Akihiko; Ukai, Shigeharu

    2016-01-01

    Dissimilar-metal joints between ODS-RAFM (oxide-dispersion-strengthened reduced activation ferritic/martensitic) steels and JLF-1 steel were fabricated by hot isostatic pressing (HIP) at 1000 - 1100degC with a cooling rate of 5degC/min. After the HIP, it was always quenched martensite for JLF-1 steel. However, coarse precipitates were found in 9Cr-ODS. Additional annealing experiments to simulate HIP conditions were conducted for 9Cr-ODS with cooling rate ranged from 0.5 to 36degC/min at 800 - 1100degC. The results showed that, to form quenched martensite for 9Cr-ODS, the HIP temperature should be above 1000degC with cooling rate no less than 25dgeC/min. When the cooling rate is increased to 36degC/min, the microstructure of 9Cr-ODS is quenched martensite with precipitate size similar as that before HIP. If the limitation of precipitate size in 9Cr-ODS is 0.2 µm, HIP temperature above 1050degC with cooling rate no less than 30degC/min is needed. In this case, post-weld heat treatment (PWHT) with only tempering is necessary to recover the microstructure of 9Cr-ODS to tempered martensite. For 12Cr-ODS, the HIP temperature and cooling rate has no effect on hardness and precipitate size. PWHT is not necessary for the single-metal joint of 12Cr-ODS from the view point of precipitation control. However, for the dissimilar-metal joints between ODS-RAFM steels and JLF-1 steel, the PWHT condition should be comprehensively determined by considering microstructural evolution of each part in the joints after HIP. (author)

  9. Study of the stability of the nanometer-sized oxides dispersed in ODS steels under ion irradiations

    International Nuclear Information System (INIS)

    Lescoat, M.-L.

    2012-01-01

    Oxide Dispersion Strengthened (ODS) Ferritic-Martensitic (FM) alloys are expected to play an important role as cladding materials in Generation IV sodium fast reactors operating in extreme temperature (400-500 C) and irradiation conditions (up to 200 dpa). Since nano-oxides give ODS steels their high temperature strength, the stability of these particles is an important issue. The present study evaluates the radiation response of nano-oxides by the use of in-situ and ex-situ ion irradiations performed on both Fe18Cr1W0,4Ti +0,3 Y 2 O 3 and Fe18Cr1W0,4Ti + 0.3 MgO ODS steels. In particular, the results showed that Y-Ti-O nano-oxides are quite stable under very high irradiation dose, namely 219 dpa at 500 C, and that the oxide interfacial structures are likely playing an important role on the behavior under irradiation (oxide stability and point defect recombination. (author) [fr

  10. Micromechanical analysis of martensite distribution on strain localization in dual phase steels by scanning electron microscopy and crystal plasticity simulation

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, M. [Department of Mechanical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of); Ziaei-Rad, S., E-mail: szrad@cc.iut.ac.ir [Department of Mechanical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of); Saeidi, N. [Department of Materials Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of); Jamshidian, M. [Department of Mechanical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of)

    2016-07-18

    The morphology and distribution of the dispersed martensite islands in the ferrite matrix plays a key role in the formation of shear bands in dual phase steels. In this study, we investigate the relationship between the martensite dispersion and the strain localization regions due to the formation of shear bands in fine-grained DP 780 steel, employing experimental observations as well as numerical simulations. SEM studies of the deformed microstructure showed that voids nucleated at ferrite-martensite interface within larger ferrite grains and regions with low local martensite fraction. The experimental results were precisely analyzed by finite element simulations based on the theory of crystal plasticity. A parametric study was then performed to obtain a deeper insight in to the effect of martensite dispersion on the strain localization of the neighboring ferrite. Crystal plasticity simulation results revealed that in a more regular structure compared to a random structure, a greater region of the ferrite phase contributes to accommodate plasticity. In addition, these regions limit the formation of main shear bands by creating barriers against stress concentration regions, results in lower growth and interaction of stress concentration regions with each others.

  11. Effect of bainitic transformation temperature on the mechanical behavior of cold-rolled TRIP steels studied with in-situ high-energy X-ray diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Fu, B. [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Yang, W.Y., E-mail: wyyang@ustb.edu.cn [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Li, L.F.; Sun, Z.Q. [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing 100083 (China)

    2014-05-01

    The effect of bainitic transformation temperature (400 and 450 °C) after intercritical annealing on the mechanical behavior of a low alloyed C–Mn–Al–Si cold-rolled TRIP steel was investigated using the in-situ high-energy X-ray diffraction technique. It was found that the mechanical behaviors of TRIP steels were dominated by the micromechanical behaviors of constituent phases, such as yield strength of each phase and stress partitioning among different phases, as well as the transformation kinetics of retained austenite during plastic deformation. The microstructures obtained at different bainitic transformation temperatures were similar, but exhibited obviously different mechanical behaviors. The retained austenite in the sample treated at 450 °C with lower carbon content and yield strength was less stable and transformed into martensite at a relatively faster speed during deformation leading to a higher ultimate tensile strength but a smaller uniform elongation. In addition, stress partitioning among constituent phases was also obtained for the investigated steels in such a way that the ferrite matrix undertook smaller stresses and the bainitic ferrite, martensite and retained austenite bore larger ones during plastic deformation. The retained austenite in the sample treated at 400 °C with higher carbon content displayed significantly higher strength and relatively stronger work-hardening capabilities during deformation in comparison to those of the sample treated at 450 °C.

  12. Fracture toughness of the F-82H steel-effect of loading modes, hydrogen, and temperature

    International Nuclear Information System (INIS)

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

    1996-01-01

    The effects of loading mode, hydrogen, and temperature on fracture toughness and tearing modulus were examined for a ferritic/martensitic steel (F-82H). The introduction of a shear load component, mode III, significantly decreased the initiation and propagation resistance of cracks compared to the opening load, mode I, behavior. Mode I crack initiation and propagation exhibited the highest resistance. A minimum resistance occurred when the mode I and mode III loads were nearly equal. The presence of 4 wppm hydrogen decreased the cracking resistance compared to behavior without H regardless of the loading mode. The minimum mixed-mode fracture toughness with the presence of hydrogen was about 30% of the hydrogen-free mode I fracture toughness. The mixed-mode toughness exhibited a lesser sensitivity to temperature than the mode I toughness. The J IC value was 284 kJ/m 2 at room temperature, but only 60 kJ/m 2 at -55 C and 30 kJ/m 2 at -90 C. The ductile to brittle transition temperature (DBTT) was apparently higher than -55 C. (orig.)

  13. Neural-network analysis of irradiation hardening in low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Kemp, R. [Department of Materials Science and Metallurgy, University of Cambridge, Pembroke Street, Cambridge CB2 3QZ, UK (United Kingdom)]. E-mail: rk237@cam.ac.uk; Cottrell, G.A. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB, UK (United Kingdom); Bhadeshia, H.K.D.H. [Department of Materials Science and Metallurgy, University of Cambridge, Pembroke Street, Cambridge CB2 3QZ, UK (United Kingdom); Odette, G.R. [Department of Mechanical and Environmental Engineering and Department of Materials, University of California Santa Barbara, Santa Barbara, CA 93106 (United States); Yamamoto, T. [Department of Mechanical and Environmental Engineering and Department of Materials, University of California Santa Barbara, Santa Barbara, CA 93106 (United States); Kishimoto, H. [Department of Mechanical and Environmental Engineering and Department of Materials, University of California Santa Barbara, Santa Barbara, CA 93106 (United States)

    2006-02-01

    An artificial neural network has been used to model the irradiation hardening of low-activation ferritic/martensitic steels. The data used to create the model span a range of displacement damage of 0-90 dpa, within a temperature range of 273-973 K and contain 1800 points. The trained model has been able to capture the non-linear dependence of yield strength on the chemical composition and irradiation parameters. The ability of the model to generalise on unseen data has been tested and regions within the input domain that are sparsely populated have been identified. These are the regions where future experiments could be focused. It is shown that this method of analysis, because of its ability to capture complex relationships between the many variables, could help in the design of maximally informative experiments on materials in future irradiation test facilities. This will accelerate the acquisition of the key missing knowledge to assist the materials choices in a future fusion power plant.

  14. Helium implanted RAFM steels studied by positron beam Doppler Broadening and Thermal Desorption Spectroscopy

    International Nuclear Information System (INIS)

    Carvalho, I; Schut, H; Fedorov, A; Luzginova, N; Desgardin, P; Sietsma, J

    2013-01-01

    Reduced Activation Ferritic/Martensitic steels are being extensively studied because of their foreseen application in fusion and Generation IV fission reactors. To mimic neutron irradiation conditions, Eurofer97 samples were implanted with helium ions at energies of 500 keV and 2 MeV and doses of 5x10 15 -10 16 He /cm 2 , creating atomic displacements in the range 0.07–0.08 dpa. The implantation induced defects were characterized by positron beam Doppler Broadening (DB) and Thermal Desorption Spectroscopy (TDS). The DB data could be fitted with one or two layers of material, depending on the He implantation energy. The S and W values obtained for the implanted regions suggest the presence of not only vacancy clusters but also positron traps of the type present in a sub-surface region found on the reference sample. The traps found in the implanted layers are expected to be He n V m clusters. For the 2 MeV, 10 16 He/cm 2 implanted sample, three temperature regions can be observed in the TDS data. Peaks below 450 K can be ascribed to He released from vacancies in the neighbourhood of the surface, the phase transition is found at 1180 K and the peak at 1350 K is likely caused by the migration of bubbles.

  15. The dose dependence of fracture toughness Of F82H steel

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States); Tanigawa, H.; Ando, M.; Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Odette, G. [UCSB, Santa-Barbara, Dept. of Mechanical Engineering UCSB, AK (United States); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States)

    2007-07-01

    Full text of publication follows: The ferritic-martensitic steel F82H is a primary candidate low-activation material for fusion applications, and it is being investigated in the joint U.S. Department of Energy-Japan Atomic Energy Agency. As a part of this program, several capsules containing fracture toughness specimens were irradiated in High-Flux Isotope Reactor. These specimens were irradiated to a wide range of doses from 3.5 to 25 dpa. The range of irradiation temperature was from 250 deg. C to 500 deg. C. This paper summarizes the changes in fracture toughness transition temperature and decrease in the ductile fracture toughness as result of various irradiation conditions. It is shown that in the 3.5 to 25 dpa dose range, irradiation temperature plays the key rote in determination of the shift of the transition temperature. Highest embrittlement observed at 250 deg.C and the lowest at 500 deg. C. At a given irradiation temperature, shift of the fracture toughness transition temperature increases slightly with dose within the studied dose range. It appears that main gain in transition temperature shift occurred during initial {approx}5 dpa of irradiation. The present data are compared to the available published trends. (authors)

  16. Tantalum-Addition Effect on Tensile and Creep Properties in 9Cr-0.5Mo-2W-V-Nb Steels

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, Sung Ho; Back, Jong Hyuk; Kim, Tae Kyu; Lee, Chan Bock

