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Sample records for ferritic steel irradiated

  1. Irradiation embrittlement of ferritic stainless steels

    International Nuclear Information System (INIS)

    Suganuma, K.; Kayano, H.

    1984-01-01

    The characteristics of the irradiation embrittlement of some ferritic stainless steels were examined by tensile tests. Steels selected in this investigation were classified into three groups: chi phase, precipitation hardened Fe-13Cr steels; tempered martensitic Fe-12Cr steels; and low alloy steels. The latter steels were chosen in order to compare the irradiation embrittlement characteristics with those of stainless steels. The stainless steels were superior to the low alloy steels with regard to the irradiation embrittlement (the changes in both ductile-brittle transition temperature (DBTT) and unstable plastic flow transition temperature (UPFTT)), irrespective of whether these stainless steels had chi phase precipitated structures or tempered martensitic structures. The suppression of the DBTT increase owing to irradiation results from low yield stress increase Δσsub(y) and high |[dσsub(y)(u)/dT]|, where u denotes unirradiated, in the stainless steels. The suppression of the UPFTT results from the high work hardening rate or the high work exponent and the low Lueders strain in the stainless steels. These characteristics of irradiation embrittlement in the ferritic stainless steels are thought to be caused by the defect structure, which is modified by Cr atoms. (author)

  2. Mechanical behaviour of ferritic martensitic steels irradiated in Phenix. Introduction at the Icone irradiation

    International Nuclear Information System (INIS)

    Seran, J.L.

    1988-01-01

    Ferritic-martensitic steels are actually possible candidates for material of fast neutron reactors hexagonal tubes. These steels possess a swelling and a creep resistance better than of classic 316 austenitic steels and present out of irradition, mechanical characteristics suitable for the proposed application and good manufacturing properties and sodium compatibility. In ferritic steels irradiation effects came forward at low temperature that for austenitic steels. In the precedent seminary we have shown that the maximum of swelling was unknown and takes probably place at a temperature below 400 0 C. The same question sets up concerning the localization of temperature range in which the mechanical characteristics of ferritic steels are affected by irradiation. In this communication, we give the first results permitting to compare the mechanical properties in traction and in resilience observed after a 50 atom displacement irradiation on a F17 ferritic steel, a EM12 ferritic-martensitic steel and a EM 10 martensitic steel [fr

  3. 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

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1997-06-01

    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.

  6. Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

    Science.gov (United States)

    Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

    2017-09-01

    Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-11-01

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

  8. Phase stability of oxide dispersion-strengthened ferritic steels in neutron irradiation

    International Nuclear Information System (INIS)

    Yamashita, S.; Oka, K.; Ohnuki, S.; Akasaka, N.; Ukai, S.

    2002-01-01

    Oxide dispersion-strengthened ferritic steels were irradiated by neutrons up to 21 dpa and studied by microstructural observation and microchemical analysis. The original high dislocation density did not change after neutron irradiation, indicating that the dispersed oxide particles have high stability under neutron irradiation. However, there is potential for recoil resolution of the oxide particles due to ballistic ejection at high dose. From the microchemical analysis, it was implied that some of the complex oxides have a double-layer structure, such that TiO 2 occupied the core region and Y 2 O 3 the outer layer. Such a structure may be more stable than the simple mono-oxides. Under high-temperature irradiation, Laves phase was the predominant precipitate occurring at grain boundaries α phase and χ phase were not observed in this study

  9. 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

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

    International Nuclear Information System (INIS)

    Zhang, Z.W.; Liu, C.T.; Wang, X-.L.; Miller, M.K.; Ma, D.; Chen, G.; Williams, J.R.; Chin, B.A.

    2012-01-01

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

  11. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. Microstructural evolution of P92 ferritic/martensitic steel under Ar+ ion irradiation at elevated temperature

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  14. Hardness distribution and effect of irradiation in FSW-ODS ferritic steels

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Kasada, Ryuta; Kimura, Akihiko; Nagasaka, Takuya; Sokolov, M.A.; Yamamoto, T.

    2014-01-01

    Oxide dispersion strengthened ferritic steels (ODS-FS) have been considered as one of the most promising structural materials for advanced nuclear systems such as fusion reactors and next generation fission reactors, because of its excellent elevated temperature strength, corrosion and radiation resistance. Especially, irradiation resistance is a critical issue for the high performance of ODS-FS. In this study, effects of the irradiation on hardness properties of friction stri processed (FSP) ODS-FS were investigated. FSP technique was employed on ODS-FS. A plate specimen was cut out from the cross section and irradiated to 1.2 dpa at 573K in the High Flux Isotope Reactor (HFIR). To investigate the effect of neutron irradiation on processed area, the hardness distributions were evaluated on the cross section. Hardness of FSP ODS-FS was various with each microstructure after irradiation to 1.2 dpa at 573K. The increase of Vickers hardness was significant in the stirred zone and heat affected zone. Base material exhibited the lowest hardening about 38HV. Since nano-oxide particles in stirred zone showed identical mean diameter and number density, it is considered that hardening differences between stirred zone and base material is due to differences in initial dislocation density. (author)

  15. 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

  16. 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.

  17. 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)

  18. 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

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

    International Nuclear Information System (INIS)

    Pareige, P.

    1996-04-01

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

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

    Science.gov (United States)

    Topbasi, Cem

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

  1. 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

  2. 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.

  3. Radiation response of ODS ferritic steels with different oxide particles under ion-irradiation at 550 °C

    Science.gov (United States)

    Song, Peng; Morrall, Daniel; Zhang, Zhexian; Yabuuchi, Kiyohiro; Kimura, Akihiko

    2018-04-01

    In order to investigate the effects of oxide particles on radiation response such as hardness change and microstructural evolution, three types of oxide dispersion strengthened (ODS) ferritic steels (named Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS), mostly strengthened by Y-Ti-O, Y-Al-O and Y-Zr-O dispersoids, respectively, were simultaneously irradiated with iron and helium ions at 550 °C up to a damage of 30 dpa and a corresponding helium (He) concentration of ∼3500 appm to a depth of 1000-1300 nm. A single iron ion beam irradiation was also performed for reference. Transmission electron microscopy revealed that after the dual ion irradiation helium bubbles of 2.8, 6.6 and 4.5 nm in mean diameter with the corresponding number densities of 1.1 × 1023, 2.7 × 1022 and 3.6 × 1022 m-3 were observed in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS, respectively, while no such bubbles were observed after single ion irradiation. About 80% of intragranular He bubbles were adjacent to oxide particles in the ODS ferritic steels. Although the high number density He bubbles were observed in the ODS steels, the void swelling in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS was still small and estimated to be 0.13%, 0.53% and 0.20%, respectively. The excellent swelling resistance is dominantly attributed to the high sink strength of oxide particles that depends on the morphology of particle dispersion rather than the crystal structure of the particles. In contrast, no dislocation loops were produced in any of the irradiated steels. Nanoindentation measurements showed that no irradiation hardening but softening was found in the ODS ferritic steels, which was probably due to irradiation induced dislocation recovery. The helium bubbles in high number density never contributed to the irradiation hardening of the ODS steels at these irradiation conditions.

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

    Science.gov (United States)

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

    1998-10-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  6. Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

    International Nuclear Information System (INIS)

    James, L.A.

    1977-01-01

    The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  8. 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.

  9. Articles comprising ferritic stainless steels

    Science.gov (United States)

    Rakowski, James M.

    2016-06-28

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

  10. Cluster dynamics modeling of Mn-Ni-Si precipitates in ferritic-martensitic steel under irradiation

    Science.gov (United States)

    Ke, Jia-Hong; Ke, Huibin; Odette, G. Robert; Morgan, Dane

    2018-01-01

    Mn-Ni-Si precipitates (MNSPs) are known to be responsible for irradiation-induced hardening and embrittlement in structural alloys used in nuclear reactors. Studies have shown that precipitation of the MNSPs in 9-Cr ferritic-martensitic (F-M) alloys, such as T91, is strongly associated with heterogeneous nucleation on dislocations, coupled with radiation-induced solute segregation to these sinks. Therefore it is important to develop advanced predictive models for Mn-Ni-Si precipitation in F-M alloys under irradiation based on an understanding of the underlying mechanisms. Here we use a cluster dynamics model, which includes multiple effects of dislocations, to study the evolution of MNSPs in a commercial F-M alloy T91. The model predictions are calibrated by data from proton irradiation experiments at 400 °C. Radiation induced solute segregation at dislocations is evaluated by a continuum model that is integrated into the cluster dynamics simulations, including the effects of dislocations as heterogeneous nucleation sites. The result shows that MNSPs in T91 are primarily irradiation-induced and, in particular, both heterogeneous nucleation and radiation-induced segregation at dislocations are necessary to rationalize the experimental observations.

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

    NARCIS (Netherlands)

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

    2003-01-01

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

  12. multi-scale modeling of helium transport and fate in irradiated nano-structured ferritic alloys and tempered martensitic steels

    International Nuclear Information System (INIS)

    Yamamoto, T.; Odette, G.; Hribernik, M.; Kurtz, R.J.; Wirth, B.

    2007-01-01

    Full text of publication follows: We describe the development and application of a multi-scale model of the transport and fate of He in irradiated nano-structured ferritic alloys (NFAs) and tempered martensitic steels (TMS) that are candidates for use in fusion first-wall and blanket structural applications. We focus on NFAs that have remarkable creep strength provided by a high density of Y-Ti-O solute clusters and oxides nano-features (NF) that not only impede dislocation motion, but can also provide fine scale helium bubble nucleation sites and vacancy-interstitial recombination centers. Key characteristics of NFAs are 1) a high density (∼10 24 m -3 ) of small (∼2-4 nm diameter) NF, 2) fine to ultra-fine crystallite grain sizes and 3) high dislocation densities. The size and number density of these features can be modified by appropriate thermo-mechanical treatments. We employ molecular dynamics (MD) simulations to assess the binding and migration energies of He and defects with each other and at various trapping sites such as coherent precipitate interfaces, dislocation jogs and representative grain boundaries. Kinetic Lattice Monte Carlo (KLMC) simulations are used to determine migration mechanisms and diffusion coefficients of substitutional and interstitial helium. KLMC is also used to model helium and vacancy clustering on precipitate interfaces, on dislocation lines and in grain boundaries. The effects of radiation induced vacancies and self-interstitial atoms are modeled in detail, including their aggregation in loops and cavities. The MD and KLMC simulations provide critical information for rate theory and cluster dynamics models that follow point defect and helium transport and partitioning to, and recycling between, matrix cavities, precipitates, dislocations and grain boundaries and the precipitation of helium bubbles and possible conversion to growing voids. The effects of irradiation variables like the irradiation temperature (300-800 deg. C

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Aubert, M.; Mathieu, B.; Petrequin, P.

    1983-06-01

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 530 0 C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  18. Characterization and assessment of ferritic/martensitic steels

    International Nuclear Information System (INIS)

    Ehrlich, K.; Harries, D.R.; Moeslang, A.

    1997-02-01

    Ferritic/martensitic steels are candidate structural materials for a DEMO fusion reactor and are investigated intensively within the frame of the European Long Term Fusion Technology Programme. This report summarizes general features of ferritic/martensitic steels and gives a broad overview on the available data of the 9-12% CrMoVNb steels MANET I and II. The data include informations on the physical metallurgy, the transformation and hardening/tempering behaviour as well as results on tensile, creep, creep-rupture, isothermal and thermal fatigue, charpy impact and fracture toughness properties. Other topics are corrosion tests of helium or high pressure water coolants, compatibility with breeding and neutron multiplier materials, advanced welding techniques, and a short review on fabrication and technology of these steels. In the relevant temperature region from ambient temperatures to 450 C a widespread field of results on pre-, postirradiation and in-situ mechanical properties is available up to a few dpa and up to 500 appm helium. Special emphasis has been put on the recent world-wide optimization of these steels. New 7-10% CrWVTa steels have been developed with significantly improved impact and fracture toughness properties. Initial results from unirradiated and neutron irradiated charpy specimens from various heats are favourable and showed a general improvement of the fracture toughness properties. These ferritic/martensitic steels also satisfy the criteria of reduced long-term activation. The potential for fusion applications is discussed together with some guidelines for required R and D. (orig.)

  19. 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.

  20. Plasma spot welding of ferritic stainless steels

    International Nuclear Information System (INIS)

    Lesnjak, A.; Tusek, J.

    2002-01-01

    Plasma spot wedding of ferritic stainless steels studied. The study was focused on welding parameters, plasma and shieldings and the optimum welding equipment. Plasma-spot welded overlap joints on a 0.8 mm thick ferritic stainless steel sheet were subjected to a visual examination and mechanical testing in terms of tension-shear strength. Several macro specimens were prepared Plasma spot welding is suitable to use the same gas as shielding gas and as plasma gas , i. e. a 98% Ar/2% H 2 gas mixture. Tension-shear strength of plasma-spot welded joint was compared to that of resistance sport welded joints. It was found that the resistance welded joints withstand a somewhat stronger load than the plasma welded joints due to a large weld sport diameter of the former. Strength of both types of welded joints is approximately the same. (Author) 32 refs

  1. Ferritic stainless steels: corrosion resistance + economy

    International Nuclear Information System (INIS)

    Remus, A.L.

    1976-01-01

    Ferritic stainless steels provide corrosion resistance at lower cost. They include Type 409, Type 439, 18SR, 20-Mo (1.6 Mo), 18-2 (2 Mo), 26-1S, E-Brite 26-1, 29 Cr-4 Mo, and 29 Cr-4 Mo-2 Ni. Their corrosion and mechanical properties are examined. Resistance to stress-corrosion cracking is an advantage compared to austenitic types

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

    International Nuclear Information System (INIS)

    Dai, Y.

    1996-01-01

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

  3. 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.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  5. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

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

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

    International Nuclear Information System (INIS)

    Quick, N.R.

    1976-01-01

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

  8. Plasma spot welding of ferritic stainless steels

    Directory of Open Access Journals (Sweden)

    Lešnjak, A.

    2002-06-01

    Full Text Available Plasma spot welding of ferritic stainless steels is studied. The study was focused on welding parameters, plasma and shielding gases and the optimum welding equipment. Plasma-spot welded overlap joints on a 0.8 mm thick ferritic stainless steel sheet were subjected to a visual examination and mechanical testing in terms of tension-shear strength. Several macro specimens were prepared. Plasma spot welding is suitable to use the same gas as shielding gas and as plasma gas, i.e., a 98 % Ar/2 % H 2 gas mixture. Tension-shear strength of plasma-spot welded joints was compared to that of resistance-spot welded joints. It was found that the resistance welded joints withstand a somewhat stronger load than the plasma welded joints due to a larger weld spot diameter of the former. Strength of both types of welded joints is approximately the same.

    El artículo describe el proceso de soldeo de aceros inoxidables ferríticos por puntos con plasma. La investigación se centró en el establecimiento de los parámetros óptimos de la soldadura, la definición del gas de plasma y de protección más adecuado, así como del equipo óptimo para la realización de la soldadura. Las uniones de láminas de aceros inoxidables ferríticos de 0,8 mm de espesor, soldadas a solape por puntos con plasma, se inspeccionaron visualmente y se ensayaron mecánicamente mediante el ensayo de cizalladura por tracción. Se realizaron macro pulidos. Los resultados de la investigación demostraron que la solución más adecuada para el soldeo por puntos con plasma es elegir el mismo gas de plasma que de protección. Es decir, una mezcla de 98 % de argón y 2 % de hidrógeno. La resistencia a la cizalladura por tracción de las uniones soldadas por puntos con plasma fue comparada con la resistencia de las uniones soldadas por resistencia por puntos. Se llegó a la conclusión de que las uniones soldadas por resistencia soportan una carga algo mayor que la uniones

  9. 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.

  10. Small-angle neutron scattering investigation of the nanostructure of ferritic-martensitic 12%-chromium steels

    Science.gov (United States)

    Bogdanov, S. G.; Goshchitskii, B. N.; Parkhomenko, V. D.; Leontieva-Smirnova, M. V.; Chernov, V. M.

    2014-01-01

    The nanostructure (nanoparticle distribution) of ferritic-martensitic 12%-chromium steels EK-181 (Fe-12Cr-2W-V-Ta-B) and ChS-139 (Fe-12Cr-2W-V-Ta-B-Nb-Mo) subjected to different modes of mechanical and heat treatments and neutron irradiation has been investigated using small-angle neutron scattering. The samples have been studied in the initial state and after neutron irradiation (IVV-2M reactor) at a temperature of 80°C with fluences of 1018, 1019, and 5 × 1019 cm-2 ( E ≥ 0.1 MeV). The nanostructure of the steels is characterized by precipitations of nanoparticles with two characteristic sizes of 1.0-1.5 and 7-8 nm. The dependence of the nanostructure parameters on the composition of the steels and on the conditions of heat treatment and irradiation has been discussed.

  11. Residual stress studies of austenitic and ferritic steels

    International Nuclear Information System (INIS)

    Chrenko, R.M.

    1978-01-01

    Residual studies have been made on austenitic and ferritic steels of the types used as structural materials. The residual stress results presented here will include residual stress measurements in the heat-affected zone on butt welded Type 304 stainless steel pipes, and the stresses induced in Type 304 austenitic stainless steel and Type A508 ferritic steel by several surface preparations. Such surface preparation procedures as machining and grinding can induce large directionality effects in the residual stresses determined by X-ray techniques and some typical data will be presented. A brief description is given of the mobile X-ray residual stress apparatus used to obtain most of the data in these studies. (author)

  12. 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)

  13. Nitrogen alloying of the 12% Cr martensitic-ferritic steel

    Science.gov (United States)

    Kudryavtsev, A. S.; Artem'eva, D. A.; Mikhailov, M. S.