    2011-01-01

    Ferritic/martensitic steels (FMS) are being considered prospectively as cladding materials of a SFR fuel in Gen-IV nuclear systems. There are sound technical justifications for these material selections, and the adoption of the FM steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for a Gen-IV SFR. To develop an improved FM steel for the Gen-IV SFR fuel cladding in Korea, a R and D program has been progressed since 2007. Categories of materials considered in the program included 8∼12% Cr FM steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel (hereafter Gr. 92) to offset the difficulties encountered with commercial available high Cr (8∼12%) steels. Since the fuel cladding in a Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the developed cladding should thus sustain both superior irradiation and temperature stabilities during its operational life. The newly developed advanced steel should also overcome severe drawbacks: mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. The aim of this study is to investigate the effect of Ta addition on the tensile and creep properties of the three alloys which are designed, manufactured and tested. Their properties are obtained and compared for developing new FM fuel cladding materials

  17. Microstructure and mechanical properties of resistance-spot-welded joints for A5052 aluminum alloy and DP 600 steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jianbin [College of Automotive Collaborative Innovation Center, Chongqing University, No. 174, Shazheng Street, Shapingba District, Chongqing 400044 (China); Yuan, Xinjian, E-mail: xinjianyuan@yahoo.com [College of Materials Science and Engineering, Chongqing University, No. 174, Shazheng Street, Shapingba District, Chongqing 400044 (China); Hu, Zhan; Sun, Changzheng; Zhang, Yanxin; Zhang, Yuxuan [College of Materials Science and Engineering, Chongqing University, No. 174, Shazheng Street, Shapingba District, Chongqing 400044 (China)

    2016-10-15

    The microstructure and mechanical properties of resistance-spot-welded A5052 aluminum alloy and DP 600 dual-phase steel joint were studied. The fusion zone (FZ) and heat-affected zone (HAZ) of DP 600 exhibited lath martensite and ferrite-martensite structures, respectively. The microstructure of FZ and HAZ in the A5052 side was composed of cellular crystals and the boundary region of FZ exhibited a columnar crystal morphology. A Fe{sub 2}Al{sub 5} intermetallic compound (IMC) layer with 3.3 μm thickness was found adjacent to the DP 600 side, whereas a needle-shaped Fe{sub 4}Al{sub 13} IMC layer with length of 0.67 μm to 15.8 μm was found adjacent to the aluminum alloy side. The maximum tensile shear load of the A5052/DP 600 joint was 5.5 KN, with a corresponding molten nugget diameter of 6.3 mm. The fracture morphology of the optimized A5052/DP 600 joint was mainly an elongated dimple fracture accompanied by cleavage fracture. - Highlights: •A5052 and DP 600 with large gaps in properties were investigated by RSW. •The microstructures of RSW joints in DP 600/A5052 were examined detailedly. •The micro/macro-characteristics and strength relations of joints were analyzed.

  18. Investigations on the growth kinetics of Laves phase precipitates in 12% Cr creep-resistant steels: Experimental and DICTRA calculations

    Energy Technology Data Exchange (ETDEWEB)

    Prat, O. [Max Planck Institute fuer Eisenforschung GmbH, Max Planck Strasse 1, 40237 Duesseldorf (Germany)] [Universidad de Concepcion, Departamento de Ingenieria de Materiales, Edmundo Larenas 270, Concepcion (Chile); Garcia, J., E-mail: jose.garcia@helmholtz-berlin.de [Helmholtz-Zentrum Berlin fuer Materialien und Energie GmbH, Hahn-Meitner-Platz 1, 14109 Berlin (Germany); Rojas, D. [Max Planck Institute fuer Eisenforschung GmbH, Max Planck Strasse 1, 40237 Duesseldorf (Germany); Carrasco, C. [Universidad de Concepcion, Departamento de Ingenieria de Materiales, Edmundo Larenas 270, Concepcion (Chile); Inden, G. [Max Planck Institute fuer Eisenforschung GmbH, Max Planck Strasse 1, 40237 Duesseldorf (Germany)

    2010-10-15

    The growth kinetics of Laves phase precipitates (type Fe{sub 2}W) in the early stage of creep (650 deg. C for 10,000 h) in two 12% Cr ferrite-martensitic steels has been investigated. In one alloy the Laves phase formed on tempering, while in the second alloy the Laves phase precipitated during creep. Kinetic simulations were performed using the software DICTRA. The particle size of the Laves phase was measured on transmission electron microscopy samples. The equilibrium phase fraction of the Laves phase was reached in the first thousand hours. Simulations of particle growth showed good agreement with the experimental results. Competitive growth between M{sub 23}C{sub 6} and the Laves phase showed that M{sub 23}C{sub 6} carbides reached their equilibrium after 12 days, whereas the Laves phase reached equilibrium after 3 months. Simulations of the influence of the interfacial energy and addition of Co, Cu and Si on Laves phase precipitation are presented.

  19. Dry sliding wear behaviour of heat treated iron based powder metallurgy steels with 0.3% Graphite + 2% Ni additions

    International Nuclear Information System (INIS)

    Tekeli, S.; Gueral, A.

    2007-01-01

    To determine the effect of various heat treatments on the microstructure and dry sliding wear behaviour of iron based powder metallurgy (PM) steels, atomized iron powder was mixed with 0.3% graphite + 2% Ni. The mixed powders were cold pressed at 700 MPa and sintered at 1200 deg. C for 30 min under pure Ar gas atmosphere. One of the sintered specimens was quenched from 890 deg. C and then tempered at 200 deg. C for 1 h. The other sintered specimens were annealed at different intercritical heat treatment temperatures of 728 and 790 deg. C and water quenched. Through this intercritical annealing heat treatment, the specimens with various ferrite + martensite volume fractions were produced. Wear tests were carried out on the quenched + tempered and intercritically annealed specimens under dry sliding conditions using a pin-on-disk type machine at constant load and speed and the results were compared in terms of microstructure, hardness and wear strength. It was seen that hardness and wear strength in intercritically annealed specimens were higher than that of quenched + tempered specimen

  20. Microstructural evolutions of friction stir welded F82H steel for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sang Hoon; Shim, Jae Won; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Tani Gawa, Hiro Yasu [JAEA, Rokasho (Japan); Fujii, Hideto Shi [Osaka Univ., Osaka (Japan); Kim Ura, Aki Hiko [Kyoto Univ., Kyoto (Japan)

    2012-10-15

    A blanket is the most important component functionalized as plasma confining, tritium breeding, heat exchanging, and irradiation shielding from severe thermo neutron loads in a fusion reactor. Its structure consists of first walls, side walls, a back board, and coolant channels mainly made of reduced activation ferritic/martensitic (RAFM) steel, which is the most promising candidate as a structural material for fusion reactors. To fabricate this blanket structure, some welding and joining methods have being carefully applied. However, when fusion welding, such as tungsten inert gas (TIG) welding, electron beam, and laser welding was performed between F82H and itself, the strength of welds significantly deteriorated due to the development of {delta} ferrite and precipitate dissolution. Post welding heat treatment (PWHT) should be followed to restore the initial microstructure. Nevertheless, microstructural discontinuity inevitably occurs between the weld metal, heat affected zone and base metal and this seriously degrades the entire structural stability under pulsed operation at high temperature in test blanket module (TBM). A phase transformation can also be an issue to be solved, which leads to a difficult replacement of the blanket module. Therefore, a reliable and field applicable joining technique should be developed not to accompany with PWHT after the joining process. Friction stir welding (FSW) is one of the solid state processes that does not create a molten zone at the joining area, so the degradation of the featured microstructures may be avoided or minimized. In this study, FSW was employed to join F82H steels to develop a potential joining technique for RAFM steel. The microstructural features on the joint region were investigated to evaluate the applicability of the FSW.

  1. High Temperature Uniaxial Compression and Stress-Relaxation Behavior of India-Specific RAFM Steel

    Science.gov (United States)

    Shah, Naimish S.; Sunil, Saurav; Sarkar, Apu

    2018-05-01

    India-specific reduced activity ferritic martensitic steel (INRAFM), a modified 9Cr-1Mo grade, has been developed by India as its own structural material for fabrication of the Indian Test Blanket Module (TBM) to be installed in the International Thermonuclear Energy Reactor (ITER). The extensive study on mechanical and physical properties of this material has been currently going on for appraisal of this material before being put to use in the ITER. High temperature compression, stress-relaxation, and strain-rate change behavior of the INRAFM steel have been investigated. The optical microscopic and scanning electron microscopic characterizations were carried out to observe the microstructural changes that occur during uniaxial compressive deformation test. Comparable true plastic stress values at 300 °C and 500 °C and a high drop in true plastic stress at 600 °C were observed during the compression test. Stress-relaxation behaviors were investigated at 500 °C, 550 °C, and 600 °C at a strain rate of 10-3 s-1. The creep properties of the steel at different temperatures were predicted from the stress-relaxation test. The Norton's stress exponent (n) was found to decrease with the increasing temperature. Using Bird-Mukherjee-Dorn relationship, the temperature-compensated normalized strain rate vs stress was plotted. The stress exponent (n) value of 10.05 was obtained from the normalized plot. The increasing nature of the strain rate sensitivity (m) with the test temperature was found from strain-rate change test. The low plastic stability with m 0.06 was observed at 600 °C. The activation volume (V *) values were obtained in the range of 100 to 300 b3. By comparing the experimental values with the literature, the rate-controlling mechanisms at the thermally activated region of high temperature were found to be the nonconservative movement of jogged screw dislocations and thermal breaking of attractive junctions.

  2. Retention and features of deuterium detrapping from radiation-induced damages in steels

    International Nuclear Information System (INIS)

    Tolstolutskaya, G.D.; Ruzhytskiy, V.V.; Karpov, S.A.; Kopanets, I.E.

    2009-01-01

    The accelerators and ion-beam analysis techniques are used for simulation of displacement damage and detailed investigation of distribution profiles of damage and impurity gas atoms (especially helium and hydrogen) in the irradiation of targets for a wide ranges of doses and particle energies. The influence of preimplanted helium and heavy ion-induced damage on deuterium trapping in austenitic and ferritic/martensitic steels was studied. The results obtained for 18Cr10NiTi stainless steel show that ion-implanted deuterium is weakly trapped by defects produced in 5 keV D + displacement cascades. The effective trapping temperature interval is between 300 and 600 K. The characteristics of trapping and the temperature range of hydrogen isotopes retention in traps formed by prior implantation of helium depend on the concentration of implanted helium and on the type of defects developed. The formation of helium bubbles in 18Cr10NiTi steel causes an order of magnitude increase in the content of retained deuterium atoms in the range of temperature 300-600 K and extends the interval of effective trapping temperatures to 1000 K. Energetic heavy-ion irradiation (1.4 MeV Ar + ) has been used for modeling defect cluster formation under displacement cascade conditions to simulate fusion reactor environments. It was found that retention of hydrogen and deuterium strongly increased in this case. It is shown that the presence of a surface-passive film considerably shifts the gas release interval to higher temperatures and reduces the deuterium surface recombination coefficient by several orders of magnitude.

  3. An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation

    Science.gov (United States)

    Hosemann, P.; Swadener, J. G.; Kiener, D.; Was, G. S.; Maloy, S. A.; Li, N.

    2008-03-01

    The superior properties of ferritic/martensitic steels in a radiation environment (low swelling, low activation under irradiation and good corrosion resistance) make them good candidates for structural parts in future reactors and spallation sources. While it cannot substitute for true reactor experiments, irradiation by charged particles from accelerators can reduce the number of reactor experiments and support fundamental research for a better understanding of radiation effects in materials. Based on the nature of low energy accelerator experiments, only a small volume of material can be uniformly irradiated. Micro and nanoscale post irradiation tests thus have to be performed. We show here that nanoindentation and micro-compression testing on T91 and HT-9 stainless steel before and after ion irradiation are useful methods to evaluate the radiation induced hardening.

  4. Irradiation proposition of ferritic steels in a russian reactor

    International Nuclear Information System (INIS)

    Seran, J.L.; Decours, J.; Levy, L.

    1987-04-01

    Using the low temperatures of russian reactors, a sample irradiation is proposed to study mechanical properties and swelling of martensitic steels (EM10, T91, 1.4914, HT9), ferrito-martensitic (EM12) and ferritic (F17), at temperatures lower than 400 0 C [fr

  5. Science.gov (United States)

    Scapin, M.; Peroni, L.; Fichera, C.; Cambriani, A.

    2014-08-01

    High chromium ferritic/martensitic steel T91 (9% Cr, 1% Mo), on account of its radiation resistance, is a candidate material for nuclear reactor applications. Its joining by an impact method to create a cold joint is tested in the realm of scoping tests toward the safe operation of nuclear fuels, encapsulated in representative T91 materials. Hitherto, T91 mechanical characterization at high strain rates is relatively unknown, particularly, in relation to impact joining and also to nuclear accidents. In this study, the mechanical characterization of T91 steel was performed in tension by varying the strain-rate (10-3 up to 104 s-1) and temperature (20-800°C) on dog-bone specimens, using standard testing machines or Hopkinson Bar apparati. As expected, the material is both temperature and strain-rate sensitive and different sets of parameters for the Johnson-Cook strength model were extracted via a numerical inverse procedure, in order to obtain the most suitable set to be used in this field of applications.

  6. Void swelling and phase stability in different heats of cold-drawn type 1.4970 stainless steel after heavy-ion irradiation

    International Nuclear Information System (INIS)

    Vaidya, W.V.; Knoblauch, G.; Ehrlich, K.

    1982-01-01

    The present investigations were undertaken with the aim to understand, to what extent variations of the tube fabrication parameters and slight modifications in the chemical composition might influence the swelling behaviour of Type 1.4970 stainless steel. The parameters varied were: Variations in the manufacturing parameters for coldworked tubes (type and degree of drawing, solution-annealing temperature and thermomechanical treatments), and variations in minor elements (C, Ti, Mo) within the specified range of chemical composition. In addition, the Si-content and the Ti/C ratio - the so-called stabilization - were changed within a broader range. The samples were irradiated with 46 MeV-Ni-ions to 64 dpa at 575 0 C and swelling as well as austenite stability, formation of precipitates and other microstructural changes were investigated by TEM. Though the austenite was stable under irradiation with respect to ferrite/martensite-transformation, the cold-drawn alloys showed a tendency to recrystallize during irradiation and exhibited lean precipitation. With respect to swelling, the only parameter that substantially reduced it, was the high Si addition; otherwise the alloys were practically insensitive to changes in the investigated parameters. These results are discussed in terms of the radiation-induced recrystallization and the high Si-effect, both of which are found to be beneficial in reducing swelling. (orig.)

  7. Synergistic effects on dislocation loops in reduced-activation martensitic steel investigated by single and sequential hydrogen/helium ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Weiping [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Luo, Fengfeng [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Institute of Applied Physics, Jiangxi Academy of Sciences, Nanchang 330029 (China); Yu, Yanxia; Zheng, Zhongcheng; Shen, Zhenyu [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2016-10-15

    Single-beam and sequential-beam irradiations were performed to investigate the H/He synergistic effect on dislocation loops in reduced-activation ferritic/martensitic (RAFM) steels. The irradiations were carried out with 10 keV H{sup +}, 18 keV He{sup +} and 160 keV Ar{sup +}, alone and in combination at 723 K. He{sup +} single-beam irradiation induced much larger dislocation loops than that induced by both H{sup +} and Ar{sup +} single-beam irradiation. H{sup +} post-irradiation after He{sup +} irradiation further increased the size of dislocation loops, whilst He{sup +} post-irradiation or Ar{sup +} post-irradiation following H{sup +} irradiation only slightly increased the size of dislocation loops. The experiment results indicate that pre-implanted H{sup +} can drastically inhibit the growth while post-implanted H{sup +} can significantly enhance the growth of dislocation loops induced by He{sup +} irradiation. The mechanisms behind the complex synergistic phenomena between H and He and the different roles that H and He played in the growth of dislocation loops are discussed.

  8. Void swelling and phase stability in different heats of cold-drawn type 1.4970 stainless steel after heavy-ion irradiation

    International Nuclear Information System (INIS)

    Vaidya, W.V.; Knoblauch, G.; Ehrlich, K.

    1982-01-01

    The present investigations were undertaken with the aim to understand, to what extent variations of the tube fabrication parameters and slight modifications in the chemical composition might influence the swelling behavior of Type 1.4970 stainless steel. The parameters varied were: variations in the manufacturing parameters for cold-worked tubes (type and degree of drawing, solution-annealing temperature and thermomechanical treatments), and variations in minor elements (C, Ti, Mo) within the specified range of chemical composition. In addition, the Si-content and the Ti/C ratio - the so-called stabilization - were changed within a broader range. The samples were irradiated with 46 MeV-Ni-ions to 64 dpa at 575 0 C and swelling as well as austenite stability, formation of precipitates and other microstructural changes were investigated by TEM. Though the austenite was stable under irradiation with respect to ferrite/martensite-transformation, the cold-drawn alloys showed a tendency to recrystallize during irradiation and exhibited lean precipitation. With respect to swelling, the only parameter that substantially reduced it, was the high Si addition; otherwise the alloys were practically insensitive to changes in the investigated parameters. These results are discussed in terms of the radiation-induced recrystallization and the high-Si-effect, both of which are found to be beneficial in reducing swelling

  9. Effect of welding on creep damage evolution in P91B steel

    Energy Technology Data Exchange (ETDEWEB)

    Baral, J., E-mail: jayshree2k4@gmail.com [Metallurgical and Materials Engineering, Indian Institute of Kharagpur, WB 721302 (India); Swaminathan, J. [CSIR–National Metallurgical Laboratory, Jamshedpur 831007 (India); Chakrabarti, D.; Ghosh, R.N. [Metallurgical and Materials Engineering, Indian Institute of Kharagpur, WB 721302 (India)

    2017-07-15

    Study of creep behavior of base metal (without weld) and welded specimens of P91B steel over a range of temperatures (600–650 °C) and stresses (50–180 MPa) showed similar values of minimum creep-rates for both specimens at higher stress regime (>100 MPa) whilst, significantly higher creep rates in the case of welded specimens at lower stress regime. Considering that welded specimen is comprised of two distinct structural regimes, i.e. weld affected zone and base metal, a method has been proposed for estimating the material parameters describing creep behavior of those regimes. Stress–strain distribution across welded specimen predicted from finite element analysis based on material parameters revealed preferential accumulation of stress and creep strain at the interface between weld zone and base metal. This is in-line with the experimental finding that creep rupture preferentially occurs at inter-critical heat affected zone in welded specimens owing to ferrite-martensite structure with coarse Cr{sub 23}C{sub 6} particles. - Highlights: •Comparison of creep properties of welded and virgin specimens of P91B steel. •At lower stresses (<100 MPa) welded samples show higher minimum creep-rate. •Creep rupture at inter-critical heat affected zone (IC-HAZ) in welded specimens. •FEA showing accumulation of creep strain in weld/base metal interface. •Precipitate free soft ferrite matrix accumulates strain and weakens IC-HAZ.

  10. Corrosion of martensitic steels in flowing 17Li83Pb alloy

    International Nuclear Information System (INIS)

    Flament, T.; Fauvet, P.; Hocde, B.; Sannier, J.

    1988-01-01

    Corrosion of three martensitic steels - 1.4914, HT9 and T91 - in the presence of flowing 17Li83Pb is investigated in thermal convection loops Tulip entirely made of 1.4914 steel. Two 3000-hour tests were performed at maximal temperatures of respectively 450 and 475 0 C with a δT of 60 0 C and an alloy velocity of about 0.08 m.s -1 . In both tests, corrosion is characterized by an homogeneous dissolution of the steel without formation of a corrosion layer. Corrosion rate is constant and very temperature dependent: the sound-metal loss of 1.4914 steel is 22 μm. year -1 at 450 0 C and 40 μm.year -1 at 475 0 C. Behaviours of 1.4914 and HT9 steels are very similar whereas T91 steel is about 20% less corroded

  11. High temperature deformation behavior, thermal stability and irradiation performance in Grade 92 steel

    Science.gov (United States)

    Alsagabi, Sultan

    The 9Cr-2W ferritic-martensitic steel (i.e. Grade 92 steel) possesses excellent mechanical and thermophysical properties; therefore, it has been considered to suit more challenging applications where high temperature strength and creep-rupture properties are required. The high temperature deformation mechanism was investigated through a set of tensile testing at elevated temperatures. Hence, the threshold stress concept was applied to elucidate the operating high temperature deformation mechanism. It was identified as the high temperature climb of edge dislocations due to the particle-dislocation interactions and the appropriate constitutive equation was developed. In addition, the microstructural evolution at room and elevated temperatures was investigated. For instance, the microstructural evolution under loading was more pronounced and carbide precipitation showed more coarsening tendency. The growth of these carbide precipitates, by removing W and Mo from matrix, significantly deteriorates the solid solution strengthening. The MX type carbonitrides exhibited better coarsening resistance. To better understand the thermal microstructural stability, long tempering schedules up to 1000 hours was conducted at 560, 660 and 760°C after normalizing the steel. Still, the coarsening rate of M23C 6 carbides was higher than the MX-type particles. Moreover, the Laves phase particles were detected after tempering the steel for long periods before they dissolve back into the matrix at high temperature (i.e. 720°C). The influence of the tempering temperature and time was studied for Grade 92 steel via Hollomon-Jaffe parameter. Finally, the irradiation performance of Grade 92 steel was evaluated to examine the feasibility of its eventual reactor use. To that end, Grade 92 steel was irradiated with iron (Fe2+) ions to 10, 50 and 100 dpa at 30 and 500°C. Overall, the irradiated samples showed some irradiation-induced hardening which was more noticeable at 30°C. Additionally

  12. Post irradiation examination of RAF/M steels after fast reactor irradiation up to 33 dpa and < 340 C (ARBOR1). RAFM steels. Metallurgical and mechanical characterisation. Final report for TW2-TTMS-001b, D9

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, C. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). EURATOM, Inst. fuer Materialforschung, Programm Kernfusion

    2010-07-01

    In an energy generating fusion reactor structural materials will be exposed to very high dpa-levels of about 100 dpa. Due to this fact and because fast reactor irradiation facilities in Europe are not available anymore, a reactor irradiation at the State Scientific Center of the Russian Federation with its Research Institute of Atomic Reactors (SSC RIAR), Dimitrovgrad, had been performed in the fast reactor BOR 60 with an instrumented test rig. This test rig contained tensile, impact and Low Cycle Fatigue type specimens used at FZK since many years. Samples of actual Reduced Activation Ferritic/Martensitic (RAF/M) -steels (e.g. EUROFER 97) had been irradiated in this reactor at a lower temperature (< 340 C) up to a damage of 33 dpa. This irradiation campaign was called ARBOR 1. Starting in 2003 one half of these irradiated samples were post irradiation examined (PIE) by tensile testing, low cycle fatigue testing and impact testing under the ISTC Partner Contract 2781p in the hot cells of SSC RIAR. In the post irradiation instrumented impact tests a significant increase in the Ductile to Brittle Transition Temperature as an effect of irradiation has been detected. During tensile testing the strength values are increasing and the strain values reduced due to substantial irradiation hardening. The hardening rate is decreasing with increasing damage level, but it does not show saturation. The low cycle fatigue behaviour of all examined RAF/M - steels show at total strain amplitudes below 1 % an increase of number of cycles to failure, due to irradiation hardening. From these post irradiation experiments, like tensile, low cycle fatigue and impact tests, radiation induced design data, e.g. for verification of design codes, can be generated.