    2017-08-01

    The influence of the nitrogen content on the structure and mechanical properties of heat and corrosion resistant 12% Cr martensitic-ferritic steel developed at the Central Research Institute of Structural Materials Prometey has been studied. Steel containing 0.061 wt % nitrogen possesses a high level of mechanical properties. The decrease in the nitrogen content to 0.017 wt % leads to an increase of structurally free ferrite fraction in the steel, a decrease in the density of dislocations, a decrease of structural dispersity and the absence of finely dispersed precipitates of niobium and vanadium nitrides and carbides. As a result, there is a decrease in the strength properties, especially in the heat resistance.

  14. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

  15. Determination of delta ferrite volumetric fraction in austenitic stainless steel

    International Nuclear Information System (INIS)

    Almeida Macedo, W.A. de.

    1983-01-01

    Measurements of delta ferrite volumetric fraction in AISI 304 austenitic stainless steels were done by X-ray diffraction, quantitative metallography (point count) and by means of one specific commercial apparatus whose operational principle is magnetic-inductive: The Ferrite Content Meter 1053 / Institut Dr. Foerster. The results obtained were comparated with point count, the reference method. It was also investigated in these measurements the influence of the martensite induced by mechanical deformation. Determinations by X-ray diffraction, by the ratio between integrated intensities of the ferrite (211) and austenite (311) lines, are in excelent agreement with those taken by point count. One correction curve for the lectures of the commercial equipment in focus was obtained, for the range between zero and 20% of delta ferrite in 18/8 stainless steels. It is demonstrated that, depending on the employed measurement method and surface finishing of the material to be analysed, the presence of martensite produced by mechanical deformation of the austenitic matrix is one problem to be considered. (Author) [pt

  16. Determination of delta ferrite volumetric fraction in austenitic stainless steels

    International Nuclear Information System (INIS)

    Almeida Macedo, W.A. de.

    1983-01-01

    Measurements of delta ferrite volumetric fraction in AISI 304 austenitic stainless steels were done by X-ray difraction, quantitative metallography (point count) and by means of one specific commercial apparatus whose operational principle is magnetic-inductive: The Ferrite Content Meter 1053 / Institut Dr. Forster. The results obtained were comparated with point count, the reference method. It was also investigated in these measurements the influence of the martensite induced by mechanical deformation. Determinations by X-ray diffraction, by the ratio between integrated intensities of the ferrite (211) and austenite (311) lines, are in excelent agreement with those taken by point count. One correction curve for the lectures of the commercial equipment in focus was obtained, for the range between zero and 20% of delta ferrite in 18/8 stainless steels. It is demonstrated that, depending on the employed measurement method and surface finishing of the material to be analysed, the presence of martensite produced by mechanical deformation of the austenitic matrix is one problem to be considered. (Author) [pt

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

    Science.gov (United States)

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

    2017-02-01

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

  18. 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

  19. Stability under irradiation of a fine dispersion of oxides in a ferritic matrix

    International Nuclear Information System (INIS)

    Monnet, I.

    1999-01-01

    Oxide dispersion strengthened (ODS) ferritic-martensitic steels are being considered for high temperature, high fluence nuclear applications, like fuel pin cladding in Fast Breeder Reactors. ODS alloys offer improved out of pile strength characteristics at temperature above 550 deg.C and ferritic-martensitic matrix is highly swelling resistant. A clad in an ODS ferritic steel, call DY (Fe-13Cr-1,5Mo+TiO 2 +Y 2 O 3 ) has been irradiated in the experimental reactor Phenix. Under irradiation oxide dissolution occurs. Microstructural observations indicated that oxide evolution is correlated with the dose and consist in four phenomena: the interfaces of oxide particles with the matrix become irregular, the uniform distribution of the finest oxide ( 2 O 3 , Y 2 O 3 , MgO or MgAl 2 O 4 . These materials were irradiated with charged particles in order to gain a better understanding of the mechanisms of dissolution. Irradiation with 1 MeV Helium does not induce any modification, neither in the chemical modification of the particles nor in their spatial and size distribution. Since most of the energy of helium ions is lost by inelastic interaction, this result proves that this kind of interaction does not induce oxide dissolution. Irradiation with 1 MeV or 1.2 MeV electrons leads to a significant dissolution with a radius decrease proportional to the dose. These experiments prove that oxide dissolution can be induced by Frenkel pairs alone, provided that metallic atoms are displaced. The comparison between irradiation with ions (displacements cascades) and electrons (Frenkel pairs only) shows the importance of free point defects in the dissolution phenomena. For all the irradiations (ions or electrons) the spinel MgAl 2 O 4 seems more resistant than Y 2 O 3 to dissolution, and MgO and Al 2 O 3 are even less resistant. This is the order of stability under irradiation of bulk oxides. (author)

  20. Effects of irradiation on mechanical properties of HIP-bonded reduced-activation ferritic/martensitic steel F82H first wall

    International Nuclear Information System (INIS)

    Kazuyuki, Furuya; Eiichi, Wakai; Kenji, Miyamoto; Masato, Akiba; Masayoshi, Sugimoto

    2007-01-01

    HIP-bonded regions in the first wall of a fusion blanket are subjected to intense neutron irradiation. The purpose of this study is to investigate the influence of radiation damage on the tensile properties of the HIP-bonded regions. Tensile tests have been performed on specimens taken from a HIP-bonded mock-up structure, made to simulate the different fabrication processes. The neutron irradiation was carried out at about 423 K and 523 K to doses up to about 2 dpa. The tensile tests were performed at room temperature, irradiation temperatures and at 623 K. The main results are as follows: (1) Before irradiation, the tensile properties in the HIP-interface were equivalent to those of the matrix region. (2) Rupture did not occur at the HIP-interface of irradiated material. (3) The tensile properties in irradiated material were not notably affected due to manufacturing/fabricating histories. (4) Changes in properties produced by irradiation at 423 K show significant recovery for a test temperature of 673 K

  1. Mechanical alloying of lanthana-bearing nanostructured ferritic steels

    International Nuclear Information System (INIS)

    Pasebani, S.; Charit, I.; Wu, Y.Q.; Butt, D.P.; Cole, J.I.

    2013-01-01

    A novel nanostructured ferritic steel powder with the nominal composition Fe–14Cr–1Ti–0.3Mo–0.5La 2 O 3 (wt.%) was developed via high energy ball milling. La 2 O 3 was added to this alloy instead of the traditionally used Y 2 O 3 . The effects of varying the ball milling parameters, such as milling time, steel ball size and ball to powder ratio, on the mechanical properties and microstructural characteristics of the as-milled powder were investigated. Nanocrystallites of a body-centered cubic ferritic solid solution matrix with a mean size of approximately 20 nm were observed by transmission electron microscopy. Nanoscale characterization of the as-milled powder by local electrode atom probe tomography revealed the formation of Cr–Ti–La–O-enriched nanoclusters during mechanical alloying. The Cr:Ti:La:O ratio is considered “non-stoichiometric”. The average size (radius) of the nanoclusters was about 1 nm, with number density of 3.7 × 10 24 m −3 . The mechanism for formation of nanoclusters in the as-milled powder is discussed. La 2 O 3 appears to be a promising alternative rare earth oxide for future nanostructured ferritic steels

  2. Non-uniformity of hot plastic strain of stainless steels with austenitic-ferritic structure

    International Nuclear Information System (INIS)

    Laricheva, L.P.; Peretyat'ko, V.N.; Rostovtsev, A.N.; Levius, A.M.

    1987-01-01

    Non-uniformity of hot strain of stainless steels of various alloying was investigated. Steels with austenite and δ-ferrite structure of two classes were chosen for investigation: 08Kh18N10T steel of austenitic class and 08Kh21N5T steel of austenitic-ferritic class. Tests were conducted for samples subjected to preliminary thermal treatment: heating up to 1250 deg C, holding during 0.5 h, cooling in water. The heat treatment enabled to produce large grains of austenite and δ-ferrite (about 30 μm) in 08Kh21N5T steel, and sufficient amount of δ-ferrite (up to 50%) in 08Kh18N10T steel. It is shown that hot strain of austenitic-ferritic steels is non-uniform. δ-ferrite strain is more pronounced as compared to austenite. The ratio of mean δ-ferrite strain to the mean austenite strain grows with increase of the degree of general steel strain and temperature. The ratio of mean phase strains in 08Kh18N10T steel is higher as compared to 08Kh21N5T steel, general strain and temperature being equal. Temperature effect on the ratio of δ-ferrite and austenite strains is more pronounced for 08Kh18N10T steel. It is explaind by the value of ratios of phase strain resistance and temperature effect on them

  3. SELECTIVE SEPARATION OF URANIUM FROM FERRITIC STAINLESS STEELS

    Science.gov (United States)

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

    1963-05-14

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

  4. Thin slab processing of acicular ferrite steels with high toughness

    Energy Technology Data Exchange (ETDEWEB)

    Reip, Carl-Peter; Hennig, Wolfgang; Hagmann, Rolf [SMS Demag Aktiengesellschaft, Duesseldorf (Germany); Sabrudin, Bin Mohamad Suren; Susanta, Ghosh; Weng Lan Lee [Megasteel Sdn Bhd, Banting (Malaysia)

    2005-07-01

    Near-net-shape casting processes today represent an important option in steelmaking. High productivity and low production cost as well as the variety of steel grades that can be produced plus an excellent product quality are key factors for the acceptance of such processes in markets all over the world. Today's research focuses on the production of pipe steel with special requirements in terms of toughness at low temperatures. The subject article describes the production of hot strip made from acicular ferritic / bainitic steel grades using the CSP thin-slab technology. In addition, the resulting strength and toughness levels as a function of the alloying concepts are discussed. Optimal control of the CSP process allows the production of higher-strength hot-rolled steel grades with a fine-grain acicular-ferritic/bainitic microstructure. Hot strip produced in this way is characterized by a high toughness at low temperatures. In a drop weight tear test, transition temperatures of up to -50 deg C can be achieved with a shear-fracture share of 85%. (author)

  5. Further application of the cleavage fracture stress model for estimating the T{sub 0} of highly embrittled ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sreenivasan, P.R.

    2016-02-15

    The semi-empirical cleavage fracture stress model (CFS), based on the microscopic cleavage fracture stress, s{sub f}, for estimating the ASTM E1921 reference temperature (T{sub 0}) of ferritic steels from instrumented impact testing of unprecracked Charpy V-notch specimens is further confirmed by test results for additional steels, including steels highly embrittled by thermal aging or irradiation. In addition to the ferrite-pearlite, bainitic or tempered martensitic steels (which was examined earlier), acicular or polygonal ferrite, precipitation-strengthened or additional simulated heat affected zone steels are also evaluated. The upper limit for the applicability of the present CFS model seems to be T{sub 41J} ∝160 to 170 C or T{sub 0} or T{sub Qcfs} (T{sub 0} estimate from the present CFS model) ∝100 to 120 C. This is not a clear-cut boundary, but indicative of an area of caution where generation and evaluation of further data are required. However, the present work demonstrates the applicability of the present CFS model even to substantially embrittled steels. The earlier doubts expressed about T{sub Qcfs} becoming unduly non-conservative for highly embrittled steels has not been fully substantiated and partly arises from the necessity of modifications in the T{sub 0} evaluation itself at high degrees of embrittlement suggested in the literature.

  6. Martensitic — Ferritic steels and their applications

    Science.gov (United States)

    Prnka, T.; Mazanets, K.; Zil'bernagel', A.

    1981-07-01

    Biphase low-carbon low-alloy steels (0.065-0.13% C, 1-2% Mn, 0.3-1.5% Si, 0.5% Cr, with 0.1% V or 0.1-0.4% Mo) are presently used in the form of strips up to 10 mm thick for parts manufactured by cold pressing (automobile parts such as disks and wheel rims).

  7. Fatigue and fracture behavior of low alloy ferritic forged steels

    International Nuclear Information System (INIS)

    Chaudhry, V.; Sharma, A.K.; Muktibodh, U.C.; Borwankar, Neeraj; Singh, D.K.; Srinivasan, K.N.; Kulkarni, R.G.

    2016-01-01

    Low alloy ferritic steels are widely used in nuclear industry for the construction of pressure vessels. Pressure vessel forged low alloy steels 20MnMoNi55 (modified) have been developed indigenously. Experiments have been carried out to study the Low Cycle Fatigue (LCF) and fracture behavior of these forged steels. Fully reversed strain controlled LCF testing at room temperature and at 350 °C has been carried out at a constant strain rate, and for different axial strain amplitude levels. LCF material behavior has been studied from cyclic stress-strain responses and the strain-life relationships. Fracture behavior of the steel has been studied based on tests carried out for crack growth rate and fracture toughness (J-R curve). Further, responses of fatigue crack growth rate tests have been compared with the rate evaluated from fatigue precracking carried out for fracture toughness (J-R) tests. Fractography of the samples have been carried out to reveal dominant damage mechanisms in crack propagation and fracture. The fatigue and fracture properties of indigenously developed low alloy steel 20MnMoNi55 (modified) steels are comparable with similar class of steels. (author)

  8. 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)

  9. 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

  10. 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.

  11. Austenitic stainless steel-to-ferritic steel transition joint welding for elevated temperature service

    International Nuclear Information System (INIS)

    King, J.F.; Goodwin, G.M.; Slaughter, G.M.

    1978-01-01

    Transition weld joints between ferritic steels and austenitic stainless steels are required for fossil-fired power plants and proposed nuclear plants. The experience with these dissimilar-metal transition joints has been generally satisfactory, but an increasing number of failures of these joints is occurring prematurely in service. These concerns with transition joint service history prompted a program to develop more reliable joints for application in proposed nuclear power plants

  12. Martensitic/ferritic super heat-resistant 650 C steels

    Energy Technology Data Exchange (ETDEWEB)

    Agamennone, R.; Blum, W. [IWW-LS1, Univ. Erlangen-Nuernberg, Erlangen (Germany); Berger, C.; Granacher, J.; Scholz, A.; Wang, Y. [IfW, TU Darmstadt, Darmstadt (Germany); Ehlers, J.; Ennis, P.J.; Quadakkers, J.W.; Singheiser, L. [IWV2, Forschungszentrum Juelich GmbH, Juelich (Germany); Inden, G.; Knezevic, V.; Sauthoff, G.; Vilk, J. [Max-Planck-Inst. fuer Eisenforschung GmbH, Duesseldorf (Germany)

    2002-07-01

    World-wide demand for higher steam parameters of ultra super critical (USC) Power Plants has led to developments of new materials with improved high-temperature properties. A new project aims at new ferritic creep-resistant steels for application at 650 C and 300 bar. The critical issues are improvement of long-term creep strength as well as oxidation and corrosion resistance. The aim of the present research is to design new super heat-resistant martensitic/ferritic 9-12%Cr steels using basic principles and concepts of physical metallurgy, to test and optimise model alloys and to investigate and clarify their behaviour under long-term creep conditions with emphasis on microstructural stability and corrosion resistance. Model alloys have been designed, produced and tested with respect to deformation and corrosion. The design of model alloys has been supported by theoretical simulations and transmission electron microscopy investigations. First results for various modified 12%Cr model steels are reported, which indicate a high potential for reaching sufficient creep and corrosion resistance at 650 C. The work with further optimisation of composition and microstructure is in progress. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-30

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

  14. 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.

  15. Foucault current testing of ferritic steel fuel cans

    International Nuclear Information System (INIS)

    Stossel, A.

    1984-10-01

    The analysis of impedance involved by a Foucault current test of ferritic steel tubes, is quite different from the classical analysis which refers to non-magnetic tubes; more particularly, volume defects are considered as magnetic anomalies. Contrarily to current instructions which recommend to test the product in a satured magnetic state, it is very interesting to work with a continuous energizing field, comparatively low, corresponding to a sequenced magnetization, of which value is obtained according to the magnetic structure of the product. This analysis is useful when testing fast reactor fuel cans [fr

  16. Material physical properties of 12 chromium ferritic steel

    International Nuclear Information System (INIS)

    Ando, Masanori; Wakai, Takashi; Aoto, Kazumi

    2003-09-01

    High chromium ferritic steel is an attractive candidate for structural material of the next Fast Breeder Reactor, since both of thermal properties and high temperature strength of the steel are superior to those of conventional austenitic stainless steels. In this study, physical properties of 12Cr steels are measured and compared to those obtained in the previous studies to discuss about stochastic dispersions. The effect of measurement technique on Young's modulus and the influence of the specimen size on coefficient of thermal expansion are also investigated. The following conclusions are obtained. (1) Young's modulus of 12Cr steels obtained in this study tends to larger than those obtained in the previous studies especially in high temperature. Such a discrepancy is resulted from the difference in measurement technique. It was clarified that Young's modulus obtained by free vibration method is more adequate those obtained by the cantilever characteristic vibration method. Therefore, the authors recommend using the values obtained by free vibration method as Young's modulus of 12Cr steels. (2) Both instant and mean coefficient of thermal expansion of 12Cr steels obtained in this study is in a good agreement with those obtained in the previous studies. However, the obviously different values are obtained from the measurement by large size specimens. Such a discrepancy is resulted from heterogeneous during heating process of the specimens. Therefore, the authors recommend using the values obtained by φ4 x 20 mm specimens as instant and mean coefficient of thermal expansion of 12Cr steels. (3) Specific heat of 12Cr steels obtained in this study agree with those obtained in the previous studies with a few exceptions. (4)Thermal conductivity of 12Cr steels obtained in this study agree with those obtained in the previous studies. (5) It was confirmed that instant and mean coefficient of thermal expansion, density, specific heat and thermal conductivity of 12Cr steels

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

    Science.gov (United States)

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

    2010-03-16

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

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

    Science.gov (United States)

    2012-03-01

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

  19. Effect of residual stress on fatigue crack propagation at 200 C in a welded joint austenitic stainless steel - ferritic steel

    International Nuclear Information System (INIS)

    Zahouane, A.I.; Gauthier, J.P.; Petrequin, P.