  13. Embrittlement of Intercritically Reheated Coarse Grain Heat-Affected Zone of ASTM4130 Steel

    Science.gov (United States)

    Li, Liying; Han, Tao; Han, Bin

    2018-04-01

    In this investigation, a thermal welding simulation technique was used to investigate the microstructures and mechanical properties of the intercritically reheated coarse grain heat-affected zone (IR CGHAZ) of ASTM4130 steel. The effect of post weld heat treatment (PWHT) on the toughness of IR CGHAZ was also analyzed. The toughness of IR CGHAZ was measured by means of Charpy impact, and it is found that IR CGHAZ has the lowest toughness which is much lower than that of the base metal regardless of whether PWHT is applied or not. The as-welded IR CGHAZ is mainly composed of ferrite, martensite, and many blocky M-A constituents distributing along grain boundaries and subgrain boundaries in a near-connected network. Also, the prior austenite grains are still as coarse as those in the coarse grain heat-affected zone (CGHAZ). The presence of the blocky M-A constituents and the coarsened austenite grains result in the toughness deterioration of the as-welded IR CGHAZ. Most of the blocky M-A constituents are decomposed to granular bainite due to the effect of the PWHT. However, PWHT cannot refine the prior austenite grains. Thus, the low toughness of IR CGHAZ after PWHT can be attributed to two factors, i.e., the coarsened austenite grains, and the presence of the remaining M-A constituents and granular bainite, which are located at grain boundaries and subgrain boundaries in a near-connected network. The absorbed energy of the IR CGHAZ was increased by about 3.75 times, which means that the PWHT can effectively improve the toughness but it cannot be recovered to the level of base metal.

  14. Materials design data for reduced activation martensitic steel type F82H

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. E-mail: tavassoli@cea.fr; Rensman, J.-W.; Schirra, M.; Shiba, K

    2002-11-01

    This paper presents materials data for design of ITER test blanket modules with the reduced activation ferritic martensitic steel type F82H as structural material. From the physical properties databases, variations of modulus of elasticity, density, thermal conductivity, thermal diffusivity, specific heat, mean and instantaneous linear coefficients of thermal expansion versus temperature are derived. Also reported are Poisson's ratio and magnetic properties. From the tension test results, nominal and minimum stress values of S{sub y} and S{sub u} are derived and used for calculation of allowable primary membrane stress intensity S{sub m}. Likewise, uniform and total elongations, as well as reduction of area data, are used for calculation of minimum and true ductility at rupture values. From the instrumented Charpy impact and fracture toughness test data, ductile to brittle transition temperature, toughness and behavior of material in different fracture modes are evaluated. The effect of specimen size and geometry are discussed but preference is given to standard size specimens. From the fatigue data, total strain range versus number of cycles to failure curves are plotted and used to derive fatigue design curves, using a reduction factor of 2 on strain and a reduction factor of 20 on number of cycles to failure. Cyclic hardening curves are also derived and compared with monotonic hardening curves. From the creep data, time dependent allowable stresses S{sub r} and S{sub t} are calculated. Combination of tension and creep results are used to deduce S{sub mt} and isochronus curves. Finally, irradiated and aged materials data are compared to insure that the safety margins incorporated in unirradiated design limits are not exceeded.

  15. On the way to high resolution TEM characterization of dual ion beam irradiated ODS steels

    International Nuclear Information System (INIS)

    Hsiung, L.; Tumey, S.; Fluss, M. J.; King, W.; Marian, J.; Kuntz, J.; Dasher, B. El; Serruys, Y.; Willaime, F.; Kimura, A.

    2009-01-01

    Fission and fusion energy application of ODS steels while appearing promising requires that many key science issues be resolved. Among these issues are our incomplete understanding of the effect of irradiation on low-temperature fracture properties, the role of fusion relevant helium and hydrogen transmutation gases on the deformation and fracture of irradiated material at low and high temperatures, radiation-induced solute segregation and phase stability, mechanisms of swelling suppression in ODS steels, and the effects of radiation damage on localized deformation. While planning to focus on all these issues we are particularly interested in the atomic scale mechanism by which helium is mitigated by the nano scale particles. In order to obtain insight we are performing analytical transmission electron microscopy (AEM), high resolution electron microscopy (HRTEM) to investigate micro-structural and micro-compositional changes and property alterations of Fe-Cr ferritic/martensitic and ODS steels driven by temperature and ion-beam irradiation with Fe, H, and He. As a beginning to a collaboration between LLNL and CEA-Saclay, we have carried out an irradiation of four specimens, Fe, Fe14%Cr, and two ODS steels (14% Cr and 16% Cr) using the dual beam facility at CEA-Saclay (JANNuS). An Fe 8+ beam was implanted at 24 MeV and helium was implanted through a degrader wheel with energies between 1.7 MeV and 1.3 MeV. The nominal radiation parameters were 40 to 25 DPA, 10 to 25 appm He/DPA ratio, and specimen temperatures of ∼425 deg. C. Our goal is to compare the evolved microstructure with respect to the accumulation of helium at or near the particle matrix interface. Preparatory to this first study we have made many hi-resolution analyses of the nano-particles in the two ODS steels which serve as a base line for comparison with the TEM post irradiation examination reported here. These base line studies are reported separately at this conference. (author)

  16. Formation of Al{sub 2}O{sub 3}/FeAl coatings on a 9Cr-1Mo steel, and corrosion evaluation in flowing Pb-17Li loop

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, Sanjib, E-mail: sanjib@barc.gov.in [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Paul, Bhaskar [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Chakraborty, Poulami [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Kishor, Jugal; Kain, Vivekanand [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Dey, Gautam Kumar [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Materials Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2017-04-01

    Iron aluminide coating layers were formed on a ferritic martensitic grade 9Cr-1Mo (P 91) steel using pack aluminizing process. The formation of different aluminide compositions such as orthorhombic-Fe{sub 2}Al{sub 5}, B2-FeAl and A2-Fe(Al) on the pack chemistry and heat treatment conditions have been established. About 4–6 μm thick Al{sub 2}O{sub 3} scale was formed on the FeAl phase by controlled heat treatment. The corrosion tests were conducted using both the FeAl and Al{sub 2}O{sub 3}/FeAl coated specimens in an electro-magnetic pump driven Pb-17Li Loop at 500 °C for 5000 h maintaining a flow velocity of 1.5 m/s. The detailed characterization studies using scanning electron microscopy, back-scattered electron imaging and energy dispersive spectrometry revealed no deterioration of the coating layers after the corrosion tests. Self-healing oxides were formed at the cracks generated in the aluminide layers during thermal cycling and protected the base alloy (steel) from any kind of elemental dissolution or microstructural degradation. - Highlights: •Al{sub 2}O{sub 3}/FeAl coating produced on P91 steel by pack aluminizing and heat treatment. •Corrosion tests of coated steel conducted in flowing Pb-17Li loop at 500 °C for 5000 h. •Coating was protective against molten metal corrosion during prolonged exposure. •Self-healing protective oxides formed in the cracks generated in aluminide layers.

  17. Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

    International Nuclear Information System (INIS)

    Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

    2015-01-01

    Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

  18. Effects of helium on ductile brittle transition behavior of reduced activation ferritic steels after high concentration he implantation at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, A.; Ejiri, M.; Nogami, S.; Ishiga, M.; Abe, K. [Tohoku Univ., Dept. of Quantum Science and Energy Engr, Sendai (Japan); Kasada, R.; Kimura, A. [Kyoto Univ., Institute of Advanced Energy (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

    2007-07-01

    Full text of publication follows: Influence of Helium (He) on fracture behavior of reduced activation ferritic/martensitic steels including Oxide Dispersion Strengthening (ODS) steels and F82H were examined. To study the He effects on fracture behavior of these steels after He bubble formation conditions, higher concentration of He implantation at around 550 C were performed and examined the relationship between microstructure evolution and fracture behavior of the steels. The 1.5CVN mini size Charpy specimens were used to evaluate impact test behavior. Reduced activation ferritic ODS steels, 9Cr-ODS and 12Cr-ODS steels were examine. F82H was also examined as reference material. Helium implantation was performed by a cyclotron of Tohoku University with a beam of 50 MeV {alpha}-particles at temperature around 550 C. A tandem-type energy degrader system was used to implant He into the specimen from the irradiated surface to the range of 50 MeV {alpha}-particles, that was about 380 {mu}m in iron. Implanted He concentration were about 1000 appm. Charpy impact test was performed using a instrumented impact test apparatus in Oarai branch of IMR, Tohoku University. Analyses of absorbed energy change and fracture surface were carried out. Vickers hardness test was also carried out on He implanted area of the 1.5CVN specimen to estimate irradiation hardening. Microstructural observation was performed by TEM. In the case of F82H, DBTT increased by the 1000 appm He implantation condition was about 80 C and grain boundary fracture surface was only observed in the He implanted area of all the ruptured specimens in brittle manner. On the other hand, DBTT shift and fracture mode change of He implanted 9Cr-ODS steel was not observed after He implantation. Microstructural observation showed that He bubble formation on the lath boundaries and grain boundaries were significant in F82H, but the bubble segregation on grain boundary in ODS steel was not apparent. The bubble formation

  19. Development of IN-RAFM steel and fabrication technologies for Indian TBM

    Energy Technology Data Exchange (ETDEWEB)

    Albert, S.K., E-mail: shaju@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603 102 (India); Laha, K.; Bhaduri, A.K.; Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam 603 102 (India); Rajendrakumar, E. [TBM Division, Institute of Plasma Research, Bhat, Gandhinagar 382428 (India)

    2016-11-01

    Highlights: • India specific RAFM steel IN-RAFM has been produced in commercial scale and physical property data for the same have been generated. • Mechanical properties of the commercial heats are comparable to that of the laboratory heats. • IN-RAFM steel filler wires has been produced indigenously. • Process parameters have been optimized so that toughness of the fusion zone of the EB and laser welds is comparable to that of the base metal even at sub-zero temperatures. • Fabrication of the mock ups of different components of Indian TBM has been initiated. - Abstract: India has indigenously developed its own Indian Reduced Activation Ferritic Martensitic Steel (IN-RAFM steel) which, it is proposed to use for fabrication of Indian TBM to be installed in ITER. With an objective of qualifying this material for use in ITER, extensive testing of this material to generate the necessary data required to offer this material for Particular Material Appraisal of ITER has been undertaken. Accordingly, data generation on creep, fatigue and physical properties of this steel is in progress. Physical properties like thermal diffusivity, heat capacity and thermal expansion coefficient for this steel at various temperatures have been measured and thermal conductivity estimated from the measured physical properties. Values obtained for these physical properties are comparable with those reported for other RAFM steels. Results of creep tests that are in progress also indicate that creep is comparable to that of other RAFM steels. Low cycle fatigue (LCF) behaviour of this steel has been studied in air with total strain control in the strain amplitude range of ±0.25–1% and temperatures range of 300–873 K at a constant strain rate of 3 × 10{sup −3} s{sup −1}. Results confirm that fatigue life of IN-RAFM steel is at par with that of the European Union grade Eurofer 97 RAFM steel. During production of IN-RAFM steel plates of different thicknesses, it is observed

  20. Assessment of the influence of surface finishing and weld joints on the corrosion/oxidation behaviour of stainless steels in lead bismuth eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Munoz, F.J., E-mail: fco.javier@ciemat.es [Structural Materials Division, CIEMAT, Building 30, Avda. Complutense 22, Madrid 28040 (Spain); Soler-Crespo, L.; Gomez-Briceno, D. [Structural Materials Division, CIEMAT, Building 30, Avda. Complutense 22, Madrid 28040 (Spain)

    2011-09-01

    The objective of this paper is to gain some insight into the influence of the surface finishing in the oxidation/corrosion behaviour of 316L and T91 steels in lead bismuth eutectic (LBE). Specimens of both materials with different surface states were prepared (as-received, grinded, grinded and polished, and electrolitically polished) and oxidation tests were carried out at 775 and 825 K from 100 to 2000 h for two different oxygen concentrations and for H{sub 2}/H{sub 2}O molar ratios of 3 and 0.03. The general conclusion for these tests is that the effect of surface finishing on the corrosion/protection processes is not significant under the tested conditions. In addition the behaviour of weld joints, T91-T91 Tungsten Inert Gas (TIG) and T91-316L have been also studied under similar conditions. The conclusions are that, whereas T91-T91 welded joint shows the same corrosion properties as the parent materials for the conditions tested, AISI 316L-T91 welded joint, present an important dissolution over seam area that it associated to the electrode 309S used for the fabrication process.

  1. Assessment of the influence of surface finishing and weld joints on the corrosion/oxidation behaviour of stainless steels in lead bismuth eutectic

    Science.gov (United States)

    Martín-Muñoz, F. J.; Soler-Crespo, L.; Gómez-Briceño, D.

    2011-09-01

    The objective of this paper is to gain some insight into the influence of the surface finishing in the oxidation/corrosion behaviour of 316L and T91 steels in lead bismuth eutectic (LBE). Specimens of both materials with different surface states were prepared (as-received, grinded, grinded and polished, and electrolitically polished) and oxidation tests were carried out at 775 and 825 K from 100 to 2000 h for two different oxygen concentrations and for H 2/H 2O molar ratios of 3 and 0.03. The general conclusion for these tests is that the effect of surface finishing on the corrosion/protection processes is not significant under the tested conditions. In addition the behaviour of weld joints, T91-T91 Tungsten Inert Gas (TIG) and T91-316L have been also studied under similar conditions. The conclusions are that, whereas T91-T91 welded joint shows the same corrosion properties as the parent materials for the conditions tested, AISI 316L-T91 welded joint, present an important dissolution over seam area that it associated to the electrode 309S used for the fabrication process.

  2. Assessment of the influence of surface finishing and weld joints on the corrosion/oxidation behaviour of stainless steels in lead bismuth eutectic

    International Nuclear Information System (INIS)

    Martin-Munoz, F.J.; Soler-Crespo, L.; Gomez-Briceno, D.

    2011-01-01

    The objective of this paper is to gain some insight into the influence of the surface finishing in the oxidation/corrosion behaviour of 316L and T91 steels in lead bismuth eutectic (LBE). Specimens of both materials with different surface states were prepared (as-received, grinded, grinded and polished, and electrolitically polished) and oxidation tests were carried out at 775 and 825 K from 100 to 2000 h for two different oxygen concentrations and for H 2 /H 2 O molar ratios of 3 and 0.03. The general conclusion for these tests is that the effect of surface finishing on the corrosion/protection processes is not significant under the tested conditions. In addition the behaviour of weld joints, T91-T91 Tungsten Inert Gas (TIG) and T91-316L have been also studied under similar conditions. The conclusions are that, whereas T91-T91 welded joint shows the same corrosion properties as the parent materials for the conditions tested, AISI 316L-T91 welded joint, present an important dissolution over seam area that it associated to the electrode 309S used for the fabrication process.

  3. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    International Nuclear Information System (INIS)

    Srinivasan, G.; Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K.

    2011-01-01

    Highlights: → Weld microstructure produced by RAFMS filler wires are free from delta ferrite. → Cooling rates of by weld thermal cycles influences the presence of delta ferrite. → Weld parameters modified with higher pre heat temperature and high heat input. → PWHT optimized based on correlation of hardness between base and weld metals. → Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in

  4. Effect of mechanical alloying and compaction parameters on the mechanical properties and microstructure of EUROFER 97 ODS steel

    International Nuclear Information System (INIS)

    Ramar, A.; Oksiuta, Z.; Baluc, N.; Schaeublin, R.

    2006-01-01

    Oxide dispersion strengthened (ODS) ferritic / martensitic (F/M) steels appear to be promising candidates for the future fusion reactor. Their inherent properties, good thermal conductivity, swelling resistance and low radiation damage accumulation, deriving from the base material EUROFER 97, are further enhanced by the presence of the fine dispersion of oxide particles. They would allow in principle for a higher operating temperature of the fusion reactor, which improves its thermal efficiency. In effect, their strength remains higher than the base material with increasing temperature. Their creep properties are also improved relatively to the base material. It is the pinning of dislocations at dispersed oxide particles that helps to improve the high temperature mechanical properties. EUROFER97 is a reduced activation F/M steel, whose chemical composition is 8.9 wt. % Cr, 1.1 wt. % W, 0.47 wt. % Mn, 0.2 wt. % V, 0.14 wt. % Ta and 0.11 wt. % C and Fe for the balance. A new ODS F/M steel based on EUROFER 97 is developed with the strengthening material as Y 2 O 3 maintained at 0.3wt% based on our past experience. The ODS powder is produced by a different powder metallurgy route. The Eurofer 97 atomized powder with particle sizes around 45 μm is ball milled in argon atmosphere in a planetary ball mill together with Yttria particles with sizes about 10 to 30 nm. The milled powders are now canned in a steel container. They are degassed at 450 o C for 3 hours under a vacuum of 10-5 mbar. The canned sample is sealed in vacuum and finally compacted by hot isostatic pressing (HIP) in argon atmosphere under a pressure of 180 MPa at 1000 o C for 1 hour. Electron microscopy and X-ray diffraction observations are done at regular intervals during ball milling to identify changes in the particle and crystallite size and in particular with the solubility of Yttria in the matrix. Further, The microstructure and mechanical properties of final compacted material is assessed. The

  5. Influence of surface roughness and melt superheat on HDA process to form a tritium permeation barrier on RAFM steel

    Energy Technology Data Exchange (ETDEWEB)

    Purushothaman, J. [B.S. Abdur Rahman University, Chennai 600048 (India); MTD, MMG, IGCAR, Kalpakkam 603102 (India); Ramaseshan, R., E-mail: seshan@igcar.gov.in [TFCS, SND, MSG, IGCAR, Kalpakkam 603102 (India); Albert, S.K. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Rajendran, R. [B.S. Abdur Rahman University, Chennai 600048 (India); Gowrishankar, N. [IP Rings Ltd., Maraimalainagar, Chennai 603209 (India); Ramasubbu, V. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Murugesan, S.; Dasgupta, Arup [PMG, MMG, IGCAR, Kalpakkam 603102 (India); Jayakumar, T. [MTD, MMG, IGCAR, Kalpakkam 603102 (India)

    2015-12-15

    Highlights: • Surface modified RAFMS samples were subjected to HDA and thermal oxidation. • Sample modified by SB process showed better coating and interface morphology. • Aluminized samples at 740 °C for 2 min showed Fe{sub 2}Al{sub 9}Si{sub 2} intermetallic phase. • Oxidized samples showed Fe{sub 2}Al{sub 8}Si, Fe{sub 2}Al{sub 3}Si{sub 3} and Fe{sub 3}Al{sub 2}Si{sub 3} intermetallic phases. • A uniform permeation barrier Al{sub 2}O{sub 3} was formed on the coating of oxidized HDA samples. - Abstract: The most optimal candidate material for fabrication of Test Blanket Module (TBM) in the installation of ITER and future fusion reactors is Reduced Activation Ferritic Martensitic (RAFM) steel, yet one of the major challenges that need to be addressed with RAFM is minimizing the loss of tritium in a reactor environment through the formation of tritium permeation barrier. One of the most promising methods for the tritium permeation barrier is through duplex coating with Al{sub 2}O{sub 3}/Fe–Al which is well known to reduce tritium permeation rate by several orders of magnitude. The present work aims to form an alumina layer on RAFM steel by a two-step method, which consists of (i) Hot Dip Aluminizing (HDA) and (ii) conversion of Al into alumina by a subsequent oxidation process. In addition, the influence of surface roughness of the substrate, superheat condition of the Al alloy melt and its composition on microstructural properties of coating before and after oxidation were investigated using OM, SEM–EDS, XRD, indentation micro hardness and scratch test. The experimental results confirmed the formation of alumina layer on RAFM steel after the HDA and oxidation process. Moreover, the surface roughness of the substrate, melt superheat of Al alloy and its composition are found to have a significant influence on the microstructure, thickness, micro-hardness, nature of intermetallic compounds formed and adhesion strength of the coating.

  6. Corrosion mechanism of T91 steel by Pb-Bi eutectic used as spallation target:importance for accelerator driven system; Caracterisation de l'endommagement de parois de galeries par tomographie electrique: application en laboratoire souterrain

    Energy Technology Data Exchange (ETDEWEB)

    Grislin-Mouezy, A

    2005-07-01

    The electrical tomography monitoring is one of the basic technique used in applied geophysics. This method allows to obtain the electrical resistivity distribution from the electrical potential measurements. The excavation of a new gallery in the underground rock laboratory at Mont Terri offers the possibility to follow and characterize the damaged zone in a spatial and temporal way. Successive acquired data sets have been carried out during several months and the results have been compared with the geological observations and the studies of the stress field. These results show that the resistivity changes are correlated with the local tectonics, the bedding and the stress field near the barriers. On account of the cylindrical geometry of the gallery, a modelling program has been developed in cylindrical co-ordinates. A program of inversion by simulated annealing has been developed too and tested on synthetical data sets. (O.M.)