    1988-01-01

    Fatigue resistance of heterogeneous welded joints between austenitic stainless steels and ferritic steels is evaluated for reactor components and more particularly effect of residual stress on fatigue crack propagation in a heterogeneous welded joint. Residual stress is measured by the hole method in which a hole is drilled through the center of a strain gage glued the surface of the materials. In the non uniform stress field a transmissibility function is used for residual stress calculation. High compression residual stress in the ferritic metal near the interface ferritic steel/weld slow down fatigue crack propagation. 5 tabs., 15 figs., 19 refs [fr

  20. Alloys influence in ferritic steels with hydrogen attack

    International Nuclear Information System (INIS)

    Moro, L; Rey Saravia, D; Lombardich, J; Saggio, M; Juan, A; Blanco, J

    2003-01-01

    Materials exposed to a corrosive environment and high temperatures, are associated with a decrease of their mechanical properties and embitterment.At room temperatures atomic hydrogen diffuses easily through metals structure, it accumulates in lattice defects forming molecular hydrogen and generating cracking due to internal stresses.Under high temperatures the phenomenon is more complex.The steels in these conditions present different structures of precipitates, that the change under creep conditions period.In this work it is determined the influence of Cr and V alloys, the changes of ferritic steel resistance in a corrosive environment and high temperatures.1.25 Cr 1 Mo 0.25 V and 2.25Cr 1 Mo under different loads and temperatures previously attacked by hydrogen environment.The hydrogen is induced by the electrolytic technique, optimizing the choice of temperatures, current density, electrolyte, etc. In order to control an adequate cathode charge, a follow up procedure is carried out by electronic barrier microscopy.After the attack, the material is settled at room temperatures for certain period of time, to allow the hydrogen to leave and evaluate the residual damage.Creep by torsion assays, under constant load and temperature is used as an experimental technique.With the outcome data curves are drawn in order to study the secondary creep rate, with the applied load and temperature, determining the value of stress exponent n and the activation energy Q.Comparing to equal assays to the same ferritic steels but non attacked by hydrogen, these values allows the prediction of microstructure changes present during these tests

  1. 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.

  2. 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.

  3. Irradiation embrittlement of pressure vessel steels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Vacek, M.

    1975-01-01

    A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

  4. Co-Sintering behaviour of zirconia-ferritic steel composites

    Directory of Open Access Journals (Sweden)

    Alexander Michaelis

    2016-08-01

    Full Text Available The combination of metallic and ceramic materials allows the combination of positive properties of both and can be applied in various industrial fields. At the moment, the deployment of these composites faces difficult and complex manufacturing. One attempt, which offers a short process route and a high degree of flexibility regarding design is a combined shaping (co-shaping with a combined sintering (co-sintering. The article will show co-sintering results of different metal-ceramic symmetric and asymmetric multi-layered tapes, consisting of yttria stabilized zirconia combined with a ferritic iron chromium steel. Focus is on the densification and co-sintering behaviour of ceramic layers depending on the sintering behaviour of metallic layers. Co-sintered composites were characterized by field emission scanning electron microscopy, x-ray diffraction measurements and in terms of adhesive tensile strength.

  5. 78 FR 63517 - Control of Ferrite Content in Stainless Steel Weld Metal

    Science.gov (United States)

    2013-10-24

    ... steel welds, the original version of this guide, Safety Guide 31, ``Control of Stainless Steel Welding... NUCLEAR REGULATORY COMMISSION [NRC-2012-0231] Control of Ferrite Content in Stainless Steel Weld.... Nuclear Regulatory Commission (NRC) is issuing a revision to Regulatory Guide (RG) 1.31, ``Control of...

  6. In-situ TEM and ion irradiation of ferritic materials

    International Nuclear Information System (INIS)

    Kirk, M.A.; Baldo, P.M; Liu, A.C.Y.; Ryan, E.A.; Birtcher, R.C.; Yao, Z.; Xu, S.; Jenkins, M.L.; Hernandez-Mayoral, M.; Kaoumi, D.; Motta, A.T.

    2009-01-01

    The intermediate voltage electron microscope-tandem user facility in the Electron Microscopy Center at Argonne National Laboratory is described. The primary purpose of this facility is electron microscopy with in situ ion irradiation at controlled sample temperatures. To illustrate its capabilities and advantages a few results of two outside user projects are presented. The motion of dislocation loops formed during ion irradiation is illustrated in video data that reveals a striking reduction of motion in Fe-8%Cr over that in pure Fe. The development of extended defect structure is then shown to depend on this motion and the influence of nearby surfaces in the transmission electron microscopy thin samples. In a second project, the damage microstructure is followed to high dose (200 dpa) in an oxide dispersion strengthened ferritic alloy at 500 C, and found to be qualitatively similar to that observed in the same alloy neutron irradiated at 420 C.

  7. 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)

  8. In-Pile creep rupture properties of ODS ferritic steel claddings

    International Nuclear Information System (INIS)

    Kaito, T.; Uwaba, T.; Mizuta, S.; Ito, C.; Kagota, E.; Kitamura, R.; Ohtsuka, S.; Inoue, M.; Asayama, T.; Ukai, S.; Furukawa, T.; Inoue, T.

    2007-01-01

    Full text of publication follows: Oxide Dispersion Strengthened (ODS) ferritic steels are the most prospective material for both advanced sodium cooled fast breeder reactor (SFR) fuels and fusion reactor components. In the SFR core, superior radiation resistance and high temperature capability are essential for fuel pin cladding tubes which will be exposed to high neutron doses up to 250 dpa relevant to peak burnup of 250 GWd/t in high temperature flowing sodium ranging from 673 K to 973 K. Japan Atomic Energy Agency (JAEA) has been developing two types of ODS steels, which are 9Cr-ODS steel (9Cr-0.13C-2W-0.2Ti-0.35Y 2 O 3 ) and 12Cr-ODS steel (12Cr-0.05C-2W-0.3Ti-0.25Y 2 O 3 ). For the cladding tubes, internal creep rupture strength is one of the most important properties; for example, internal pressure gradually increases with burnup and finally reaches at 120 MPa in the highest burnup fuel pins. In order to examine irradiation effect on creep rupture strength of the ODS steels, an in-pile internal creep rupture test has been conducted in the experimental fast reactor JOYO using Material Testing Rig with Temperature Control (MARICO). Twenty-four pressurized tube specimens made from both 9Cr- and 12Cr-ODS steels have been irradiated at temperatures of 943 K, 973 K and 1023 K up to 20 dpa. Hoop stress for each specimen was varied with filling helium gas volume to attain predetermined pressure ranging from 45 MPa to 155 MPa at desired test temperature. Small amount of xenon and krypton mixed gas with unique isotopic composition was also filled into each specimen and released into cover gas systems after creep rupture in order to identify its creep rupture time by analyzing gas species by means of Laser Resonance Ionization Mass Spectrometry (RIMS). In MARICO test, 14 creep ruptures have been detected by the end of February 2007. Up to now, no irradiation effect on creep rupture strength of the ODS steels has been distinguished. This indicates that nanometer size

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.

    1996-12-31

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

  10. Modeling of Ni Diffusion Induced Austenite Formation in Ferritic Stainless Steel Interconnects

    DEFF Research Database (Denmark)

    Chen, Ming; Molin, Sebastian; Zhang, L.

    2015-01-01

    Ferritic stainless steel interconnect plates are widely used in planar solid oxide fuel cell (SOFC) or electrolysis cell (SOEC) stacks. During stack production and operation, nickel from the Ni/YSZ fuel electrode or from the Ni contact component diffuses into the IC plate, causing transformation...... of the ferritic phase into an austenitic phase in the interface region. This is accompanied with changes in volume and in mechanical and corrosion properties of the IC plates. In this work, kinetic modeling of the inter-diffusion between Ni and FeCr based ferritic stainless steel was conducted, using the CALPHAD...

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

    Science.gov (United States)

    2010-10-01

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

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

    Science.gov (United States)

    2010-10-01

    ... VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-20 Low temperature operation—ferritic steels with properties enhanced by heat treatment (modifies UHT-5(c), UHT-6, UHT-23, and UHT-82... 46 Shipping 2 2010-10-01 2010-10-01 false Low temperature operation-ferritic steels with...

  13. Effect of the damage by radiation on the reference temperature T0 of ferritic steel

    International Nuclear Information System (INIS)

    Villanueva O, A.

    2004-01-01

    Presently work studies the effect that produces the irradiation in ferritic steels, on the reference temperature T 0 (intrinsic characteristic of the fracture tenacity in the area of ductile-fragile transition), applying the approach of the Master curve that is based on the norm Astm E-1921. For it it was elaborated a methodology and procedure for test tubes type Charpy according to the standard before mentioned. Due to the ferritic steels are used mainly in pressure vessels to the reactor (RPV) of nuclear power plants; in the samples it was simulated the effect of the damage for irradiation through a thermal treatment that induced the precipitation of the carbides and sulfurs in the limits of grain (one of the modifications suffered in the irradiated materials); it was made a comparison later with material samples in initial state (without thermal treatment), used as witness sample, by means of assays of fracture mechanics, specifically flexion in three points; this way with it to observe the effect of the damage for irradiation in the reference temperature (T 0 ). This temperature (T 0 ) it is a very important parameter in the mechanical property of the material called fracture tenacity; which at the moment gives the rule for the verification of structural integrity of the RPV. As a result of this it was observed an increase in the reference temperature in the material in fragilezed state with respect to the initial state of 31.75 C. They were carried out metallographic analysis and fractographs of the assayed surface finding carbide inclusions and sulfurs that in theory of the Master Curve they are initiators of cracks and of a possible catastrophic flaw of the material. At the moment the Division of Scientific Investigation of the ININ is carrying out activities in the Nucleo electric Central of Laguna Verde (CNLV) related with the program of surveillance of the materials of the vessel of the unit 2, as well as projects of structural integrity financed by the

  14. 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

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

    Science.gov (United States)

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

    2015-12-01

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

  16. In situ 3D monitoring of corrosion on carbon steel and ferritic stainless steel embedded in cement paste

    KAUST Repository

    Itty, Pierre-Adrien

    2014-06-01

    In a X-ray microcomputed tomography study, active corrosion was induced by galvanostatically corroding steel embedded in cement paste. The results give insight into corrosion product build up, crack formation, leaching of products into the cracks and voids, and differences in morphology of corrosion attack in the case of carbon steel or stainless steel reinforcement. Carbon steel was homogeneously etched away with a homogeneous layer of corrosion products forming at the steel/cement paste interface. For ferritic stainless steel, pits were forming, concentrating the corrosion products locally, which led to more extensive damage on the cement paste cover. © 2014 Elsevier Ltd.

  17. ROLE OF STRUCTURE IS IN THE PROCESS OF FERRITIC-PEARLITIC STEEL EROSION

    Directory of Open Access Journals (Sweden)

    O. A. Kuzin

    2010-09-01

    Full Text Available The results of study of influence of structure on mechanical properties and behavior of ferrite-perlite steels under the action of contact loads are presented. It is shown that the formation of the widmanstatten pattern has a negative impact on the performance of steels under static loads but a positive effect on their durability.

  18. CYCLIC RECRYSTALLIZATION OF FERRITE IN HOT-ROLLED LOW-CARBON SHEET STEEL WITH STRUCTURETEXTURAL HETEROGENEITY

    Directory of Open Access Journals (Sweden)

    A. M. Nesterenko

    2009-01-01

    Full Text Available It is determined that in the process of soaking at subcritical temperature 680 °C in hot-rolled rolling of low-carbon steel 08 ps recrystallization is developed with heterogeneous fu ll repeat change of the steel ferrite change by its section.

  19. 77 FR 60478 - Control of Ferrite Content in Stainless Steel Weld Metal

    Science.gov (United States)

    2012-10-03

    ... Metal AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... draft regulatory guide (DG), DG-1279, ``Control of Ferrite Content in Stainless Steel Weld Metal.'' This... stainless steel weld metal. Revision 4 updates the guide to remove references to outdated standards and to...

  20. Stability under irradiation of a fine dispersion of oxides in a ferritic matrix; Stabilite sous irradiation de particules d'oxydes finement dispersees dans des alliages ferritiques

    Energy Technology Data Exchange (ETDEWEB)

    Monnet, I

    1999-07-01

    Oxide dispersion strengthened (ODS) ferritic-martensitic steels are being considered for high temperature, high fluence nuclear applications, like fuel pin cladding in Fast Breeder Reactors. ODS alloys offer improved out of pile strength characteristics at temperature above 550 deg.C and ferritic-martensitic matrix is highly swelling resistant. A clad in an ODS ferritic steel, call DY (Fe-13Cr-1,5Mo+TiO{sub 2}+Y{sub 2}O{sub 3}) has been irradiated in the experimental reactor Phenix. Under irradiation oxide dissolution occurs. Microstructural observations indicated that oxide evolution is correlated with the dose and consist in four phenomena: the interfaces of oxide particles with the matrix become irregular, the uniform distribution of the finest oxide (< 20 nm) disappear, the modification of oxide composition, and a halo of fine oxides appear around the larger oxides. The use of such a material requires a study of oxide stability under irradiation, since the oxide particles provide the desired mechanical properties. The study is based on two types of alloys, the DY and four ferritic steels Fe-9Cr-1Mo reinforced by Al{sub 2}O{sub 3}, Y{sub 2}O{sub 3}, MgO or MgAl{sub 2}O{sub 4}. These materials were irradiated with charged particles in order to gain a better understanding of the mechanisms of dissolution. Irradiation with 1 MeV Helium does not induce any modification, neither in the chemical modification of the particles nor in their spatial and size distribution. Since most of the energy of helium ions is lost by inelastic interaction, this result proves that this kind of interaction does not induce oxide dissolution. Irradiation with 1 MeV or 1.2 MeV electrons leads to a significant dissolution with a radius decrease proportional to the dose. These experiments prove that oxide dissolution can be induced by Frenkel pairs alone, provided that metallic atoms are displaced. The comparison between irradiation with ions (displacements cascades) and electrons (Frenkel

  1. Development and Application of High-Cr Ferritic Stainless Steels as Building Exterior Materials

    International Nuclear Information System (INIS)

    Kim, Yeong H.; Lee, Yong H.; Lee, Yong D.

    2008-01-01

    Stainless Steels have been widely used as a building exterior materials in Asian countries for the last decade. It is required for the materials in this field to have an aesthetic appearance,a relatively high strength, and an excellent corrosion resistance. Other metallic materials such as copper, aluminum, and carbon steels have been also used as the exterior materials. Considering the cost of maintenance, stainless steel, having the outstanding corrosion resistance, is replacing other materials in the several parts in the building exteriors. Ferritic stainless steel has been applied as the roofing materials because its thermal expansion is much smaller than that of austenitic stainless steel. Therefore, it is suitable for the large-scale construction such as airport terminal, convention center, and football stadium. To improve the corrosion resistance of the ferritic stainless steels, the modification of alloy composition has been studied to develop new grade materials and the progress in the surface technology has been introduced. Corrosion properties, of these materials were evaluated in the laboratory and in the field for longer than two years. High-Cr ferritic stainless steel showed excellent corrosion resistance to the atmospheric environments. In the region close to the sea, the corrosion resistance of high-Cr ferritic stainless steel was much superior to that of other materials, which may prove this steel to be the appropriate materials for the construction around seashore. In some of the large constructions around seashore in South Korea, high-Cr ferritic stainless steels have been used as the building exterior materials for six years

  2. 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...... lamellar structure with a + duplex fibre texture develops in both the modified 9Cr-1Mo steel and PM2000 during DPD to high strains. The strength is improved significantly, but the thermal stability is largely reduced. A very pronounced orientation dependent recovery and recrystallization take...... place, when both steels after DPD are annealed. Both oriented nucleation and oriented growth of oriented lamellae are demonstrated to account for such an orientation dependence. The underlying mechanisms are discussed, including the differences in stored energy, structural variation, and recovery...

  3. Modifying ferritic stainless steels for solid oxide fuel cell applications

    Science.gov (United States)

    Laney, Scot Jason

    2007-12-01

    One of the most important problem areas associated with the solid oxide fuel cells is selection of a cost effective material for use as the interconnect component of the cell. Metals are now being considered as materials for this component, with ferritic stainless steels being the leading candidate. This work evaluates methods to combat the problem areas, namely rapid growth rate and vaporization of the oxide scale, that hinder the use of these materials. Oxidation experiments have been performed in dry and wet single atmosphere exposures as well as a dual environment exposure to simulate the conditions in a working SOFC. Measurements of the electrical properties of the oxides that formed were also performed. Commercial alloys, E-Brite and Crofer 22APU, were tested to form a baseline and resultant oxidation and electrical behaviors match those found in the literature. Isothermal oxidation tests for short exposure times have also led to a possible mechanism for the formation of the MnCr2O4 layer on Crofer. All of these tests were then replicated on a series of experimental Fe-22Cr-XTi (X=0-4) alloys. These alloys are shown to form a rutile layer analogous to the MnCr2O4 layer on Crofer. While this layer does prevent some chromia vaporization, the consequences due to the presence of Ti in the chromia include increased growth rate, decreased resistivity, extensive internal oxidation and nitridation of Ti, and a change of the growth direction of the chromia. The alloys containing ˜2--3 wt%Ti appear to offer the best combination of oxidation, electrical, and mechanical properties. Coatings of lanthanum chromites and ferrites were also tested and shown to be very sensitive to exposure condition, resulting in the formation of pores, and to coating thickness, where thicker coatings are subject to cracking. Finally, reactive element oxide doping was attempted to slow the oxide growth rate for E-Brite (CeO2 doping) and for the Fe-Cr-Ti alloys (CeO 2 and La2O3 doping). A

  4. Swelling in simple ferritic alloys irradiated to high fluence

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Residual stress distribution in ferritic to austenitic steel joints made by laser welding,

    OpenAIRE

    Iordachescu, Mihaela; Ruiz Hervías, Jesús; Luzin, V.; Scutelnicu, Elena; Valiente Cancho, Andrés; Ocaña Moreno, José Luis

    2013-01-01

    In this study, autogenous laser welding was used to join thin plates of low carbon ferritic and austenitic stainless steel. Due to the differences in the thermo-physical properties of base metals, this kind of weld exhibits a complex microstructure, which frequently leads to an overall loss of joint quality. Four welded samples were prepared by using different sets of processing parameters, with the aim of minimizing the induced residual stress field. The dissimilar austenitic-ferritic joints...