  7. Development of filler wires for welding of reduced activation ferritic martenstic steel for India's test blanket module of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, G., E-mail: gsrini@igcar.gov.in [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India); Arivazhagan, B.; Albert, S.K.; Bhaduri, A.K. [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamilnadu (India)

    2011-06-15

    Highlights: > Weld microstructure produced by RAFMS filler wires are free from delta ferrite. > Cooling rates of by weld thermal cycles influences the presence of delta ferrite. > Weld parameters modified with higher pre heat temperature and high heat input. > PWHT optimized based on correlation of hardness between base and weld metals. > Optimised mechanical properties achieved by proper tempering of the martensite. - Abstract: Indigenous development of reduced activation ferritic martensitic steel (RAFMS) has become mandatory to India to participate in the International Thermo-nuclear Experimental Reactor (ITER) programme. Optimisation of RAFMS is in an advanced stage for the fabrication of test blanket module (TBM) components. Simultaneously, development of RAFMS filler wires has been undertaken since there is no commercial filler wires are available for fabrication of components using RAFMS. Purpose of this study is to develop filler wires that can be directly used for both tungsten inert gas welding (TIG) and narrow gap tungsten inert gas welding (NG-TIG), which reduces the deposited weld metal volume and heat affected zone (HAZ) width. Further, the filler wires would also be used for hybrid laser welding for thick section joints. In view of meeting all the requirements, a detailed specification was prepared for the development of filler wires for welding of RAFM steel. Meanwhile, autogenous welding trials have been carried out on 2.5 mm thick plates of the RAFM steel using TIG process at various heat inputs with a preheat temperature of 250 deg. C followed by various post weld heat treatments (PWHT). The microstructure of the weld metal in most of the cases showed the presence of some delta-ferrite. Filler wires as per specifications have also been developed with minor variations on the chemistry against the specified values. Welding parameters and PWHT parameters were optimised to qualify the filler wires without the presence of delta-ferrite in the weld

  8. Experimental investigation of high He/dpa microstructural effects in neutron irradiated B-alloyed Eurofer97 steel by means of small angle neutron scattering (SANS and electron microscopy

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2016-12-01

    Full Text Available High He/dpa microstructural effects have been investigated, by means of small-angle neutron scattering (SANS and transmission electron microscopy (TEM, in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2V, 1.08W wt%, B contents variable between 10 and 1000ppm, neutron irradiated at the High Flux Reactor of the JRC-Petten at temperatures between 250 °C and 450 °C, up do a dose level of 16 dpa. Under these irradiation parameters, B activation is expected to produce corresponding helium contents variable between 80 and 5600appm, with helium bubble distributions relevant for the technological applications. The SANS measurements were carried out under magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. Increasing the estimated helium content from 400 to 5600appm, the analysis of the SANS cross-sections yields an increase in the volume fraction, attributed to helium bubbles, of almost one order of magnitude (from 0.007 to 0.038; furthermore, the difference between nuclear and magnetic SANS components is strongly reduced. These results are discussed in correlation with TEM observations of the same samples and are tentatively attributed to the effect of drastic microstructural changes in Eurofer97-1 for high He/dpa ratio values, possibly relating to the dissolution of large B-carbides due to transmutation reactions.

  9. Control of molten salt corrosion of reduced activation steel for fusion applications by metallic beryllium

    International Nuclear Information System (INIS)

    Calderoni, P.; Sharpe, P.; Nishimura, H.; Terai, T.

    2007-01-01

    Full text of publication follows: In 2001 the INL started a research program as a part of the 2. Japan/US Program on Irradiation Tests for Fusion Energy Research (JUPITER-II collaboration) aimed at the characterization of the 2LiF-BeF2 (Flibe) molten salt as a breeder and coolant material for fusion applications. A key objective of the work was to demonstrate chemical compatibility between Flibe and potential fusion structural materials once suitable fluoride potential control methods are established. A series of tests performed at INL demonstrated that this can be achieved by contacting the salt with metallic beryllium, and the results have been published in recent years. A further step was to expose two specimens of low activation ferritic/martensitic steel 9Cr-2W JLF-1 to static corrosion tests that include an active corrosion agent (hydrofluoric gas) and fluoride potential control (metallic Be) at 530 C, and the results of the tests are presented in this paper. The specimen and a beryllium rod were simultaneously immersed in the molten salt through gas tight fittings mounted on risers extending from the top lid of the test vessel; the beryllium rod was extracted after 5 hours, while the sample was left in the salt for 250 hours during which salt samples were withdrawn from the melt at fixed intervals. A diagnostic system based on the measurement of reacting HF through on-line titration was coupled with the analysis of metallic components in the salt samples that were dissolved and analyzed using inductively coupled plasma atomic emission spectroscopy (ICP-AES). Impurity levels of oxygen, nitrogen and carbon were determined from pieces of the solidified melt using Leco analytical systems. The results confirmed the expected correlation of the HF recovery with the concentration of metallic elements dissolved in the salt because of specimen corrosion. The metals concentration falls below the detectable limit when the beryllium rod is inserted and increases when the

  10. Investigation on the suitability of plasma sprayed Fe-Cr-Al coatings as tritium permeation barrier

    International Nuclear Information System (INIS)

    Fazio, C.; Serra, E.; Benamati, G.

    1999-01-01

    Results on the fabrication of a tritium permeation barrier by spraying Fe-Cr-Al powders are described. The sprayed coatings were deposited at temperatures below the A c1 temperature of the ferritic-martensitic steel substrate and no post-deposition heat treatment was applied. The aim of the investigation was the determination of the efficiency of the coatings to act as tritium permeation barrier. Metallurgical investigations as well as hydrogen isotope permeation measurements were carried out onto the produced coatings. The depositions were performed on ferritic-martensitic steels by means of three types of spray techniques: high velocity oxy fuel, air plasma spray and vacuum plasma spray. (orig.)

  11. Corrosion behavior of Si-enriched steels for nuclear applications in liquid lead–bismuth

    Energy Technology Data Exchange (ETDEWEB)

    Kurata, Yuji, E-mail: kurata.yuji@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken 319-1195 (Japan)

    2013-06-15

    Highlights: ► The corrosion behavior of Si-enriched steels in liquid lead–bismuth was studied. ► The corrosion tests were conducted at the two controlled oxygen levels. ► The Si addition reduces the scale thickness under the high oxygen condition. ► The Si addition has no significant effect under the low oxygen condition. -- Abstract: The corrosion behavior of Si-enriched steels in liquid lead–bismuth was studied in order to develop accelerator driven systems for transmutation of long-lived radioactive wastes and lead–bismuth cooled fast reactors. The corrosion tests of 316SS, Si-enriched 316SS, Mod.9Cr–1Mo steel (T91) and Si-enriched T91 were conducted at 550 °C in liquid lead–bismuth at the two controlled oxygen levels. Both the additions of 2.5 wt.% Si to 316SS and 1.5 wt.% Si to T91 had the effect of reducing the thickness of oxide layer in liquid lead–bismuth at the high oxygen concentration (2.5 × 10{sup −5} wt.%). Although the Si addition to 316SS reduced the depth of ferritization caused by Ni dissolution in liquid lead–bismuth at the low oxygen concentration (4.4 × 10{sup −8} wt.%), it could not suppress the ferritization and the penetration of Pb and Bi completely. The Si addition to T91 did not have the effect of preventing the penetration of Pb and Bi in the liquid lead–bismuth at the low oxygen concentration. The oxide scales formed on both Si-enriched steels did not have sufficient corrosion resistance under the low oxygen condition.

  12. Irradiation and inhomogeneity effects on ductility and toughness of (ODS)-7 -13Cr steels

    International Nuclear Information System (INIS)

    Preininger, D.

    2007-01-01

    Full text of publication follows: The superimposed effect of irradiation defect and structural inhomogeneity formation on tensile ductility and dynamic toughness of ferritic-martensitic 7-13CrW(Mo)VTa(Nb) and oxide dispersion-strengthened (ODS)-7-13CrWVTa(Ti)- RAFM steels has been examined by work hardening and local stress/strain-induced ductile fracture models. Structural inhomogeneities which strongly promoting plastic instability and localized flow might be formed by the applied fabrication process, high dose irradiation and additionally further during deformation by enhanced local dislocation generation around fine particles or due to slip band formation with localized heating at high impact strain rates ε'. The work hardening model takes into account superimposed dislocation multiplication from stored dislocations, dispersions and also grain boundaries as well as annihilation by cross-slip. Analytical relations have been deduced from the model describing uniform ductility and ductile upper shelf energy (USE) observed from Charpy-impact testes. Especially, the influence of different irradiation defects like atomic clusters, dislocation loops and coherent chromium-rich α'- precipitates have been considered together with effects from strain rate as well as irradiation (TI) and test temperature TT. Strengthening by clusters and more pronounced by dislocation loops formed at higher TI>250 deg. C reduces uniform ductility and also distinctly stronger dynamic toughness USE. A superimposed hardening by the α'- formation in higher Cr containing 9-13Cr steels strongly reduces toughness assisted by a combined grain-boundary embrittlement with reduction of the ductile fracture stress. But that improves work hardening and uniform ductility as observed particularly due to nano-scale Y 2 O 3 - dispersions in ODS-RAFM steels. For ODS- steels additionally the strength-induced reduction of toughness is diminished by a combined microstructural-induced increase of the ductile

  13. Materials for innovative lead alloy cooled nuclear systems: Overview

    International Nuclear Information System (INIS)

    Mueller, Georg; Weisenburger, Alfons; Fetzer, Renate; Heinzel, Annette; Jianu, Adrian

    2015-01-01

    One of the most challenging issues for all future innovative nuclear systems including Gen IV reactors are materials. The selection of the structural materials determines the design which has to consider the properties and the availability of the materials. Beside general requirements for material properties that are common for all fast reactor types specific issues arise from coolant compatibility. The high solubility of steel alloying elements in liquid Pb-alloys at reactor relevant temperatures is clearly detrimental. Therefore, all steels that are considered as structural materials have to be protected by dissolution barriers. The most common barriers for steels under consideration are oxide scales that form in situ during operation. However, increasing the temperature above 500 deg. C will result either in dissolution attack or in enhanced oxidation. For higher temperatures additional barriers like alumina forming surface alloys are discussed and investigated. Mechanical loads like creep stress and fretting will act on the steels. These mechanical loads will interact with the coolant and can increase the negative effects. For a LFR (Lead Fast Reactor) Demonstrator and MYHRRA (ADS) austenitic steels (316L) are selected for most in core components. The 15-15Ti is the choice for the fuel cladding of MYHRRA and a Pb cooled demonstrator. For an industrial LFR (Lead Fast Reactor) the ferritic martensitic steel T91 was selected as fuel clad material due to its improved irradiation resistance. T91 is in both designs the material to be used for the heat exchanger. Surface alloying with alumina forming alloys is considered to assure material functionality at higher temperatures and is therefore selected for fuel cladding of the ELFR and the heat exchanger tubes. This presentation will give an overview on the selected materials for innovative Pb alloy cooled nuclear systems considering, beside pure compatibility, the influence of mechanical interaction like creep and

  14. Radiation effects in structural materials of spallation targets

    Science.gov (United States)

    Jung, P.

    2002-02-01

    Effects of radiation damage by protons and neutrons in structural materials of spallation neutron sources are reviewed. Effects of atomic displacements, defect mobility and transmutation products, especially hydrogen and helium, on physical and mechanical properties are discussed. The most promising candidate materials (austenitic stainless steels, ferritic/martensitic steels and refractory alloys) are compared, and needed investigations are identified.

  15. Cooperation on impingement wastage experiment of Mod. 9Cr-1Mo steel using SWAT-1R sodium-water reaction test facility

    International Nuclear Information System (INIS)

    Beauchamp, F.; Allou, A.; Nishimura, M.; Umeda, R.