  6. Ferritic stainless steel composite slabs : Experimental study of longitudinal shear transfer

    OpenAIRE

    Ferrer Ballester, Miquel; Marimón Carvajal, Federico; Arrayago Luquin, Itsaso; Mirambell Arrizabalaga, Enrique

    2014-01-01

    The objective of this work is to carry out the procedure described in Eurocode 4 to evaluate the longitudinal shear transfer capability of conventional steel sheeting open-rib profile with embossments, usually rolled in conventional galvanized steel, being rolled now in ferritic stainless steel 1.4003 alloy. Finally, the results of both composite floor slabs are compared. Two methodologies have been used to evaluate the longitudinal shear resistance in composite slabs, the m-k method and t...

  7. Characterization of friction stir welded joint of low nickel austenitic stainless steel and modified ferritic stainless steel

    Science.gov (United States)

    Mondal, Mounarik; Das, Hrishikesh; Ahn, Eun Yeong; Hong, Sung Tae; Kim, Moon-Jo; Han, Heung Nam; Pal, Tapan Kumar

    2017-09-01

    Friction stir welding (FSW) of dissimilar stainless steels, low nickel austenitic stainless steel and 409M ferritic stainless steel, is experimentally investigated. Process responses during FSW and the microstructures of the resultant dissimilar joints are evaluated. Material flow in the stir zone is investigated in detail by elemental mapping. Elemental mapping of the dissimilar joints clearly indicates that the material flow pattern during FSW depends on the process parameter combination. Dynamic recrystallization and recovery are also observed in the dissimilar joints. Among the two different stainless steels selected in the present study, the ferritic stainless steels shows more severe dynamic recrystallization, resulting in a very fine microstructure, probably due to the higher stacking fault energy.

  8. Characterisation of high-temperature damage mechanisms of oxide dispersion strengthened (ODS) ferritic steels

    International Nuclear Information System (INIS)

    Salmon-Legagneur, Hubert

    2017-01-01

    The development of the fourth generation of nuclear power plants relies on the improvement of cladding materials, in order to achieve resistance to high temperature, stress and irradiation dose levels. Strengthening of ferritic steels through nano-oxide dispersion allows obtaining good mechanical strength at high temperature and good resistance to irradiation induced swelling. Nonetheless, studies available from open literature evidenced an unusual creep behavior of these materials: high anisotropy in time to rupture and flow behavior, low ductility and quasi-inexistent tertiary creep stage. These phenomena, and their still unclear origin are addressed in this study. Three 14Cr ODS steels rods have been studied. Their mechanical behavior is similar to those of other ODS steels from open literature. During creep tests, the specimens fractured by through crack nucleation and propagation from the lateral surfaces, followed by ductile tearing once the critical stress intensity factor was reached at the crack tip. Tensile and creep properties did not depend on the chemical environment of specimens. Crack propagation tests performed at 650 C showed a low value of the stress intensity factor necessary to start crack propagation. The cracks followed an intergranular path through the smaller-grained regions, which partly explains the anisotropy of high temperature strength. Notched specimens have been used to study the impact of the main loading parameters (deformation rate, temperature, stress triaxiality) on macroscopic crack initiation and stable propagation, from the central part of the specimens. These tests allowed revealing cavities created during high temperature loading, but unexposed to the external environment. These cavities showed a high chemical reactivity of the free surfaces in this material. The performed tests also evidenced different types of grain boundaries, which presented different damage development behaviors, probably due to differences in local

  9. Modeling of Ni Diffusion Induced Austenite Formation in Ferritic Stainless Steel Interconnects

    DEFF Research Database (Denmark)

    Chen, Ming; Alimadadi, Hossein; Molin, Sebastian

    2017-01-01

    Ferritic stainless steel interconnect plates are widely used in planar solid oxide fuel cell and electrolysis cell stacks. During stack production and operation, nickel from the Ni/yttria stabilized zirconia fuel electrode or from the Ni contact component layer diffuses into the interconnect plate......, causing transformation of the ferritic phase into an austenitic phase in the interface region. This is accompanied with changes in volume, and in mechanical and corrosion properties of the interconnect plates. In this work, kinetic modeling of the inter-diffusion between Ni and FeCr based ferritic...

  10. Formation of oxides particles in ferritic steel by using gas-atomized powder

    International Nuclear Information System (INIS)

    Liu Yong; Fang Jinghua; Liu Donghua; Lu Zhi; Liu Feng; Chen Shiqi; Liu, C.T.

    2010-01-01

    Oxides dispersion strengthened (ODS) ferritic steel was prepared by using gas-atomized pre-alloyed powder, without the conventional mechanical alloying process. By adjusting the volume content of O 2 in the gas atmosphere Ar, the O level in the ferritic powder can be well controlled. The O dissolves uniformly in the ferritic powder, and a very thin layer of oxides forms on the powder surface. After hot deformation, the primary particle boundaries, which retain after sintering, can be disintegrated and near fully dense materials can be obtained. The oxide layer on the powder surface has a significant effect on the microstructural evolution. It may prevent the diffusion in between the primary particles during sintering, and may dissolve and/or induce the nucleation of new oxides in the ferritic matrix during recrystallization. Two kinds of oxide particles are found in the ferritic steel: large (∼100 nm) Ti-rich and fine (10-20 nm) Y-Ti-rich oxides. The hardness of the ferritic steel increases with increasing annealing temperatures, however, decreases at 1400 deg. C, due to the coarsening of precipitates and the recrystallization microstructure.

  11. Microstructure and tensile properties of yttrium nitride dispersion-strengthened 14Cr–3W ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Liqing [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); School of Mechanical and Mining Engineering, University of Queensland, Brisbane 4067, QLD (Australia); Liu, Zuming, E-mail: lzm@csu.edu.cn [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Chen, Shiqi; Guo, Yang [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China)

    2015-12-15

    Highlights: • Innovative nano yttrium nitride dispersion strengthened steels were fabricated. • Higher content of additives accelerate the steel-ceramic powder milling process more. • Steel with high content (3%) of YN dispersoids can obtain good performance at 500 °C. - Abstract: 14Cr–3W ferritic steel powders were mechanically milled with microscale yttrium nitride (YN) particles to fabricate particle dispersion-strengthened ferritic steels. After hot consolidation and annealing, the steel matrix was homogeneously dispersed with nano-scale YN particles. The steel containing 0.3 wt.% YN particles exhibited a yield strength of 1445 MPa at room temperature. Its total elongation was 10.3%, and the fracture surface exhibited mixed ductile and quasi-cleavage fracture morphologies. The steel with a much higher content of YN particles (3 wt.%) in its matrix was much stronger (1652 MPa) at room temperature at the cost of ductility. In particular, it exhibited a high yield strength (1350 MPa) with applicable ductility (total elongation > 10%) at 500 °C. This study has developed a new kind of reinforcement particle to fabricate high-performance ferritic steels.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  13. Passivation behavior of a ferritic stainless steel in concentrated alkaline solutions

    Directory of Open Access Journals (Sweden)

    Arash Fattah-alhosseini

    2015-10-01

    Full Text Available The passivation behavior of AISI 430 ferritic stainless steel was investigated in concentrated alkaline solutions in relation to several test parameters, using electrochemical techniques. Increasing solution pH (varying from 11.5 to 14.0 leads to an increase in the corrosion rate of the alloy. Mott–Schottky analysis revealed that passive films formed on AISI 430 ferritic stainless steel behave as n-type semiconductor and the donor densities increased with pH. Electrochemical impedance spectroscopy (EIS results showed that the reciprocal capacitance of the passive film is directly proportional to its thickness, which decreases with pH increase. The results revealed that for this ferritic stainless steel in concentrated alkaline solutions, decreasing the solution pH offers better conditions for forming passive films with higher protection behavior, due to the growth of a much thicker and less defective film.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-11-02

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

  15. Fractographic correlations with mechanical properties in ferritic martensitic steels

    Science.gov (United States)

    Das, Arpan; Chakravartty, Jayanta Kumar

    2017-12-01

    The ultimate continuum of a material is nothing but the process called fracture. Fracture surface retains the imprint of the entire deformation history undergone in a material. Hence, it is possible to derive the approximate deformation and fracture properties of a material from a systematic fracture feature analysis. There has been large volume of literature available in the open domain correlating different mechanical and fracture responses of reduced activation ferritic martensitic grade steels under various testing conditions/circumstances with corresponding microstructural interpretation. There has been no such literature available to establish the relationship between the two-dimensional fracture geometry/topography with its corresponding deformation and mechanical properties of the material as a function of testing temperature, which has been the primary aim in the current investigation. A comprehensive literature survey has been carried out to realize this fact. In order to establish the above hypothesis, many tensile experiments were carried out at constant strain rate by systematic variation of the test temperature. The initial void volume fraction or the inclusion content of material was kept unaltered and the test temperature has been varied orderly on different multiple specimens to vary the deformation-induced nucleation sites of micro voids (i.e. different carbides, phase interfaces, dislocation pile up etc), which results in a change of fracture topography under uniaxial tensile deformation. A conventional metallographic technique followed by optical microscopy has been employed to understand the basic morphologies and characteristics of the alloy exposed at different temperatures. Fractographic investigation of the broken tensile specimens at various temperatures is carried out to measure the fracture features by using quantitative fractography on representative scanning electron fractographs through image processing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pasebani, Somayeh [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Charit, Indrajit, E-mail: icharit@uidaho.edu [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Wu, Yaqiao; Burns, Jatuporn; Allahar, Kerry N.; Butt, Darryl P. [Department of Materials Science and Engineering, Boise State University, Boise, ID 83725 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Cole, James I. [Idaho National Laboratory, Idaho Falls, ID 83401 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Alsagabi, Sultan F. [Department of Chemical and Materials Engineering, University of Idaho, Moscow, ID 83844 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States)

    2016-03-15

    A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La{sub 2}O{sub 3} (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and1050 °C for 45 min. Significant densification occurred at temperatures greater than 950 °C with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 °C, the number density of Cr–Ti–La–O-enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 × 10{sup 24} m{sup −3}. The La + Ti:O ratio was close to 1 after SPS at 950 and 1050 °C; however, the number density of nanoclusters decreased at 1050 °C. With SPS above 950 °C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.

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

    Science.gov (United States)

    Rakowski, James M.

    2013-09-10

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

  18. 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.

  19. Behavior of the elements in the mechanically alloyed and cast ferritic steels and a type 316 stainless steel in a flowing sodium environment

    International Nuclear Information System (INIS)

    Suzuki, T.; Mutoh, I.

    1988-01-01

    Sodium corrosion behavior of a mechanically alloyed ferritic steel, dispersion-strengthened with addition of Y 2 0 3 and Ti, two kinds of melted/cast ferritic steels and a Type 316 stainless steel was examined by using a non-isothermal sodium loop system, constructed of another Type 316 stainless steel, with a direct resistance electrical heater. The sodium conditions were 675 0 C, 4.0 m/s in velocity and 1-2 ppm oxygen concentration and a cumulative exposure time of the specimens was about 3000 h. The absorption of Ni and selective dissolution of Cr played an important role in the corrosion of the mechanically alloyed ferritic steel as in the case of the cast ferritic steels. However, the region of Ni absorption and Cr diminution was deeper than that of the cast ferritic steels. Peculiar finding for the mechanically alloyed ferritic steel was the corroded surface with irregularly shaped protuberance, that might be related with formation of sodium titanate, and the absorption of carbon and nitrogen to form carbide and nitride of titanium. It seems that these facts resulted in the irregular weight loss of the specimens, which depended on the downstream position and the cumulative exposure time. However, the tensile properties of the mechanically alloyed ferritic steel did not noticeably change by the sodium exposure

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

    International Nuclear Information System (INIS)

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

    1983-01-01

    A modified 9 Cr-1 Mo ferritic steel has been selected for development by the US Department of Energy as an alternative structural material for breeder reactor applications in which type 304 stainless steel or 2 1/4 Cr-1 Mo steel is currently being used. This developmental alloy is a modification (primarily through additions of niobium and vanadium) of a commercial 9 Cr-1 Mo steel already available in the US. A similar commercial alloy is available in the UK. Meanwhile, a 9 Cr-2 Mo alloy (EM12) is now used in France and a 12 Cr-1 Mo alloy (HT9) has been recommended for high temperature service in both Europe and the US. This paper compares the yield strength and ultimate tensile strength properties of the American modified 9 Cr-1 Mo steel with those of the American and British commercial 9 Cr-1 Mo steels. In addition, the creep rupture properties of the modified alloy are compared with those of the two commercial 9 Cr-1 Mo steels, 2 1/4 Cr-1 Mo steel, HT9, EM12, and type 304 stainless steel. The overall conclusion of the study is that the modified 9 Cr-1 Mo steel displays tensile and creep strengths matching or exceeding those of the other ferritic materials examined, and being at least comparable to those of type 304 stainless steel from room temperature to about 625 0 C. 15 references, 11 figures, 3 tables

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

    Science.gov (United States)

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

    2014-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  3. Use of ferritic steels in breeder reactors worldwide

    International Nuclear Information System (INIS)

    Patriarca, P.

    1983-01-01

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

  4. Ferrite and Perlite Hardening in Copper-Alloyed Steels and Irons

    Science.gov (United States)

    Bataev, I. A.; Stepanova, N. V.; Bataev, A. A.; Razumakov, A. A.

    2017-10-01

    The paper presents transmission electron microscopy (TEM) investigations of ɛ-copper formation in ferritic grains and perlitic colonies of irons and steels alloyed with copper. It is shown that copper-enriched inclusions substantially differ in size and shape. The most disperse are particles produced by decomposition of α-phase in iron due to oversaturated copper. The size of particles appeared after austenite decomposition is approximately an order of magnitude larger. After the formation of ɛ-copper particles in ɛ-phase, they incorporate both in ferrite and partially in cementite laminas during the formation of lamellar perlite. Fine particles of ɛ-copper locating inside ferritic grains and in ferritic layers in perlite, restrain the dislocation mobility and have an additional hardening effect on iron-carbon alloys.

  5. Kinetics modeling of delta-ferrite formation and retainment during casting of supermartensitic stainless steel

    DEFF Research Database (Denmark)

    Nießen, Frank; Tiedje, Niels Skat; Hald, John

    2017-01-01

    , equilibrium calculations and the Scheil model in Thermo-Calc, and validated by using microscopy and energy dispersive X-ray spectroscopy for chemical analysis on a cast ingot. The kinetics model showed that micro-segregation from solidification homogenizes within 2–3 s (70 °C) of cooling, and that retained δ......The kinetics model for multi-component diffusion DICTRA was applied to analyze the formation and retainment of δ-ferrite during solidification and cooling of GX4-CrNiMo-16-5-1 cast supermartensitic stainless steel. The obtained results were compared with results from the Schaeffler diagram......-ferrite originates from the incomplete transformation to austenite. The kinetics model predicted the measured amount of δ-ferrite and the partitioning of Cr and Ni reasonably well. Further, it showed that slower cooling for the investigated alloy leads to less retained δ-ferrite, which is in excellent agreement...

  6. Diffusion of Nickel into Ferritic Steel Interconnects of Solid Oxide Fuel/Electrolysis Stacks

    DEFF Research Database (Denmark)

    Molin, Sebastian; Chen, Ming; Bowen, Jacob R.

    2013-01-01

    diffusion of nickel from the Ni/YSZ electrode or the contact layer into the interconnect plate. Such diffusion can cause austenization of the ferritic structure and could possibly alter corrosion properties of the steel. Whereas this process has already been recognized by SOFC stack developers, only...... a limited number of studies have been devoted to the phenomenon. Here, diffusion of Ni into ferritic Crofer 22 APU steel is studied in a wet hydrogen atmosphere after 250 hours of exposure at 800 °C using Ni-plated (~ 10 micron thick coatings) sheet steel samples as a model system. Even after...... this relatively short time all the metallic nickel in the coating has reacted and formed solid solutions with iron and chromium. Diffusion of Ni into the steel causes formation of the austenite FCC phase. The microstructure and composition of the oxide scale formed on the sample surface after 250 hours is similar...

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

    Science.gov (United States)

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

    2017-01-01

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

  8. The creep properties of a low alloy ferritic steel containing an intermetallic precipitate dispersion

    International Nuclear Information System (INIS)

    Batte, A.D.; Murphy, M.C.; Edmonds, D.V.

    1976-01-01

    A good combination of creep rupture ductility and strength together with excellent long term thermal stability, has been obtained from a dispersion of intermetallic Laves phase precipitate in a non-transforming ferritic low alloy steel. The steel is without many of the problems currently associated with the heat affected zone microstructures of low alloy transformable ferritic steels, and can be used as a weld metal. Following suitable development to optimize the composition and heat treatment, such alloys may provide a useful range of weldable creep resistant steels for steam turbine and other high temperature applications. They would offer the unique possibility of easily achievable microstructural uniformity, giving good long term strength and ductility across the entire welded joint

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

    Science.gov (United States)

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

    2007-03-01

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

  10. Technological investigations of ferritic chromium steels with extra low concentrations of carbon and nitrogen

    International Nuclear Information System (INIS)

    Mueller, E.

    1976-01-01

    Ferritic chromium steels with extra low concentrations of C and N were examined by the Strauss Test and by other methods. The results show that resistance against intergranular corrosion can be reached by titanium concentrations ten times higher than the sum of C and N concentrations, whereas stability against pitting corrosion can be achieved by adding 1.3 % Mo. On grounds of the excellent corrosion resistance these steels will be of interest in nuclear technology

  11. 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.

  12. 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.

  13. Effect of hardness of martensite and ferrite on void formation in dual phase steel

    DEFF Research Database (Denmark)

    Azuma, M.; Goutianos, Stergios; Hansen, Niels

    2012-01-01

    The influence of the hardness of martensite and ferrite phases in dual phase steel on void formation has been investigated by in situ tensile loading in a scanning electron microscope. Microstructural observations have shown that most voids form in martensite by evolving four steps: plastic...

  14. Constraint Effects at Brittle Fracture Initiation in a Cast Ferritic Steel

    Czech Academy of Sciences Publication Activity Database

    Dlouhý, Ivo; Chlup, Zdeněk; Kozák, Vladislav

    č. 71 (2004), s. 873-883 ISSN 0013-7944 R&D Projects: GA AV ČR IAA2041003 Institutional research plan: CEZ:AV0Z2041904 Keywords : Cast ferritic steel * transition behaviour * fracture toughness Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 1.299, year: 2004 www.sciencedirect.com

  15. On the Nature of Internal Interfaces in Tempered Martensite Ferritic Steels

    Czech Academy of Sciences Publication Activity Database

    Dronhofer, A.; Pešička, J.; Dlouhý, Antonín; Eggeler, G.

    2003-01-01

    Roč. 94, č. 5 (2003), s. 511-520 ISSN 0044-3093 R&D Projects: GA ČR GA106/99/1172 Institutional research plan: CEZ:AV0Z2041904 Keywords : Tempered martensite ferritic steels * martensite variants * orientation imaging Subject RIV: JG - Metallurgy Impact factor: 0.637, year: 2003

  16. Overlaying of type 316 austenitic stainless steel with type 430 ferritic stainless steel

    International Nuclear Information System (INIS)

    Sujith, S.; Gill, T.P.S.

    1993-01-01

    Overlaying of type 316 austenitic stainless steel vessel with type 430 ferritic stainless is proposed for liquid magnesium service. The interface in this type of bimetallic configuration has been shown to be a cause for concern as it contains a hard and brittle martensite micro constituent which becomes susceptible to cracking under certain conditions. This study was carried out to standardize the welding conditions and characterise the interface in order to obtain sound overlay. Some tests were also conducted to simulate the elevated temperature service. The investigation has shown that the interface hardness approaches 400 VPN when no preheating is employed. However, in the preheated samples, appreciable reduction in the peak hardness was observed. This has been attributed to a decrease in the cooling rate of the clad metal with an increase in the preheating temperature which results in softening of the martensite. The minimum recommended preheat is 473 K. The samples exposed to thermal cycle tests to a peak temperature of 1223 K to simulate the service condition did not show any cracking at the interface after 20 cycles of testing. Therefore, this study has demonstrated the stability of the interface between type 316 and 430 stainless steels at the intended temperature of service. (author)

  17. 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

  18. 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.

  19. High strength ferritic alloy

    International Nuclear Information System (INIS)

    1977-01-01

    A high strength ferritic steel is specified in which the major alloying elements are chromium and molybdenum, with smaller quantities of niobium, vanadium, silicon, manganese and carbon. The maximum swelling is specified for various irradiation conditions. Rupture strength is also specified. (U.K.)

  20. 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

    2017-02-02

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

  1. 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

    2017-01-01

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

  2. Residual stress distribution in ferritic-austenitic steel joints made by laser welding

    OpenAIRE

    Iordachescu, Mihaela; Ocaña Moreno, José Luis

    2013-01-01

    The present investigation addresse the influence of laser welding process-ing parameters used for joining dis-similar metals (ferritic to austenitic steel), on the induced residual stress field. Welding was performed on a Nd:YAG laser DY033 (3300 W) in a continuous wave (CW), keyhole mode. The base metals (BM) employed in this study are AISI 1010 carbon steel (CS) and AISI 304L austenitic stainless steel (SS). Pairs of dissimilar plates of 200 mm x 45 mm x 3 mm were butt joined by laser weldi...

  3. SPEED DEPENDENCE OF ACOUSTIC VIBRATION PROPAGATION FROM THE FERRITIC GRAIN SIZE IN LOW-CARBON STEEL

    Directory of Open Access Journals (Sweden)

    I. A. Vakulenko

    2015-08-01

    Full Text Available Purpose. It is determining the nature of the ferrite grain size influence of low-carbon alloy steel on the speed propagation of acoustic vibrations. Methodology. The material for the research served a steel sheet of thickness 1.4 mm. Steel type H18T1 had a content of chemical elements within grade composition: 0, 12 % C, 17, 5 % Cr, 1 % Mn, 1, 1 % Ni, 0, 85 % Si, 0, 9 % Ti. The specified steel belongs to the semiferritic class of the accepted classification. The structural state of the metal for the study was obtained by cold plastic deformation by rolling at a reduction in the size range of 20-30 % and subsequent recrystallization annealing at 740 – 750 ° C. Different degrees of cold plastic deformation was obtained by pre-selection of the initial strip thickness so that after a desired amount of rolling reduction receives the same final thickness. The microstructure was observed under a light microscope, the ferrite grain size was determined using a quantitative metallographic technique. The using of X-ray structural analysis techniques allowed determining the level of second-order distortion of the crystal latitude of the ferrite. The speed propagation of acoustic vibrations was measured using a special device such as an ISP-12 with a working frequency of pulses 1.024 kHz. As the characteristic of strength used the hardness was evaluated by the Brinell’s method. Findings. With increasing of ferrite grain size the hardness of the steel is reduced. In the case of constant structural state of metal, reducing the size of the ferrite grains is accompanied by a natural increasing of the phase distortion. The dependence of the speed propagation of acoustic vibrations up and down the rolling direction of the ferrite grain size remained unchanged and reports directly proportional correlation. Originality. On the basis of studies to determine the direct impact of the proportional nature of the ferrite grain size on the rate of propagation of sound

  4. Steam oxidation of ferritic steels: kinetics and microestructure

    Directory of Open Access Journals (Sweden)

    Aríztegui, A.

    2000-06-01

    Full Text Available The ferritic 2.25Cr–1Mo steel has been subjected to isothermal and non-isothermal oxidation treatments in water steam at several temperatures ranging from 550 to 700 °C for up to 56 days. Under isothermal conditions this steel follows a parabolic oxidation kinetics, with an activation energy of 324 kJ mol–1. This value corresponds to an apparent activation energy for the global process, which includes both outward diffusion of Fe cations and inward diffusion of oxygen. The oxidation products present in the oxide scales, which were characterised by X-ray diffraction and SEM, are in total agreement with the Fe-O phase diagram. Thus, magnetite is the most stable oxide at low temperatures and wustite starts to form above 570 °C. Further studies of the effect of cooling rate have shown that wustite formed at 700 °C transforms into magnetite during a slow cooling, whereas a rapid cooling inhibits this transformation to a certain extent. For non-isothermal oxidation treatments consisting of a holding period at 550 °C followed by a single or double 4 hours exposure at 700 °C, the oxidation process seems to occur in sequence, thus presenting an additive effect of the oxidation treatments carried out at each temperature. This effect was observed both, for the type of oxide grown, and for the kinetics of the process. Microscopic observations of the oxide scales formed after the various oxidation treatments revealed that the oxide scales are constituted by sublayers of distinct microstructure and chemical composition changing from their surface to the substrate interface.

    Se han realizado tratamientos de oxidación isotermos y no isotermos a un acero ferrítico 2,25Cr–1Mo en vapor de agua, a temperaturas comprendidas entre 550 y 700 °C y tiempos de hasta 56 días. En condiciones isotermas, este acero tiene una cinética de oxidación parabólica, con una energía de activación de 324 kJ mol–1. Este valor corresponde a una energía de

  5. Influence of the ferritic-pearlitic steel microstructure on surface roughness in broaching of automotive steels

    Science.gov (United States)

    Arrieta, I.; Courbon, C.; Cabanettes, F.; Arrazola, P.-J.; Rech, J.

    2017-10-01

    The aim of this work is to characterize the effect of microstructural parameters on surface roughness in dry broaching with a special emphasis on the ferrite-pearlite (FP) ratio. An experimental approach combining cutting and tribological tests has been developed on three grades 27MnCr5, C45, C60 covering a wide range of FP ratio. Fundamental broaching tests have been performed with a single tooth to analyse the resulting surface quality with uncoated M35 HSS tools. A specially designed open tribometer has been used to characterize the friction coefficient at the tool-chip-workpiece interface under appropriate conditions. Specific phenomena have been observed depending on the FP ratio and an interesting correlation with the tribological tests has been found. This clearly shows that friction has an important contribution in broaching and that phase distribution has to be highly considered when cutting a FP steel at a microscopic scale. This work also provides quantitative data of the friction coefficient depending on the sliding velocity and FP content which can be implemented in any analytical or numerical model of a broaching operation.

  6. Elevated temperature properties of ferritic/martensitic steels for application to future nuclear reators

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Kim, Sung Ho; Ryu, Woo Seog; Chang, Jong Hwa

    2005-12-15

    The future nuclear systems such as nuclear hydrogen production reactors and fusion reactor require low activation and radiation embrittlement resistance in addition to excellent high temperature properties because their operating temperature are even higher than those of the light water reactors. The R and D of ferritic-martensitic steels in nuclear leading centuries like USA, Japan and EU has been continued for decades of years nuclear and they commercialized several steels. Korea consider modified 9Cr-1Mo steel as a candidate materials for reactor pressure vessel of very high temperature reactor. This state-of-the art report aimed to provide informations about the applicabilities of high Cr steels and low Cr steels through the analyses of their microstructures, mechanical properties and radiation characteristics. The metallurgical understanding of background of alloy evolutions might be helpful for the establishment of research orientation.

  7. Mechanosynthesis of A Ferritic ODS (Oxide Dispersion Strengthened) Steel Containing 14% Chromium and Its Characterization

    Science.gov (United States)

    Rivai, A. K.; Dimyati, A.; Adi, W. A.

    2017-05-01

    One of the advanced materials for application at high temperatures which is aggressively developed in the world is ODS (Oxide Dispersion strengthened) steel. ODS ferritic steels are one of the candidate materials for future nuclear reactors in the world (Generation IV reactors) because it is able to be used in the reactor above 600 °C. ODS ferritic steels have also been developed for the interconnect material of SOFC (Solid Oxide Fuel Cell) which will be exposed to about 800 °C of temperature. The steel is strengthened by dispersing homogeneously of oxide particles (ceramic) in nano-meter sized in the matrix of the steel. Synthesis of a ferritic ODS steel by dispersion of nano-particles of yttrium oxide (yttria: Y2O3) as the dispersion particles, and containing high-chromium i.e. 14% has been conducted. Synthesis of the ODS steels was done mechanically (mechanosynthesis) using HEM (High Energy ball Milling) technique for 40 and 100 hours. The resulted samples were characterized using SEM-EDS (Scanning Electron Microscope-Energy Dispersive Spectroscope), and XRD (X-ray diffraction) to analyze the microstructure characteristics. The results showed that the crystal grains of the sample with 100 hours milling time was much smaller than the sample with 40 hours milling time, and some amount of alloy was formed during the milling process even for 40 hours milling time. Furthermore, the structure analysis revealed that some amount of iron atom substituted by a slight amount of chromium atom as a solid solution. The quantitative analysis showed that the phase mostly consisted of FeCr solid-solution with the structure was BCC (body-centered cubic).

  8. Development of ferritic steels for fast induced-radioactivity decay

    International Nuclear Information System (INIS)

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

    1985-01-01

    Tempering studies were conducted on eight heats of normalized chromium-tungsten steel that contained variations in the composition of chromium, tungsten, vanadium, and tantalum. Hardness measurements and optical metallographic observations were used to determine alloying effects on tempering resistance between 650 to 780 0 C. The results were compared to results for analogous chromium-molybdenum steels. 6 references, 6 figures, 1 table

  9. Effect of chromium contents on material properties of high chromium ferritic steel

    International Nuclear Information System (INIS)

    Ando, Masanori; Wakai, Takashi; Aoto, Kazumi

    2003-05-01

    High chromium ferritic steel, having both advanced thermal properties and high temperature strength, is a candidate for the structural material of the future Japanese Fast Breeder Reactor (FBR). In this study, material physical properties of several kinds of 12Cr steel and high purity Fe-Cr alloys are measured to suggest the adequate high chromium steel for the structural material of FBR. The following conclusions are obtained from measured data and the literature data of 2.25Cr-1Mo and Mod.9Cr-1Mo steels. (1) Thermal conductivity decrease with Cr contents increase. However, the difference of the thermal conductivity caused by Cr contents becomes not so significant in high temperature. (2) Thermal expansion decreases with Cr contents increase. However, Cr dope in the iron more than 30mass% is not so efficient to suppress the thermal expansion. (3) Young's modulus increase with Cr contents increase. (4) In this study, the effect of W contents on the 12Cr steels is insignificant. (5) Improving the performance against thermal stress by doping Cr is expected as far as the iron contains low Cr. (6) About results suggest that there is limited utility to improve the physical properties of high Cr ferritic steel for FBR by control of Cr contents. (author)

  10. Simulation and experimental approach to CVD-FBR aluminide coatings on ferritic steels under steam oxidation

    International Nuclear Information System (INIS)

    Leal, J.; Alcala, G.; Bolivar, F.J.; Sanchez, L.; Hierro, M.P.; Perez, F.J.

    2008-01-01

    The ferritic steels used to produce structural components for steam turbines are susceptible to strong corrosion and creep damage due to the extreme working conditions pushed to increase the process efficiency and to reduce pollutants release. The response of aluminide coatings on the P-92 ferritic steel, deposited by CVD-FBR, during oxidation in a simulated steam environment was studied. The analyses were performed at 650 deg. C in order to simulate the working conditions of a steam turbine, and 800 deg. C in order to produce a critical accelerated oxidation test. The Thermo-Calc software was used to predict the different solid phases that could be generated during the oxidation process, in both, coated and uncoated samples. In order to validate the thermodynamic results, the oxides scales produced during steam tests were characterized by different techniques such as XRD, SEM and EDS. The preliminary results obtained are discussed in the present work

  11. Effect of biaxial loading on the fracture behavior of a ferritic steel component

    International Nuclear Information System (INIS)

    Wright, D.J.; Sharples, J.K.; Gardner, L.

    1993-01-01

    The effect of biaxial loading on the ductile tearing behaviour of a through-wall crack in a ferritic steel structure under contained yield is of particular interest to the structural integrity argument for reactor pressure vessels. This results from the fact that there are many instances in practice, (for example a crack in a circumferential weld), where a significant applied stress is present in the direction parallel to the crack as well as in the perpendicular direction. Two large plate ductile tearing tests have been performed on center through-crack specimens (75 mm by 2 m by 2 m) manufactured from a ferritic steel. The first test specimen was loaded in uniaxial tension and the second test specimen was loaded biaxially. This paper presents experimental details and results of the two wide plate tests and describes the analysis work being undertaken which is required to interpret the experiments satisfactorily. Preliminary results of this analysis work are presented. 2 refs., 19 figs

  12. Simulation and experimental approach to CVD-FBR aluminide coatings on ferritic steels under steam oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Leal, J. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain); Alcala, G. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain)], E-mail: galcades@yahoo.es; Bolivar, F.J.; Sanchez, L.; Hierro, M.P.; Perez, F.J. [Universidad Complutense de Madrid, Dep. CC. Materiales e Ingenieria Metalurgica, Avenida Complutense s/n, Facultad de Ciencias Quimicas, 28040 Madrid (Spain)

    2008-07-15

    The ferritic steels used to produce structural components for steam turbines are susceptible to strong corrosion and creep damage due to the extreme working conditions pushed to increase the process efficiency and to reduce pollutants release. The response of aluminide coatings on the P-92 ferritic steel, deposited by CVD-FBR, during oxidation in a simulated steam environment was studied. The analyses were performed at 650 deg. C in order to simulate the working conditions of a steam turbine, and 800 deg. C in order to produce a critical accelerated oxidation test. The Thermo-Calc software was used to predict the different solid phases that could be generated during the oxidation process, in both, coated and uncoated samples. In order to validate the thermodynamic results, the oxides scales produced during steam tests were characterized by different techniques such as XRD, SEM and EDS. The preliminary results obtained are discussed in the present work.

  13. Corrosion studies on Cu-Ni alloys and ferritic steel in salt water for desalination service

    International Nuclear Information System (INIS)

    Shibad, P.R.; Balachandra, J.