    2013-01-01

    Conclusion: • 6 experiments were carried out in the SWAT-1R facility of JAEA Oarai R&D Center to study the wastage resistance of the Mod. 9Cr-1Mo steel (T91) straight tubes. • These experiments were performed under the cooperation between CEA and JAEA. • The experiments were conducted successfully: - all the tubes were punctured by the reaction jet, - wastage and steam/water leak rates were obtained, - experimental results brought some new determining sets of wastage data on T91. • This fruitful cooperation has contributed to: - expanding the wastage database on T91, - upgrading wastage rates prediction from modelling, - the safety demonstration of future steam generators units

  16. Opportunities for the LWR ATF materials development program to contribute to the LBE-cooled ADS materials qualification program

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Xing, E-mail: gongxingzfl@hotmail.com [Department of ATF R& D, Nuclear Fuel Research and Development Center, China Nuclear Power Technology Research Institute Co., Ltd., China General Nuclear Power Corporation (CGN), Shenzhen, 518026 (China); Li, Rui, E-mail: li-rui@cgnpc.com.cn [Department of ATF R& D, Nuclear Fuel Research and Development Center, China Nuclear Power Technology Research Institute Co., Ltd., China General Nuclear Power Corporation (CGN), Shenzhen, 518026 (China); Sun, Maozhou; Ren, Qisen [Department of ATF R& D, Nuclear Fuel Research and Development Center, China Nuclear Power Technology Research Institute Co., Ltd., China General Nuclear Power Corporation (CGN), Shenzhen, 518026 (China); Liu, Tong, E-mail: liutong@cgnpc.com.cn [Department of ATF R& D, Nuclear Fuel Research and Development Center, China Nuclear Power Technology Research Institute Co., Ltd., China General Nuclear Power Corporation (CGN), Shenzhen, 518026 (China); Short, Michael P., E-mail: hereiam@mit.edu [Department of Nuclear Science and Engineering, Massachusetts Institute of Technology (MIT), Cambridge, MA, 02139 (United States)

    2016-12-15

    Accelerator-driven systems (ADS) are a promising approach for nuclear waste disposal. Nevertheless, the principal candidate materials proposed for ADS construction, such as the ferritic/martensitic steel, T91, and austenitic stainless steels, 316L and 15-15Ti, are not fully compatible with the liquid lead-bismuth eutectic (LBE) coolant. Under some operating conditions, liquid metal embrittlement (LME) or liquid metal corrosion (LMC) may occur in these steels when exposed to LBE. These environmentally-induced material degradation effects pose a threat to ADS reactor safety, as failure of the materials could initiate a severe accident, in which fission products are released into the coolant. Meanwhile, parallel efforts to develop accident-tolerant fuels (ATF) in light water reactors (LWRs) could provide both general materials design philosophies and specific material solutions to the ADS program. In this paper, the potential contributions of the ATF materials development program to the ADS materials qualification program are evaluated and discussed in terms of service conditions and materials performance requirements. Several specific areas where coordinated development may benefit both programs, including composite materials and selected coatings, are discussed. - Highlights: • ATF materials developed for LWRs could be candidate materials for the LBE-cooled ADS program. • Similar material design and protection philosophies are utilized in both programs. • Unique challenges of LBE-cooled ADS systems could possibly be addressed by LWR ATF materials. • More coordinated testing should be performed between the ATF and ADS programs.

  17. Fe-Cr-V ternary alloy-based ferritic steels for high- and low-temperature applications

    International Nuclear Information System (INIS)

    Rieth, M.; Materna-Morris, E.; Dudarev, S.L.; Boutard, J.-L.; Keppler, H.; Mayor, J.

    2009-01-01

    The phase stability of alloys and steels developed for application in nuclear fission and fusion technology is one of the decisive factors determining the potential range of operating temperatures and radiation conditions that the core elements of a power plant can tolerate. In the case of ferritic and ferritic-martensitic steels, the choice of the chemical composition is dictated by the phase diagram for binary FeCr alloys where in the 0-9% range of Cr composition the alloy remains in the solid solution phase at and below the room temperature. For Cr concentrations exceeding 9% the steels operating at relatively low temperatures are therefore expected to exhibit the formation of α' Cr-rich precipitates. These precipitates form obstacles for the propagation of dislocations, impeding plastic deformation and embrittling the material. This sets the low temperature limit for the use of of high (14% to 20%) Cr steels, which for the 20% Cr steels is at approximately 600 deg. C. On the other hand, steels containing 12% or less Cr cannot be used at temperatures exceeding ∼600 deg. C due to the occurrence of the α-γ transition (912 deg. C in pure iron and 830 deg. C in 7% Cr alloy), which weakens the steel in the high temperature limit. In this study, we investigate the physical properties of a concentrated ternary alloy system that attracted relatively little attention so far. The phase diagram of ternary Fe-Cr-V alloy shows no phase boundaries within a certain broad range of Cr and V concentrations. This makes the alloy sufficiently resistant to corrosion and suggests that steels and dispersion strengthened materials based on this alloy composition may have better strength and stability at high temperatures. Experimental heats were produced on a laboratory scale by arc melting the material components to pellets, then by melting the pellets in an induction furnace and casting the melt into copper moulds. The compositions in weight percent (iron base) are 10Cr5V, 10Cr

  18. Corrosion of stainless steels in lead-bismuth eutectic up to 600 °C

    Science.gov (United States)

    Soler, L.; Martín, F. J.; Hernández, F.; Gómez-Briceño, D.

    2004-11-01

    An experimental program has been carried out to understand the differences in the corrosion behaviour between different stainless steels: the austenitic steels 304L and 316L, the martensitic steels F82Hmod, T91 and EM10, and the low alloy steel P22. The influence of oxygen level in Pb-Bi, temperature and exposure time is studied. At 600 °C, the martensitic steels and the P22 steel exhibit thick oxide scales that grow with time, following a linear law for the wet environment and a parabolic law for the dry one. The austenitic stainless steels show a better corrosion behaviour, especially AISI 304L. Under reducing conditions, the steels exhibit dissolution, more severe for the austenitic stainless steels. At 450 °C, all the materials show an acceptable behaviour provided a sufficient oxygen level in the Pb-Bi. At reducing conditions, the martensitic steels and the P22 steel have a good corrosion resistance, while the austenitic steels exhibit already dissolution at the longer exposures.

  19. Corrosion of stainless steels in lead-bismuth eutectic up to 600 deg. C

    International Nuclear Information System (INIS)

    Soler, L.; Martin, F.J.; Hernandez, F.; Gomez-Briceno, D.

    2004-01-01

    An experimental program has been carried out to understand the differences in the corrosion behaviour between different stainless steels: the austenitic steels 304L and 316L, the martensitic steels F82Hmod, T91 and EM10, and the low alloy steel P22. The influence of oxygen level in Pb-Bi, temperature and exposure time is studied. At 600 deg. C, the martensitic steels and the P22 steel exhibit thick oxide scales that grow with time, following a linear law for the wet environment and a parabolic law for the dry one. The austenitic stainless steels show a better corrosion behaviour, especially AISI 304L. Under reducing conditions, the steels exhibit dissolution, more severe for the austenitic stainless steels. At 450 deg. C, all the materials show an acceptable behaviour provided a sufficient oxygen level in the Pb-Bi. At reducing conditions, the martensitic steels and the P22 steel have a good corrosion resistance, while the austenitic steels exhibit already dissolution at the longer exposures

  20. Aircraft Steels

    Science.gov (United States)

    2009-02-19

    component usage. PH 13-8Mo is a precipitation-hardenable martensitic stainless steel combining excellent corrosion resistance with strength. Custom 465 is...a martensitic , age-hardenable stainless steel capable of about 1,724 MPa (250 ksi) UTS when peak-aged (H900 condition). Especially, this steel can...NOTES 14. ABSTRACT Five high strength steels (4340, 300M, AerMet 100, Ferrium S53, and Hy-Tuf) and four stainless steels (High Nitrogen, 13

  1. Microstructural development in multi-pass TIG welded F82H steels under dual-ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ogiwara, H.; Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Kishimoto, H.; Tomohiro, M.; Kohyama, A. [Kyoto Univ., Institute of Advanced Energy (Japan)

    2007-07-01

    Full text of publication follows: Reduced-activation ferritic/martensitic (RAFM) steels are the first candidate materials for blanket structural component. On fabricating blanket components, various joining technologies will be required, and it could be anticipated that some parts of the weldments will suffer from high-dose 14 MeV neutron irradiation. For example, the current Japanese water cooled test blanket module design has the weldment of the first wall and side wall at 25 mm from the surface. This will not be the case for TIER-TBM, but could be the critical issue if the same design is applied for DEMO blanket system. In this study, the effects of displacement damage and helium production on mechanical properties and microstructures of the multi-pass tungsten inert gas (TIG) welding were investigated. The welded joint used in this study is a multi-pass TIG weldment on 15-mm thick plates of F82H-IEA. The post weld heat treatment was carried out at 720 deg. C for 1 h. To obtain systematic and accurate information of the microstructural response under fusion environment, dual-ion irradiation was performed. 15 mm x 20 mm pieces which cover the whole TIG weldment were irradiated at 470 deg. C up to 20 dpa using 6.4 MeV Fe{sup +3} and/or energy-degraded 1.0 MeV He{sup +}. The damage rate was 1.0 x 10{sup -3} dpa/s, and the helium injection rate was 15 x 10{sup -3} appm He/s. After the irradiations, thin film samples were made from various regions of the weldment by focused ion beam (FIB) processor, and transmission electron microscopy (TEM) observation and hardness tests were performed. Microstructure and Vickers hardness profiles across base metal, heat affected zone (HAZ) and fusion zone (FZ) were examined before irradiation experiments. The hardness measurements revealed that the maximum hardness was observed at the last pass region of FZ, and the softest region was in the middle of the FZ and HAZ region near the transformation line. In the microstructure study

  2. Compatibility tests of steels in flowing liquid lead-bismuth

    International Nuclear Information System (INIS)

    Barbier, F.; Benamati, G.; Fazio, C.; Rusanov, A.

    2001-01-01

    The behaviour of steels exposed to flowing Pb-55Bi was evaluated. The materials tested are the two austenitic steels AISI 316L and 1.4970, and the six martensitic steels Optifer IVc, T91, Batman 27, Batman 28, EP823 and EM10 which were exposed to flowing Pb-55Bi for 1000, 2000 and 3000 h and at two temperatures (573 and 743 K). The corrosion tests were conducted in the non-isothermal loop of IPPE-Obninsk under a controlled oxygen level (10 -6 wt%). The compatibility study showed that at a lower temperature, a very thin oxide layer (<1 μm) was formed on the steels. At higher temperature, austenitic steels also exhibited a thin oxide layer sufficient to prevent their dissolution in the melt. A thicker oxide, which grew according to a parabolic law, was observed on the surface of the martensitic steels. The oxidation resistance behaviour of the martensitic steels was correlated with their alloying elements

  3. Steel making

    CERN Document Server

    Chakrabarti, A K

    2014-01-01

    "Steel Making" is designed to give students a strong grounding in the theory and state-of-the-art practice of production of steels. This book is primarily focused to meet the needs of undergraduate metallurgical students and candidates for associate membership examinations of professional bodies (AMIIM, AMIE). Besides, for all engineering professionals working in steel plants who need to understand the basic principles of steel making, the text provides a sound introduction to the subject.Beginning with a brief introduction to the historical perspective and current status of steel making together with the reasons for obsolescence of Bessemer converter and open hearth processes, the book moves on to: elaborate the physiochemical principles involved in steel making; explain the operational principles and practices of the modern processes of primary steel making (LD converter, Q-BOP process, and electric furnace process); provide a summary of the developments in secondary refining of steels; discuss principles a...