    1975-01-01

    Corrosion studies on In 838 and In 848 alloys in 3% NaCl solution, synthetic sea water and in 3% NaCl at pH3 and pH10 indicate that the latter alloy is more corrosion resistant than the former at room (28 0 C), and boiling temperature (101 0 C) and at 125 0 C. Ferritic steel is unaffected in boiling synthetic sea water. In boiling 3% NaCl solution at pH3 and pH10, (the pH values adjusted at room temperature) increase in the rate of corrosion of ferritic steel compared to that at room temperature has been observed. A fair correlation between polarization characteristics and dissolution rates in these solutions is seen for all these materials. (author)

  14. Numerical simulation of hydrogen-assisted crack initiation in austenitic-ferritic duplex steels

    International Nuclear Information System (INIS)

    Mente, Tobias

    2015-01-01

    Duplex stainless steels have been used for a long time in the offshore industry, since they have higher strength than conventional austenitic stainless steels and they exhibit a better ductility as well as an improved corrosion resistance in harsh environments compared to ferritic stainless steels. However, despite these good properties the literature shows some failure cases of duplex stainless steels in which hydrogen plays a crucial role for the cause of the damage. Numerical simulations can give a significant contribution in clarifying the damage mechanisms. Because they help to interpret experimental results as well as help to transfer results from laboratory tests to component tests and vice versa. So far, most numerical simulations of hydrogen-assisted material damage in duplex stainless steels were performed at the macroscopic scale. However, duplex stainless steels consist of approximately equal portions of austenite and δ-ferrite. Both phases have different mechanical properties as well as hydrogen transport properties. Thus, the sensitivity for hydrogen-assisted damage is different in both phases, too. Therefore, the objective of this research was to develop a numerical model of a duplex stainless steel microstructure enabling simulation of hydrogen transport, mechanical stresses and strains as well as crack initiation and propagation in both phases. Additionally, modern X-ray diffraction experiments were used in order to evaluate the influence of hydrogen on the phase specific mechanical properties. For the numerical simulation of the hydrogen transport it was shown, that hydrogen diffusion strongly depends on the alignment of austenite and δ-ferrite in the duplex stainless steel microstructure. Also, it was proven that the hydrogen transport is mainly realized by the ferritic phase and hydrogen is trapped in the austenitic phase. The numerical analysis of phase specific mechanical stresses and strains revealed that if the duplex stainless steel is

  15. Creep transients in a nuclear-grade ODS ferritic steel

    Science.gov (United States)

    Evans, R. W.; Preston, J.; Wilshire, B.; Little, E. A.

    1992-10-01

    Inflexions are detected in the steady state creep regime of a nuclear-grade 13% Cr oxide-dispersion-strengthened ferritic alloy when tested at 700-725°C in an annealed condition. This anomalous response can be suppressed by using a two-stage annealing plus ageing heat treatment designed to fully precipitate a stable chi-phase intermetallic. Mechanisms directly related to observed creep-induced precipitation of chi-phase cannot account for the creep transients, but a tentative explanation based on localized grain boundary migration is in accord with the experimental observations.

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

    International Nuclear Information System (INIS)

    Swindeman, R.W.

    1984-01-01

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

  17. Dislocation structures in cyclically strained X10CrAl24 ferritic steel

    Czech Academy of Sciences Publication Activity Database

    Petrenec, Martin; Polák, Jaroslav; Obrtlík, Karel; Man, Jiří

    2006-01-01

    Roč. 54, č. 13 (2006), s. 3429-3443 ISSN 1359-6454. [Micromechanics and Microstructure Evolution : Modeling Simulation and Experiments. Madrid, 11.09.2005-16.09.2006] R&D Projects: GA ČR(CZ) GP106/05/P521 Institutional research plan: CEZ:AV0Z20410507 Keywords : Transmission electron microscopy * Ferritic steel * Fatigue Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 3.549, year: 2006

  18. Corrosion stability of ferritic stainless steels for solid oxide electrolyser cell interconnects

    DEFF Research Database (Denmark)

    Palcut, Marián; Mikkelsen, Lars; Neufeld, Kai

    2010-01-01

    Long-term oxidation behaviour of eight ferritic steels with 20–29 wt.% chromium (F 20 T, TUS 220 M, AL 453, Crofer 22 APU, Crofer 22 H, Sanergy HT, E-Brite and AL 29-4C) has been studied. The samples were cut into square coupons, ground and annealed for 140–1000 h at 1173 K in flowing, wet hydrog...... of a superior alloy composition are given....

  19. Fretting and wear of stainless and ferritic steels in LMFBR steam generators

    International Nuclear Information System (INIS)

    Lewis, M.W.J.; Campbell, C.S.

    1981-01-01

    Steam generators for LMFBR's may be subject to both fretting wear as a result of flow-induced vibrations and to wear from larger amplitude sliding movements from thermal changes. Results of tests simulating the latter are given for stainless and ferritic steels. For the assessment of fretting wear damage, vibration assessments must be combined with data on specific wear rates. Test mechanisms used to study fretting in sodium covering impact, impact-slide and pure rubbing are described and results presented. (author)

  20. Boron Segregation and Creep in Ultra Fine Grained Tempered Martensite Ferritic Steels

    Czech Academy of Sciences Publication Activity Database

    Eggeler, G.; Dlouhý, Antonín

    2005-01-01

    Roč. 96, č. 7 (2005), s. 743-748 ISSN 0044-3093 R&D Projects: GA ČR(CZ) GA106/99/1172; GA ČR(CZ) GA106/93/0965 Institutional research plan: CEZ:AV0Z20410507 Keywords : Ultra-fine grained materials * Tempered martensite ferritic steels * Boron segregation Subject RIV: JG - Metallurgy Impact factor: 0.842, year: 2005

  1. Surface Modification of Ferritic Stainless Steel by Active Screen Plasma Nitriding

    OpenAIRE

    NII, Hiroaki; NISHIMOTO, Akio

    2012-01-01

    Plasma nitriding is a surface modification process with a low environmental impact. Active screen plasma nitriding (ASPN) is one of the new plasma nitriding technologies, and can eliminate problems related to conventional direct current plasma nitriding (DCPN). In this study, ferritic stainless steel SUS430 samples were treated by ASPN to increase their wear resistance without decreasing their corrosion resistance. ASPN was performed in a nitrogen-hydrogen atmosphere with 25%N2 + 75%H2 for 18...

  2. Effect of annealing temperature and time on microstructure and mechanical properties of high Cr ferritic casting steel

    Science.gov (United States)

    Suo, Z. Y.; Fu, L. M.; Zhang, R. N.; Wang, Y. J.; Shan, A. D.

    2017-09-01

    A new-type of high Cr ferrite cast steel was designed and investigated. Effects of annealing temperature and time on the microstructure and mechanical properties of the high Cr ferrite cast steel were studied. The results show that the microstructures of the as-cast and annealing steels are composed of ferrite and (Cr•Fe)23C6 carbide. The morphology of carbides is from long rod and the continuous network to crystal precipitation for the steels with increasing of annealing temperature and time. The impact toughness is slightly increased from 6 J/cm2 to 8 J/cm2 when the annealing temperature increases from 1180 ℃ to 1200 ℃. But the hardness is about HB 200 and no obvious differences between the as-cast and annealing steels. The most suitable annealing temperature and time are 1200 ℃ and 5 h, respectively. The wear resistance of the high Cr ferrite cast steel is increased and improved with annealing temperature and holding time at 260 ℃. The wear mechanism is changed from abrasion wear to abrasive and adhesive wear. The good wear-resistant of the high Cr ferrite cast steel is mainly attributed to the fine uniformly dispersed carbides.

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

    Science.gov (United States)

    Eghlimi, Abbas; Shamanian, Morteza; Raeissi, Keyvan

    2013-12-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    The present investigation aimed at researching the mechanical properties of the oxide dispersion strengthened (ODS) ferritic steels with different Cr content, which were fabricated through a consolidation of mechanical alloyed (MA) powders of 0.35 wt.% nano Y 2 O 3 dispersed Fe–12.0Cr–0.5Ti–1.0W (alloy A), Fe–16.0Cr–0.5Ti–1.0W (alloy B), and Fe–18.0Cr–0.5Ti–1.0W (alloy C) alloys (all in wt.%) by hot isostatic pressing (HIP) with 100 MPa pressure at 1150 °C for 3 h. The mechanical properties, including the tensile strength, hardness, and impact fracture toughness were tested by universal testers, while Young’s modulus was determined by ultrasonic wave non-destructive tester. It was found that the relationship between Cr content and the strength of ODS ferritic steels was not a proportional relationship. However, too high a Cr content will cause the precipitation of Cr-enriched segregation phase, which is detrimental to the ductility of ODS ferritic steels

  5. Mechanical behaviour of ferritic ODS steels - Temperature dependancy and history

    Czech Academy of Sciences Publication Activity Database

    Fournier, B.; Steckmeyer, A.; Rouffié, A.-L.; Malaplate, J.; Garnier, J.; Ratti, M.; Wident, P.; Ziolek, L.; Tournie, I.; Rabeau, V.; Gentzbittel, J.M.; Kruml, Tomáš; Kuběna, Ivo

    2012-01-01

    Roč. 430, 1-3 (2012), s. 142-149 ISSN 0022-3115 Institutional support: RVO:68081723 Keywords : ODS steels * fatigue * fracture mechanics Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 1.211, year: 2012

  6. 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

  7. Microstructure and mechanical properties of an oxide dispersion strengthened ferritic steel by a new fabrication route

    International Nuclear Information System (INIS)

    Guo Lina; Jia Chengchang; Hu Benfu; Li Huiying

    2010-01-01

    A reduced activation oxide dispersion strengthened (ODS) ferritic steel with nominal composition of Fe-12Cr-2.5W-0.25Ti-0.2V-0.4Y 2 O 3 (designated 12Cr-ODS) was produced by using EDTA-citrate complex method to synthesize and add Y 2 O 3 particles to an argon atomized steel powder, followed by hot isostatic pressing at 1160 deg. C for 3 h under the pressure of 130 MPa, forging at 1150 deg. C, and heat treatment at 1050 deg. C for 2 h. The microstructure, tensile, and Charpy impact properties of the 12Cr-ODS steel were investigated. Transmission electron microscopy studies indicate that the 12Cr-ODS steel exhibits the characteristic ferritic structure containing few dislocations. Tensile characterization has shown that the 12Cr-ODS steel has superior tensile strength accompanied by good elongation at room temperature and 550 deg. C. The material exhibits very attractive Charpy impact properties with upper shelf energy of 22 J and a ductile-to-brittle transition temperature (DBTT) of about -15 deg. C. The formation of small, equiaxed grains and fine dispersion of oxide particles are the main reasons for the good compromise between tensile strength and impact properties.

  8. Heavy-section steel irradiation program summary

    International Nuclear Information System (INIS)

    Corwin, W.R.; Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

    1992-01-01

    Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict failure models of the vessel material. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness (K Ic and J Ic ), crack-arrest toughness (K Ia ), ductile tearing resistance (dJ/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. (orig./GL)

  9. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects; Mecanismes de fragilisation sous irradiation aux neutrons d'alliages modeles ferritiques et d'un acier de cuve: amas de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

    The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)

  10. Tensile properties of shielded metal arc welded dissimilar joints of nuclear grade ferritic steel and austenitic stainless steel

    Science.gov (United States)

    Karthick, K.; Malarvizhi, S.; Balasubramanian, V.; Krishnan, S. A.; Sasikala, G.; Albert, Shaju K.

    2016-12-01

    In nuclear power plants, modified 9Cr-1Mo ferritic steel (Grade 91 or P91) is used for constructing steam generators (SG's) whereas austenitic stainless steel (AISI 316LN) is a major structural member for intermediate heat exchanger (IHX). Therefore, a dissimilar joint between these materials is unavoidable. In this investigation, dissimilar joints were fabricated by Shielded Metal Arc Welding (SMAW) process with Inconel 82/182 filler metals. Transverse tensile properties and Charpy V-notch impact toughness for different regions of dissimilar joints of modified 9Cr-1Mo ferritic steel and AISI 316LN austenitic stainless steel were evaluated as per the standards. Microhardness distribution across the dissimilar joint was recorded. Microstructural features of different regions were characterized by optical and scanning electron microscopy. The transverse tensile properties of the joint is found to be inferior to base metals. Impact toughness values of different regions of dissimilar metal weld joint (DMWJ) is slightly higher than the prescribed value. Formation of a soft zone at the outer edge of the HAZ will reduce the tensile properties of DMWJ. The complex microstructure developed at the interfaces of DMWJ will reduce the impact toughness values.

  11. Plastic deformation of ferritic grains in presence of ODS particles and irradiation-induced defect clusters: A 3D dislocation dynamics simulation study

    Science.gov (United States)

    Robertson, C.; Gururaj, K.

    2011-08-01

    Ferritic steels strengthened by oxide particle dispersions (ODS) are prime candidate for future nuclear applications. So far, the beneficial mechanical characteristics of ODS steels are not fully understood, in terms of dislocation-based mechanisms. In this work, three-dimensional discrete dislocation dynamics simulations were carried out to analyze pre and post-irradiation plastic deformation in ferritic grains, with and without ODS particles. In the absence of irradiation induced defect loops, ODS-grains are stronger and plastic strain is more localized than in the corresponding, particle-free grain. After irradiation however, ODS-grain become more resistant to loop-induced hardening, while plastic strain spreading is broader, with respect to particle-free grain. This effect is due to dislocations accumulating next to the precipitates, generating internal stress allowing cross-slipped dislocations to go past the irradiation induced defect loops. Cross-slip is therefore a key feature of our model, for explaining the beneficial role of ODS particles on post-irradiation plastic deformation.

  12. A Dislocation-Based Theory for the Deformation Hardening Behavior of DP Steels: Impact of Martensite Content and Ferrite Grain Size

    Directory of Open Access Journals (Sweden)

    Yngve Bergström

    2010-01-01

    Full Text Available A dislocation model, accurately describing the uniaxial plastic stress-strain behavior of dual phase (DP steels, is proposed and the impact of martensite content and ferrite grain size in four commercially produced DP steels is analyzed. It is assumed that the plastic deformation process is localized to the ferrite. This is taken into account by introducing a nonhomogeneity parameter, f(ε, that specifies the volume fraction of ferrite taking active part in the plastic deformation process. It is found that the larger the martensite content the smaller the initial volume fraction of active ferrite which yields a higher initial deformation hardening rate. This explains the high energy absorbing capacity of DP steels with high volume fractions of martensite. Further, the effect of ferrite grain size strengthening in DP steels is important. The flow stress grain size sensitivity for DP steels is observed to be 7 times larger than that for single phase ferrite.

  13. Evaluation of toughness degradation by small punch (SP) tests for neutron irradiated structural steels

    International Nuclear Information System (INIS)

    Misawa, Toshihei; Hamaguchi, Yoshikazu; Kimura, Akihiko; Eto, Motokuni; Suzuki, Masahide; Nakajima, Nobuya.

    1992-01-01

    The small punch (SP) test as one of the useful small specimen testing technique (SSTT) has been developed to evaluate the fracture toughness, ductile-brittle transition temperature (DBTT) and tensile properties for neutron irradiated structural materials. The SP tests using the miniaturized specimens of φ3 mm TEM disk and 10 mm 2 coupon were performed for six kinds of ferritic steels of F-82, F-82H, HT-9, JFMS, 2.25-1Mo and SQV2A. It was shown that the temperature dependence of SP fracture energies with scatter in miniaturized testing can give reliable information on the DBTT by use of the statistical analysis based on the Weibull distribution. A good correlation between the DBTT of the SP tests and that of the standard CVN test has been obtained for the various nuclear ferritic steels. The SP test was performed for cryogenic austenitic steels as a way of evaluating elastic-plastic fracture toughness, J IC , on the basis of a universal empirical relationship between J IC and SP equivalent fracture strain, ε-bar qf . The SP testing using the neutron irradiated specimens of 2.25Cr-1Mo, F-82, F-82H and HT-9 steels was successfully applied and presented the neutron radiation induced changes on the DBTT, fracture toughness and tensile properties. (author)

  14. Heavy-Section Steel Irradiation Program

    International Nuclear Information System (INIS)

    Rosseel, T.M.

    2000-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established

  15. Analysis of the grain size evolution for ferrite formation in Fe-C-Mn steels using a 3D model under a mixed-mode interface condition

    NARCIS (Netherlands)

    Fang, H.; Mecozzi, M.G.; Brück, E.H.; van der Zwaag, S.; van Dijk, N.H.

    2018-01-01

    A 3D model has been developed to predict the average ferrite grain size and grain size distribution for an austenite-to-ferrite phase transformation during continuous cooling of an Fe-C-Mn steel. Using a Voronoi construction to represent the austenite grains, the ferrite is assumed to nucleate at

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

    International Nuclear Information System (INIS)

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

    2017-01-01

    The nanoparticles in an Al-alloyed high-Cr oxide dispersion strengthened (ODS) ferritic steel with Hf addition, i.e., SOC-16 (Fe-15Cr-2W-0.1Ti-4Al-0.62Hf-0.35Y 2 O 3 ), have been examined by transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM). Relative to an Al-alloyed high-Cr ODS ferritic steel without Hf addition, i.e., SOC-9 (Fe-15.5Cr-2W-0.1Ti-4Al-0.35Y 2 O 3 ), the dispersion morphology and coherency of the oxide nanoparticles in SOC-16 were significantly improved. Almost all the small nanoparticles (diameter <10 nm) in SOC-16 were found to be consistent with cubic Y 2 Hf 2 O 7 oxides with the anion-deficient fluorite structure and coherent with the bcc steel matrix. The larger particles (diameter >10 nm) were also mainly identified as cubic Y 2 Hf 2 O 7 oxides with the anion-deficient fluorite structure. The results presented here are compared with those of SOC-9 with a brief discussion of the underlying mechanisms of the unusual thermal and irradiation stabilities of the oxides as well as the superior strength, excellent irradiation tolerance and extraordinary corrosion resistance of SOC-16.

  17. Preparation, characterization and application of nanosized copper ferrite photocatalysts for dye degradation under UV irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Zaharieva, Katerina, E-mail: zaharieva@ic.bas.bg [Institute of Catalysis, Bulgarian Academy of Sciences, Acad. G. Bonchev St., Block 11, 1113 Sofia (Bulgaria); Rives, Vicente, E-mail: vrives@usal.es [GIR-QUESCAT, Dpto. Química Inorgánica, Universidad de Salamanca, 37008 Salamanca (Spain); Tsvetkov, Martin, E-mail: mptsvetkov@gmail.com [Faculty of Chemistry and Pharmacy, St. Kliment Ohridski University of Sofia, 1 J. Bourchier Blvd., 1164 Sofia (Bulgaria); Cherkezova-Zheleva, Zara, E-mail: zzhel@ic.bas.bg [Institute of Catalysis, Bulgarian Academy of Sciences, Acad. G. Bonchev St., Block 11, 1113 Sofia (Bulgaria); Kunev, Boris, E-mail: bkunev@ic.bas.bg [Institute of Catalysis, Bulgarian Academy of Sciences, Acad. G. Bonchev St., Block 11, 1113 Sofia (Bulgaria); Trujillano, Raquel, E-mail: rakel@usal.es [GIR-QUESCAT, Dpto. Química Inorgánica, Universidad de Salamanca, 37008 Salamanca (Spain); Mitov, Ivan, E-mail: mitov@ic.bas.bg [Institute of Catalysis, Bulgarian Academy of Sciences, Acad. G. Bonchev St., Block 11, 1113 Sofia (Bulgaria); Milanova, Maria, E-mail: nhmm@wmail.chem.uni-sofia.bg [Faculty of Chemistry and Pharmacy, St. Kliment Ohridski University of Sofia, 1 J. Bourchier Blvd., 1164 Sofia (Bulgaria)

    2015-06-15

    {sup −3} min{sup −1}) for degradation of organic dye Malachite green under UV irradiation. - Highlights: • Copper ferrites via co-precipitation, mechanochemical and/or thermal treatment. • Nano ferrites show a superparamagnetic and collective magnetic excitations nature. • The co-precipitated Cu{sub 0.25}Fe{sub 2.75}O{sub 4} posses the highest photocatalytic activity. • The amount adsorbed Malachite Green by catalyst depends on the preparation method. • The prepared copper ferrites can be applicable as cheap adsorbents and catalysts.

  18. Preparation, characterization and application of nanosized copper ferrite photocatalysts for dye degradation under UV irradiation

    International Nuclear Information System (INIS)

    Zaharieva, Katerina; Rives, Vicente; Tsvetkov, Martin; Cherkezova-Zheleva, Zara; Kunev, Boris; Trujillano, Raquel; Mitov, Ivan; Milanova, Maria

    2015-01-01

    Malachite green under UV irradiation. - Highlights: • Copper ferrites via co-precipitation, mechanochemical and/or thermal treatment. • Nano ferrites show a superparamagnetic and collective magnetic excitations nature. • The co-precipitated Cu 0.25 Fe 2.75 O 4 posses the highest photocatalytic activity. • The amount adsorbed Malachite Green by catalyst depends on the preparation method. • The prepared copper ferrites can be applicable as cheap adsorbents and catalysts

  19. Study of corrosion resistance of AISI 444 ferritic stainless steel for application as a biomaterial

    International Nuclear Information System (INIS)

    Marques, Rogerio Albuquerque

    2014-01-01

    Ferritic stainless steels are ferromagnetic materials. This property does not allow their use in orthopedic prosthesis. Nevertheless, in some specific applications, this characteristic is very useful, such as, for fixing dental and facial prostheses by using magnetic attachments. In this study, the corrosion resistance and cytotoxicity of the AISI 444 ferritic stainless steel, with low nickel content, extra-low interstitial levels (C and N) and Ti and Nb stabilizers, were investigated for magnetic dental attachments application. The ISO 5832-1 (ASTM F-139) austenitic stainless steel and a commercial universal keeper for dental attachment (Neo-magnet System) were evaluated for comparison reasons. The first stainless steel is the most used metallic material for prostheses, and the second one, is a ferromagnetic keeper for dental prostheses (NeoM). In vitro cytotoxicity analysis was performed by the red neutral incorporation method. The results showed that the AISI 444 stainless steel is non cytotoxic. The corrosion resistance was studied by anodic polarization methods and electrochemical impedance spectroscopy (EIS), in a saline phosphate buffered solution (PBS) at 37 °C. The electronic properties of the passive film formed on AISI 444 SS were evaluated by the Mott-Schottky approach. All tested materials showed passivity in the PBS medium and the passive oxide film presented a duplex nature. The highest susceptibility to pitting corrosion was associated to the NeoM SS. This steel was also associated to the highest dopant concentration. The comparatively low levels of chromium (nearly 12.5%) and molybdenum (0.3%) of NeoM relatively to the other studied stainless steels are the probable cause of its lower corrosion resistance. The NeoM chemical composition does not match that of the SUS444 standards. The AISI 444 SS pitting resistance was equivalent to the ISO 5832-1 pointing out that it is a potential candidate for replacement of commercial ferromagnetic alloys used

  20. Identification of Age, Temperature and Radiation Effect on Ferritic Steel Microstructure Based on Artificial Intelligence

    International Nuclear Information System (INIS)

    Mike Susmikanti; Entin Hartini; Antonius Sitompul

    2008-01-01

    In the construction of nuclear installation, it is important to know the material condition used on it. Considering mechanical properties of these materials, there are some material change affected by ageing, temperature and radiation. For some years, austenitic stainless steel are used as a fuel cladding in fast breeder reactor. However this material will not sufficiently competitive from economic point of view for the next year. Experiment result on ferritic steel gave information of stronger structural properties compared to austenitic stainless steel. Modeling and simulation will support further identification of this material changing caused by such effects. Pattern recognition of these changes base on artificial intelligence is expected to support the research and development activities on nuclear structure materials. Material structure pattern of these materials, observed by SEM, are converted using image processing system. Its characteristic is then analyzed with principal component using perception method, which usually used on identifying and learning neural network system based on artificial intelligence. Specific design and input are needed to identify the change of material structure pattern before and after any applied effect. In this paper, simulation of changing identification on three types ferritic steel F17(17 Cr), EM 12 (9 CR-2 MoNbV), and EMI 0 (9 Cr-I Mo) were done. The microstructure data before and after effect are taken from some references. The whole pattern recognition process are done using MATLAB software package. (author)

  1. Identification, size classification and evolution of Laves phase precipitates in high chromium, fully ferritic steels.

    Science.gov (United States)

    Lopez Barrilao, Jennifer; Kuhn, Bernd; Wessel, Egbert

    2017-10-01

    To fulfil the new challenges of the German "Energiewende" more efficient, sustainable, flexible and cost-effective energy technologies are strongly needed. For a reduction of consumed primary resources higher efficiency steam cycles with increased operating parameters, pressure and temperature, are mandatory. Therefore advanced materials are needed. The present study focuses on a new concept of high chromium, fully ferritic steels. These steels, originally designed for solid oxide fuel cell applications, provide favourable steam oxidation resistance, creep and thermomechanical fatigue behaviour in comparison to conventional ferritic-martensitic steels. The strength of this type of steel is achieved by a combination of solid-solution hardening and precipitation strengthening by intermetallic Laves phase particles. The effect of alloy composition on particle composition was measured by energy dispersive X-ray spectroscopy and partly verified by thermodynamic modelling results. Generally the Laves phase particles demonstrated high thermodynamic stability during long-term annealing up to 40,000h at 600°C. Variations in chemical alloy composition influence Laves phase particle formation and consequently lead to significant changes in creep behaviour. For this reason particle size distribution evolution was analysed in detail and associated with the creep performance of several trial alloys. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Li, M.; Soppet, W.K.; Rink, D.L.; Listwan, J.T.; Natesan, K. (Nuclear Engineering Division)

    2012-05-10

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

  3. Characterization and Strain-Hardening Behavior of Friction Stir-Welded Ferritic Stainless Steel

    Science.gov (United States)

    Sharma, Gaurav; Dwivedi, Dheerendra Kumar; Jain, Pramod Kumar

    2017-12-01

    In this study, friction stir-welded joint of 3-mm-thick plates of 409 ferritic stainless steel (FSS) was characterized in light of microstructure, x-ray diffraction analysis, hardness, tensile strength, ductility, corrosion and work hardening properties. The FSW joint made of ferritic stainless steel comprises of three distinct regions including the base metal. In stir zone highly refined ferrite grains with martensite and some carbide precipitates at the grain boundaries were observed. X-ray diffraction analysis also revealed precipitation of Cr23C6 and martensite formation in heat-affected zone and stir zone. In tensile testing of the transverse weld samples, the failure eventuated within the gauge length of the specimen from the base metal region having tensile properties overmatched to the as-received base metal. The tensile strength and elongation of the longitudinal (all weld) sample were found to be 1014 MPa and 9.47%, respectively. However, in potentiodynamic polarization test, the corrosion current density of the stir zone was highest among all the three zones. The strain-hardening exponent for base metal, transverse and longitudinal (all weld) weld samples was calculated using various equations. Both the transverse and longitudinal weld samples exhibited higher strain-hardening exponents as compared to the as-received base metal. In Kocks-Mecking plots for the base metal and weld samples at least two stages of strain hardening were observed.

  4. Mitigation of sensitisation effects in unstabilised 12%Cr ferritic stainless steel welds

    International Nuclear Information System (INIS)

    Warmelo, Martin van; Nolan, David; Norrish, John

    2007-01-01

    Sensitisation in the heat-affected zones of ferritic stainless steel welds is typically prevented by stabilisation of the parent material with titanium or niobium, and suitable design of the overall composition to produce a suitably high ferrite factor. However, such alloy modification has proven to be economically unviable for thick gauge (>10 mm) plate products and therefore unstabilised 12%Cr (3CR12) material is still currently being used for heavy gauge structural applications in many parts of the world. The aim of the current work was to review the mechanisms responsible for sensitisation in these unstabilised ferritic stainless steels, and to characterise the sensitisation effects arising from multipass welding procedures. The objective was to determine the influence of welding parameters, and thereby to recommend mitigating strategies. Two particular sensitisation modes were found to occur in the current work, although only one was predominant and considered problematic from a practical perspective. It was found that with proper positioning of weld capping runs and control of weld overlap, it is possible to ensure that sensitising isotherms remain buried beneath the parent surface, and so reduce harmful corrosion effects

  5. Stress corrosion cracking behavior of weldments of ferritic stainless steels in high temperature pure water

    International Nuclear Information System (INIS)

    Fujiwara, Kazuo; Tomari, Haruo; Shimogori, Kazutoshi

    1985-01-01

    Considering the application of a ferritic stainless steel as heat exchanger tubing for a moisture separator reheater of light water reactors, stress corrosion cracking behavior at the weldment of commercial ferritic stainless steels in high temperature pure water was studied. Double U-bend method was used for the study and the relationship with microstructure was discussed. Welded joint of Type 439SS containing 0.021% C, 0.025% N and 0.27% Ti with In-82 type filler metal was susceptible to intergranular stress corrosion cracking if a tight crevice was provided by inserting a teflon sheet between the inner and outer specimens of double U-bend. This was attributable to the formation of chromium depleted zone due to the precipitation of chromium carbides/nitrides along ferrite grain boundaries. On the other hand welded joint of Type 444SS with 0.007% C, 0.010% N and 0.26% Nb was immune to stress corrosion cracking, and this might be attributed to the higher ratio of Nb/(C+N) content. (author)

  6. Friction Characteristics of Nitrided Layers on AISI 430 Ferritic Stainless Steel Obtained by Various Nitriding Processes

    Directory of Open Access Journals (Sweden)

    Hakan AYDIN

    2013-03-01

    Full Text Available The influence of plasma, gas and salt-bath nitriding techniques on the friction coefficient of AISI 430 ferritic stainless steel was studied in this paper. Samples were plasma nitrided in 80 % N2 + 20 % H2 atmosphere at 450 °C and 520 °C for 8 h at a pressure of 2 mbar, gas nitrided in NH3 and CO2 atmosphere at 570 °C for 13 h and salt-bath nitrided in a cyanide-cyanate salt-bath at 570 °C for 1.5 h. Characterisation of nitrided layers on the ferritic stainless steel was carried out by means of microstructure, microhardness, surface roughness and friction coefficient measurements. Friction characteristics of the nitrided layers on the 430 steel were investigated using a ball-on-disc friction-wear tester with a WC-Co ball as the counter-body under dry sliding conditions. Analysis of wear tracks was carried out by scanning electron microscopy. Maximum hardness and maximum case depth were achieved on the plasma nitrided sample at 520 ºC for 8 h. The plasma and salt-bath nitriding techniques significantly decreased the average surface roughness of the 430 ferritic stainless steel. The friction test results showed that the salt-bath nitrided layer had better friction-reducing ability than the other nitrided layers under dry sliding conditions. Furthermore, the friction characteristic of the plasma nitrided layer at 520 ºC was better than that of the plasma nitrided layer at 450 °C.DOI: http://dx.doi.org/10.5755/j01.ms.19.1.3819

  7. Neutron diffraction investigation of lattice microstrain in ferrite steel

    International Nuclear Information System (INIS)

    Camanzi, A.; Moze, O.

    1992-01-01

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

  8. Microstructure of 9 Cr-1 MoVNb steel irradiated to 40 dpa at elevated temperatures in HFIR

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-01-01

    As part of an effort by the Office of Fusion Energy to evaluate the irradiation behavior of ferritic steels, a 9 Cr-1 MoVNb alloy was irradiated in HFIR to a dose of approx. 36 dpa at temperatures of 300, 400, 500, and 600/sup 0/C. In addition to the displacement damage produced during irradiation, a transmutation reaction of nickel during HFIR irradiation resulted in the simultaneous production of approx. 30 at. ppM He. Electron microscopy disks in the normalized and tempered condition were irradiated, and the microstructures were evaluated as a function of irradiation temperature. A few small cavities were observed after irradiation at 300, 500, and 600/sup 0/C. However, a pronounced cavity microstructure was found after irradiation at 400/sup 0/C. At this temperature, the cavities had a volume-averaged diameter of 15 nm and a concentration of 1.1 x 10/sup 21/ m/sup -3/, resulting in a void-swelling contribution of 0.19%. The cavities at 400/sup 0/C were homogeneously distributed throughout the tempered martensite matrix, and showed no preference for lath boundaries or precipitate interfaces. The results are compared to those recently reported on a similarly irradiated 12 Cr-1 MoVW ferritic steel.

  9. 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.

  10. Changes of microstructures and high temperature properties during high temperature service of Niobium added ferritic stainless steels

    International Nuclear Information System (INIS)

    Fujita, Nobuhiro; Ohmura, Keiichi; Yamamoto, Akio

    2003-01-01

    To improve the fuel economy and clean the exhaust gas of automobiles, the temperature of exhaust gas is getting higher and higher. Niobium added ferritic stainless steels are often being used in automotive exhaust systems, because of their excellent heat resistant properties, especially thermal fatigue resistance, which is very important for materials of exhaust manifold. However, coarse precipitates containing niobium, which cause degradation in high temperature strength and thermal fatigue resistance, are unavoidable during high temperature service. In this study, changes of microstructures and high temperature properties in high temperature aging were investigated using several Nb added ferritic stainless steels. It has been found that the microstructure stability of Nb-Ti-Mo alloyed steels in high temperature aging is superior to that of Nb added steels. The microstructure stability leads to less degradation in high temperature strength during high temperature aging and to longer thermal fatigue lives of Nb-Ti-Mo alloyed steels than in Nb added steels

  11. Localized corrosion and stress corrosion cracking behavior of austenitic stainless steel weldments containing retained ferrite. Annual progress report, June 1, 1978--March 31, 1979

    International Nuclear Information System (INIS)

    Savage, W.F.; Duquette, D.J.

    1979-03-01

    Localized corrosion and stress corrosion cracking experiments have been performed on single phase 304 stainless steel alloys and autogeneous weldments containing retained delta ferrite as a second phase. The results of the pitting experiments show that the pressure of delta ferrite decreases localized corrosion resistance with pits initiating preferentially at delta ferrite--gamma austenite interphase boundaries. This increased susceptibility is reversible with elevated temperature heat treatments which revert the metastable ferrite phase to the equilibrium austenite phase

  12. Effect of niobium clustering and precipitation on strength of an NbTi-microalloyed ferritic steel

    International Nuclear Information System (INIS)

    Kostryzhev, A.G.; Al Shahrani, A.; Zhu, C.; Cairney, J.M.; Ringer, S.P.; Killmore, C.R.; Pereloma, E.V.

    2014-01-01

    The microstructure–property relationship of an NbTi-microalloyed ferritic steel was studied as a function of thermo-mechanical schedule using a Gleeble 3500 simulator, optical and scanning electron microscope, and atom probe tomography. Contributions to the yield stress from grain size, solid solution, work hardening, particle and cluster strengthening were calculated using the established equations and the measured microstructural parameters. With a decrease in the austenite deformation temperature the yield stress decreased, following a decrease in the number density of >20 nm Nb-rich particles and ≈5 nm Nb-C clusters, although the grain refinement contribution increased. To achieve the maximum cluster/precipitation strengthening in ferrite, the austenite deformation should be carried out in the recrystallisation temperature region where there is a limited tendency for strain-induced precipitation. Based on the analysis of cluster strengthening increment, it could be suggested that the mechanism of dislocation–cluster interaction is closer to shearing than looping

  13. 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)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hsiung, L; Fluss, M; Kimura, A

    2010-04-06

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

  16. Modelling of creep damage development in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    The physical creep damage, which is observed in fossil-fired power plants, is mainly due to the formation of cavities and their interaction. It has previously been demonstrated that both the nucleation and growth of creep cavities can be described by power functions in strain for low alloy and 12 % CrMoV creep resistant steels. It possible to show that the physical creep damage is proportional to the product of the number of cavities and their area. Hence, the physical creep damage can also be expressed in terms of the creep strain. In the presentation this physical creep damage is connected to the empirical creep damage classes (1-5). A creep strain-time function, which is known to be applicable to low alloy and 12 % CrMoV creep resistant steels, is used to describe tertiary creep. With this creep strain - time model the residual lifetime can be predicted from the observed damage. For a given damage class the remaining life is directly proportional to the service time. An expression for the time to the next inspection is proposed. This expression is a function of fraction of the total allowed damage, which is consumed till the next inspection. (orig.) 10 refs.