  4. Fatigue Properties of Aged Mod. 9Cr-1Mo

    International Nuclear Information System (INIS)

    Kim, Dae Whan; Kim, Sung Ho; Lee, Chan Bock

    2007-01-01

    Ferritic/Martensitic steel has a good mechanical properties and a lower thermal expansion coefficient than austenitic stainless steel. Mechanical property of Mod. 9Cr-1Mo steel is less than austenitic stainless steel at high temperature. High temperature mechanical properties are affected by precipitation for Mod. 9Cr-1Mo. FMS steel is used for long time at high temperature and the effect of aging on mechanical properties is very important. In this study, low cycle fatigue properties with aging were investigated

  5. Residual Stress Development in Explosive-Bonded Bi-Metal Composite Materials

    Science.gov (United States)

    2014-03-01

    solidification cracking in steels and stainless steels . He has also undertaken extensive work on improving the weld zone toughness of high strength steels ...information on microstructural characterisation at the interface of ferritic- martensitic and austenitic steels produced using the EW process, the...957. [3] I. Tatsukawa, I. Oda, ‘Residual Stress Measurements on Explosive Clad Stainless Steel ’, Trans. Japan Welding Soc., 2(2), 1971, p26-34

  6. Optimization of the Fabrication Route of Ferritic/Martensitic ODS Cladding Tubes: Metallurgical Approach and Pilgering Numerical Modeling

    International Nuclear Information System (INIS)

    Logé, R.E.; Vanegas-Marques, E.; Mocellin, K.; Toualbi, L.; Carlan, Y. de

    2013-01-01

    Conclusions: • Fabrication route of 9Cr-ODS (martensitic) alloys is well controlled. • Fabrication route of 14Cr-ODS (ferritic) should be further optimized. • The choice between a ferritic or a martensitic grade is not already done, it will depend also on the behaviour under irradiation, the corrosion resistance … • Part of the optimization can rely on numerical simulation of pilgering: • The constitutive behaviour is an essential ingredient for process optimization: appropriate cyclic laws must be used. • The numerical analysis can look at cracking risks, final yield stress, and even residual stress state or surface roughness. • HPTR laboratory approaches can be translated to the (industrial) VMR process provided some additional adjustments in the numerical code

  7. Application of aluminum diffusion coatings to mitigate the KCl-induced high-temperature corrosion

    DEFF Research Database (Denmark)

    Kiamehr, Saeed; Lomholt, T. N.; Dahl, Kristian Vinter

    2017-01-01

    Pack cementation was used to produce Fe1−xAl and Fe2Al5 diffusion coatings on ferritic-martensitic steel P91 and a Ni2Al3 diffusion coating on pure nickel. The performance of diffusion coatings against high-temperature corrosion induced by potassium chloride (KCl) was evaluated by exposing...

  8. Effect of alloy content on microstructure and microchemistry of ...

    Indian Academy of Sciences (India)

    This paper presents the results of an experimental study on the microstructural evolution in 9Cr reduced activation ferritic/martensitic steels during short term thermal exposures. Since the microstructure is strongly influenced by the alloying additions, mainly W, Ta and C contents, the effect of varying W and Ta contents on ...

  9. Influencia de la temperatura de austenización y tiempo de permanencia sobre el tamaño de grano en aceros ferrítico-martensíticos del tipo 9Cr1MoVNb utilizados en calderas supercríticas

    Directory of Open Access Journals (Sweden)

    Gutiérrez-Urrutia, L.

    2001-04-01

    Full Text Available The aim of the present work is to determine the influence of austenitizing temperature holding time and heating velocity on grain size of ferritic-martensitic type 9CrlMoVNb steels developed for Oak Ridge National Laboratories & Combustion Engineering, T91/P91[1] Steels (USA and X10CrMoVNb 9.1 (Europe. The potential benefits of this material, in term of high resistance, good ductily and oxidation resistance, weldability and in particular good high temperature creep strength, are now widely acknowledged, particulary by supercritical boilers (P = 300 bar, T = 550±50 °C. The studied steels have been produced in, high frecuency induction vacum furnaces and hot-rolled.

    Se estudia la influencia de la temperatura de austenización, tiempo de permanencia a esta temperatura y velocidad de calentamiento sobre el tamaño de grano de varios aceros ferrítico-martensíticos de última generación del tipo 9CrlMoVNb, desarrollados por Oak Ridge National Laboratories (ORNL conjuntamente con Combustion Enginering en USA, conocidos por ASME/ASTM por las designaciones T91/P91[1] y en Europa como X10CrMoVNb9.1. Son aceros ductiles y tenaces que presentan muy buenas propiedades de resistencia a fluencia, soldabilidad y conductividad térmica. Son cada vez más utilizados como tubos de alta temperatura (550±50 °C y presión (300 bar en calderas supercríticas para centrales térmicas avanzadas. Los aceros estudiados se han fabricado en hornos de inducción de alta frecuencia al vacío y laminados en caliente.

  10. Mechanical dispersion of Y2O3 nanoparticles in steel EUROFER 97: process and optimisation

    International Nuclear Information System (INIS)

    Castro, V. de; Leguey, T.; Monge, M.A.; Munoz, A.; Pareja, R.; Amador, D.R.; Torralba, J.M.; Victoria, M.

    2003-01-01

    The procedures followed to produce Y 2 O 3 -dispersed EUROFER 97 powder ready to be compacted and hot isostatic pressing processed are reported. An attrition mill has been used under controlled conditions. The compositional and microstructural characterization of the Y 2 O 3 /EUROFER powder along the different steps of the milling process has allowed optimising the processing conditions to obtain a nanosized Y 2 O 3 dispersion. TEM observations performed on Y 2 O 3 /EUROFER powder milled under these specific conditions reveal the presence of monoclinic Y 2 O 3 dispersoids having sizes around 10 nm in the ferrite/martensite matrix

  11. The Cold Gas-Dynamic Spray and Characterization of Microcrystalline and Nanocrystalline Copper Alloys

    Science.gov (United States)

    2012-12-01

    set—the vial itself is constructed of a polycarbonate shell with stainless steel end plugs with a stainless steel impactor contained within the vial...Alloying of Refractory Metals in Austenitic and Ferritic/ Martensitic Steels ,” M.S. Thesis, MAE Dept., Naval Postgraduate School, Monterey, CA, 2012...Process. From [2]. .....................................................5  Figure 3.  Individual Particle after Impact with a Steel Substrate. From [15

  12. High temperature oxidation in boiler environment of chromized steel

    Science.gov (United States)

    Alia, F. F.; Kurniawan, T.; Asmara, Y. P.; Ani, M. H. B.; Nandiyanto, A. B. D.

    2017-10-01

    The demand for increasing efficiency has led to the development and construction of higher operating temperature power plant. This condition may lead to more severe thickness losses in boiler tubes due to excessive corrosion process. Hence, the research to improve the corrosion resistance of the current operated material is needed so that it can be applied for higher temperature application. In this research, the effect of chromizing process on the oxidation behaviour of T91 steel was investigated under steam condition. In order to deposit chromium, mixture of chromium (Cr) powder as master alloy, halide salt (NH4Cl) powder as activator and alumina (Al2O3) powder as inert filler were inserted into alumina retort together with the steel sample and heated inside furnace at 1050°C for ten hours under argon gas environment. Furthermore, for the oxidation process, steels were exposed at 700°C at different oxidation time (6h-24h) under steam condition. From FESEM/EDX analysis, it was found that oxidation rate of pack cemented steel was lower than the un-packed steel. These results show that Cr from chromizing process was able to become reservoir for the formation of Cr2O3 in high temperature steam oxidation, and its existence can be used for a longer oxidation time.

  13. Positron annihilation studies on structural materials for nuclear reactors

    International Nuclear Information System (INIS)

    Rajaraman, R.; Amarendra, G.; Sundar, C.S.

    2012-01-01

    Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

  14. Quantification of the degradation of steels exposed to liquid lead-bismuth eutectic

    International Nuclear Information System (INIS)

    Schroer, C.; Voss, Z.; Novotny, J.; Konys, J.

    2006-05-01

    Metallographic and gravimetric methods of measuring the degradation of steels are introduced and compared, with emphasis on the quantification of oxidation in molten lead-bismuth eutectic (LBE). In future applications of LBE or other molten lead alloys, additions of oxygen should prevent the dissolution of steel constituents in the liquid heavy metal. Therefore, also the amount of steel constituents transferred between the steel (including the oxide scale formed on the surface) and the LBE has to be assessed, in order to evaluate the efficiency of oxygen additions with respect to preventing dissolution of the steel. For testing the methods of quantification, specimens of martensitic steel T91 were exposed for 1500 h to stagnant, oxygen-saturated LBE at 550 C, whereby, applying both metallographic and gravimetric measurements, the recession of the cross-section of sound material deviated by ± 3 μm for a mean value of 11 μm. Although the transfer of steel constituents between the solid phases and the LBE is negligible under the considered exposure conditions, the investigation shows that a gravimetric analysis is most promising for quantifying such a mass transfer. For laboratory experiments on the behaviour of steels in oxygen-containing LBE, it is suggested to make provisions for both metallographic and gravimetric measurements, since both types of methods have specific benefits in the characterisation of the oxidation process. (Orig.)

  15. Tool steels

    DEFF Research Database (Denmark)

    Højerslev, C.

    2001-01-01

    On designing a tool steel, its composition and heat treatment parameters are chosen to provide a hardened and tempered martensitic matrix in which carbides are evenly distributed. In this condition the matrix has an optimum combination of hardness andtoughness, the primary carbides provide...... resistance against abrasive wear and secondary carbides (if any) increase the resistance against plastic deformation. Tool steels are alloyed with carbide forming elements (Typically: vanadium, tungsten, molybdenumand chromium) furthermore some steel types contains cobalt. Addition of alloying elements...... serves primarily two purpose (i) to improve the hardenabillity and (ii) to provide harder and thermally more stable carbides than cementite. Assuming proper heattreatment, the properties of a tool steel depends on the which alloying elements are added and their respective concentrations....

  16. Fireside corrosion and steamside oxidation of 9-12% Cr martensitic steels exposed for long term testing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Jensen, S. A.; Rasmussen, F.

    2009-01-01

    MoV121 and HCM12 for the 12% Cr steels. The test tubes were welded in as part of the existing final superheaters in actual plants and exposure has been conducted over a ten year period (1994-2005). Compared to the older steel types, T92 and HCM12 utilise tungsten to improve their creep strength. From......To obtain long term corrosion and steam oxidation data for the 9-12%Cr ferritic steels, test tube sections have been exposed in Amager 3 and Avedore 1 coal fired power plants in Denmark (formerly run by ENERGI E2). Thus direct comparisons can be made for T91 and T92 for the 9% Cr steels and X20Cr...... Avedore I testing, T91 and T92 can be compared for exposure times up to similar to 48 000 h exposure. From Amager 3 testing, X20, HCM12 and T92 were tested; T92 has been exposed for up to 31 000 h and X20 and HCM12 have had 84 500 h exposure. Tube sections were removed for various exposure durations...

  17. Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

    International Nuclear Information System (INIS)

    Byun, T.S.; Farrell, K.; Lee, E.H.; Mansur, L.K.; Maloy, S.A.; James, M.R.; Johnson, W.R.

    2002-01-01

    This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 deg. C. Tensile testing was performed at room temperature (20 deg. C) and 164 deg. C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 deg. C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability

  18. FY17 Status Report on the Micromechanical Finite Element Modeling of Creep Fracture of Grade 91 Steel

    Energy Technology Data Exchange (ETDEWEB)

    Messner, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Truster, T. J. [Univ. of Tennessee, Knoxville, TN (United States); Cochran, K. B. [DR& C Inc.; Parks, D. M. [DR& C Inc.; Sham, T. -L. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    Advanced reactors designed to operate at higher temperatures than current light water reactors require structural materials with high creep strength and creep-fatigue resistance to ac