  17. Increasing the formability of ferritic stainless steel tube by granular medium-based hot forming

    Science.gov (United States)

    Chen, H.; Staupendahl, D.; Hiegemann, L.; Tekkaya, A. E.

    2017-09-01

    Ferritic stainless steel without the alloy constituent nickel is an economical substitution for austenitic stainless steel in the automotive industry. Its lower formability, however, oftentimes prevents the direct material substitution in forming processes such as hydroforming, necessitating new forming strategies. To extend the forming capacity of ferritic stainless steel tube, the approach of forming at elevated temperatures is proposed. Utilizing granular material as forming medium, high forming temperatures up to 900°C are realized. The forming process works by moving punches axially into the granular medium, thereby, compressing it and causing axial as well as radial pressure. In experimental and numerical investigations it is shown that interfacial friction between the granular medium and the tube inherently causes tube feed, resulting in stain states in the tension-compression region of the FLD. Formability data for this region are gained by notched tensile tests, which are performed at room temperature as well as at elevated temperatures. The measured data show that the formability is improved at forming temperatures higher than 700°C. This observed formability increase is experimentally validated using a demonstrator geometry, which reaches expansion ratios that show fracture in specimens formed at room temperature.

  18. Austenitic-ferritic stainless steels: A state-of-the-art review

    Science.gov (United States)

    Voronenko, B. I.

    1997-10-01

    Austenitic-ferritic stainless steels, more commonly known as duplex stainless steels, or DSS for short, consist of two basic phases. One is austenite, A, and the other is ferrite, F, present in about equal amounts (but not less than 30% each). The two phases owe their corrosion resistance to the high chromium content. Compared to austenitic stainless steels, ASS, they are stronger (without sacrificing ductility), resist corrosion better, and cost less due to their relatively low nickel content. DSS can be used in an environment where standard ASS are not durable enough, such as chloride solutions (ships, petrochemical plant, etc.). Due to their low nickel content and the presence of nickel, DSS have good weldability. However, they have a limited service temperature range (from -40 to 300°) because heating may cause them to give up objectionable excess phases and lower the threshold of cold brittleness in the heat-affected zone of welded joints. State-of-the art DSS are alloyed with nitrogen to stabilize their austenite, and in this respect the nitrogen does the job of nickel. Also, nitrogen enhances the strength and resistance to pitting and improves the structure of welds.

  19. Segregation and morphology on the surface of ferritic stainless steel (0 0 1)

    International Nuclear Information System (INIS)

    Fujiyoshi, H.; Matsui, T.; Yuhara, J.

    2012-01-01

    The temperature dependence of the segregation and morphology of ferritic steel (0 0 1) surfaces has been examined by a combination of Auger electron spectroscopy, low-energy electron diffraction (LEED), scanning tunneling microscopy (STM), and low-energy electron loss spectroscopy. Upon annealing ferritic steel at 500 °C, the topmost layer was observed to be mainly composed of Fe-Cr alloy. Oxygen segregation was also detected locally in the STM images. LEED showed a (1 × 1) pattern and a weak (√5 × √5)R27° reconstruction corresponding to Fe and Cr 4 O 5 , respectively. Upon annealing at 600 °C, carbon and chromium co-segregated to the surface, forming two different regions composed of CrC and Cr-based steel, while the Cr 4 O 5 domains disappeared. Upon annealing at 700 °C, nitrogen segregated to the surface, and the topmost layer was observed to be mainly composed of CrN domains with local CrC domains.

  20. Microstructure examination of Fe-14Cr ODS ferritic steels produced through different processing routes

    Science.gov (United States)

    Oksiuta, Z.; Hosemann, P.; Vogel, S. C.; Baluc, N.

    2014-08-01

    Various thermo-mechanical treatments were applied to refine and homogenise grain size and improve mechanical properties of hot-isostatically pressed (HIP) 14%Cr ODS ferritic steel. The grain size was reduced, improving mechanical properties, tensile strength and Charpy impact, however bimodal-like distribution was also observed. As a result, larger, frequently elongated grains with size above 1 μm and refined, equiaxed grains with a diameter ranging from 250 to 500 nm. Neutron diffraction measurements revealed that for HIP followed by hydrostatic extrusion material the strongest fiber texture was observed oriented parallel to the extrusion direction. In comparison with hot rolling and hot pressing methods, this material exhibited promising mechanical properties: the ultimate tensile strength of 1350 MPa, yield strength of 1280 MPa, total elongation of 21.7% and Charpy impact energy of 5.8 J. Inferior Charpy impact energy of ∼3.0 J was measured for HIP and hot rolled material, emphasising that parameters of this manufacturing process still have to be optimised. As an alternative manufacturing route, due to the uniform microstructure and simplicity of the process, hot pressing might be a promising method for production of smaller parts of ODS ferritic steels. Besides, the ductile-to-brittle transition temperature of all thermo-mechanically treated materials, in comparison with as-HIPped ODS steel, was improved by more than 50%, the transition temperature ranging from 50 to 70 °C (323 and 343 K) remains still unsatisfactory.

  1. Effect of Mechanical Alloying Atmospheres and Oxygen Concentration on Mechanical Properties of ODS Ferritic Steels

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Choi, Byoungkwon; Han, Changhee; Kim, Kibaik; Kang, Sukhoon; Chun, Youngbum; Kim, Taekyu

    2013-01-01

    Finely dispersed nano-oxide particles with a high number density in the homogeneous grain matrix are essential to achieve superior mechanical properties at high temperatures, and these unique microstructures can be obtained through the mechanical alloying (MA) and hot consolidation process. The microstructure and mechanical property of ODS steel significantly depends on its powder property and the purity after the MA process. These contents should be carefully controlled to improve the mechanical property at elevated temperature. In particular, appropriate the control of oxygen concentration improves the mechanical property of ODS steel at high temperature. An effective method is to control the mechanical alloying atmosphere by high purity inert gas. In the present study, the effects of mechanical alloying atmospheres and oxygen concentration on the mechanical property of ODS steel were investigated. ODS ferritic alloys were fabricated in various atmospheres, and the HIP process was used to investigate the effects of MA atmospheres and oxygen concentration on the microstructure and mechanical property. ODS ferritic alloys milled in an Ar-H 2 mixture, and He is effective to reduce the excess oxygen concentration. The YH 2 addition made an extremely reduced oxygen concentration by the internal oxygen reduction reaction and resulted in a homogeneous microstructure and superior creep strength

  2. Behaviour and damage of aged austenitic-ferritic steels: a micro-mechanical approach

    International Nuclear Information System (INIS)

    Bugat, St.

    2000-12-01

    The austenitic-ferritic steels are used in the PWR primary cooling system. At the running temperature (320 C), they are submitted to a slow aging, which leads to the embrittlement of the ferritic phase. This embrittlement leads to a decrease of the mechanical properties, in particular of the crack resistance of the austenitic-ferritic steels. The damage and rupture of the austenitic-ferritic steels have been approached at the ENSMP by the works of P. Joly (1992) and of L. Devilliers-Guerville (1998). These works have allowed to reveal a damage heterogeneity which induces a strong dispersion on the ductilities and the toughnesses as well as on the scale effects. Modeling including the damage growth kinetics measured experimentally, have allowed to verify these effects. Nevertheless, they do not consider the two-phase character of the material and do not include a physical model of the cleavage cracks growth which appear in the embrittled ferrite. In this study, is proposed a description of the material allowing to treat these aspects while authorizing the structure calculation. In a first part, the material is studied. The use of the ESBD allows to specify the complex morphology of these steels and crystal orientation relations between the two phases. Moreover, it is shown that the two phases keep the same crystal orientation in the zones, called bicrystals, whose size varies between 500 μm and 1 mm. The study of the sliding lines, coupled to the ESBD, allows to specify too the deformation modes of the two phases. At last, tensile and tensile-compression tests at various deformation range are carried out to characterize the macroscopic mechanical behaviour of these materials. Then, a micro-mechanical modeling of the material behaviour is proposed. This one takes into account the three scales identified at the preceding chapter. The first scale, corresponding to the laths is described as a monocrystal whose behaviour includes both an isotropic and a kinematic strain

  3. Powder processing for the fabrication of oxide dispersion strengthened ferritic steels

    International Nuclear Information System (INIS)

    Coheur, L.; De Wilde, L.

    1975-01-01

    The addition of Ti to the powder mixtures, which will finally give oxide dispersion strengthened ferritic steels, causes the appearance of large oxide particles in the end product. These particles are detrimental to the tensile properties at 700 deg C. The method for preparing powder mixtures is recalled. The effect peculiar to each mill has been shown with binary powder mixtures Fe-TiO 2 . In this case, the relationship between the end product properties and the mill type has been defined more accurately [fr

  4. Optimization of Ferritic Steel Porous Supports for Protonic Fuel Cells Working at 600°C

    DEFF Research Database (Denmark)

    Venkatachalam, Vinothini; Molin, Sebastian; Chen, Ming

    2014-01-01

    , and is particularly helpful for a porous metal supported cell because it limits the corrosion of the metal by exposure to water vapor in the anode gas. In this work, we show the effect of composition and microstructure on the high temperature corrosion and phase stability (formation of sigma phase/Laves phase......) of porous alloys. Alloys in the compositional range Fe-20%Cr to Fe-32%Cr were evaluated and the effects of surface modification on corrosion resistance were studied using thermogravimetry, x-ray diffractometry and electron microscopy. The results show that surface modified porous ferritic steels are very...

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

    DEFF Research Database (Denmark)

    Zhang, Zhenbo; Mishin, Oleg; Pantleon, Wolfgang

    2012-01-01

    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......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...

  6. Recrystallization of niobium stabilized ferritic stainless steel during hot rolling simulation by torsion tests

    Directory of Open Access Journals (Sweden)

    Flávia Vieira Braga

    2016-01-01

    Full Text Available The aim of this study was to investigate the effect of finishing hot rolling temperature in promoting interpass recrystallization on a Nb-stabilized AISI 430 ferritic stainless steel. Torsion tests were performed in order to simulate the Steckel mill rolling process by varying the temperature ranges of the finishing passes. Interrupted torsion test were also performed and interpass recrystallization was evaluated via optical microscopy and electron backscatter diffraction (EBSD. As a result of this work, it has been established, within the restrictions of a Steckel mill rolling schedule, which thermomechanical conditions mostly favor SRX.

  7. Contribution to the structural study of austeno-ferritic steels. Morphological and analytical definition of the ferritic phase

    International Nuclear Information System (INIS)

    Bathily, Alassane.

    1977-07-01

    Conditions of fast and selective austenite dissolution were defined by means of current-voltage curves using AISI 316-type materials (welding beads). The ferritic phase was isolated and identified with X-rays. The percentages of ferrite were compared gravimetrically with those obtained by traditional methods. The ferrite isolated was chemically analysed by atomic absorption, the only doubtful value being carbon. It is shown by this method that a morphological study of the solidification of the ferritic lattice is possible, even for percentages around 1% [fr

  8. Effects of delta ferrite content on the mechanical properties of E308-16 stainless steel weld metal

    Energy Technology Data Exchange (ETDEWEB)

    Edmonds, D. P.; Vandergriff, D. M.; Gray, R. J.

    1978-01-01

    The effects of ferrite content on the properties of type 308 stainless steel shielded metal-arc (SMA) welds were investigated. Welds were made at four levels of ferrite content ranging from 2 to 15 FN (Ferrite Number). Creep and tensile tests were performed. Specimens were aged at 1100/sup 0/C (593/sup 0/C) for times up to 10,000 h (36 Ms) and Charpy V-notch impact tests were performed. Chemical analysis of the original deposits, Magne-gage evaluations, and metallographic evaluation of tested specimens were made. The E308-16 stainless steel electrodes were formulated to produce SMA welds with 2, 5, 9, and 15 FN. The ferrite number was made to vary by varying the nickel and chromium concentrations. Magne-gage determinations revealed that as-welded structures contained an average of 1.8, 4.2, 9.6, and 14.5 FN, respectively. Chemical anslysis of these deposits revealed no unusually high concentrations of tramp elements that would significantly affect mechanical properties. The extra low-ferrite electrodes were made with a different core wire, which produced deposits with slightly higher molybdenum concentrations. This variation in molybdenum should affect properties only minimally. From these chemical analyses and a constitutional diagram, ferrite concentrations were calculated, and the results correlated with the Magne-gage values

  9. Microstructural characterization and formation mechanism of abnormal segregation band of hot rolled ferrite/pearlite steel

    International Nuclear Information System (INIS)

    Feng, Rui; Li, Shengli; Zhu, Xinde; Ao, Qing

    2015-01-01

    In order to further reveal the microstructural characterization and formation mechanism of abnormal segregation band of hot rolled ferrite/pearlite steel, the microstructure of this type steel was intensively studied with Scanning Auger Microprobe (SAM), etc. The results show that severe C–Mn segregation exists in the abnormal segregation band region at the center of hot rolled ferrite/pearlite steel, which results from the Mn segregation during solidification process of the continuous casting slab. The C–Mn segregation causes relative displacement of pearlite transformation curve and bainite transformation curve of C curve in the corresponding region, leading to bay-like shaped C curve. The bay-like shaped C curve creates conditions for the transformation from supercooling austenite to bainite at relatively lower cooling rate in this region. The Fe–Mn–C Atomic Segregation Zone (FASZ) caused by C–Mn segregation can powerfully retard the atomic motion, and increase the lattice reconstruction resistance of austenite transformation. These two factors provide thermodynamic and kinetic conditions for the bainite transformation, and result in the emergence of granular bainitic abnormal segregation band at the center of steel plate, which leads to lower plasticity and toughness of this region, and induces the layered fracture. - Highlights: • Scanning Auger Microprobe (SAM) is applied in the fracture analysis. • The abnormal segregation band region appears obvious C–Mn segregation. • The C–Mn segregation leads to bay-like shaped C curve. • The C–Mn segregation leads to Fe–Mn–C Atomic Segregation Zone

  10. Martensitic/ferritic super heat-resistant 650 C steels - design of model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Knezevic, V.; Sauthoff, G. [Max-Planck-Inst. fuer Eisenforschung GmbH, Duesseldorf (Germany)

    2002-07-01

    Tempered martensitic/ferritic 9-12%Cr steels are now recognized to be the most potential materials for 650 C ultra super critical (USC) Power Plants. The degradation of long-term creep strength, as a result of microstructural changes during long-term exposure at the elevated temperature, is the main problem for this group of steels. Therefore, to achieve sufficient creep resistance during the entire service life it is necessary to stabilize the microstructure by alloying with elements which provide enough solid solution and precipitation strengthening and slow down diffusion. The aim of the present study is to investigate the effect of different types of precipitates as well as alloying elements on mechanical long-term properties of new ferritic 12%Cr steels. Fine distributions of stable precipitates which block the movement of subgrain boundaries (M{sub 23}C{sub 6} carbides, Laves phase) and dislocations (MX carbonitrides) and delay coarsening of microstructure is the key to high creep strength of such steels. Furthermore, additional Laves phase, which precipitates during service, is to strengthen the alloys when M{sub 23}C{sub 6} particles become less effective. Addition of Co is to achieve an initially 100% martensitic microstructure and moreover to slow down diffusion processes and consequently coarsening of particles. The partial substitution of Co by Cu and Mn is also investigated to reduce costs. The first results of mechanical tests of the studied model alloys have shown positive effects of the addition of W as Laves phase forming element, as well as of the MX forming elements Ta and Ti. Alloying with Co has also shown beneficial effects on the creep strength of model alloys. Further optimisation of composition and microstructure is in progress. (orig.)

  11. Standard test method for determination of reference temperature, to, for ferritic steels in the transition range

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 This test method covers the determination of a reference temperature, To, which characterizes the fracture toughness of ferritic steels that experience onset of cleavage cracking at elastic, or elastic-plastic KJc instabilities, or both. The specific types of ferritic steels (3.2.1) covered are those with yield strengths ranging from 275 to 825 MPa (40 to 120 ksi) and weld metals, after stress-relief annealing, that have 10 % or less strength mismatch relative to that of the base metal. 1.2 The specimens covered are fatigue precracked single-edge notched bend bars, SE(B), and standard or disk-shaped compact tension specimens, C(T) or DC(T). A range of specimen sizes with proportional dimensions is recommended. The dimension on which the proportionality is based is specimen thickness. 1.3 Median KJc values tend to vary with the specimen type at a given test temperature, presumably due to constraint differences among the allowable test specimens in 1.2. The degree of KJc variability among specimen types i...

  12. Diffraction plane dependence of elastic constants in ferritic steel in neutron diffraction stress measurement

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Ishiwata, Masayuki; Minakawa, Noriaki; Funahashi, Satoru; Root, J.H.

    1995-01-01

    Neutron diffraction measurement have been made to investigate the diffraction plane dependence of elastic constants in ferritic steel. The measured diffraction planes were 110, 220, 112, 222 and 200. In the measurement a small tensile specimen was loaded in the tensile test rig specially designed for a neutron diffractometer. The strains obtained for five diffraction planes increased almost in proportion to the applied stress up to 230 MPa nearly equivalent to the yield stress. The mean elastic constants obtained were E=243 GPa and ν=0.28 for 110, 220 and 112, 182 GPa and 0.31 for softest 200, and 268 GPa and 0.30 for stiffest 222, respectively. The bulk elastic constants, E=222 GPa and ν=0.29, measured by the strain gauges almost agreed with the mean values for 110, 220 and 112. The Kroner elastic model is found to account for the diffraction plane dependence of Young's modulus and Poisson's ratio of the ferritic steel. (author)

  13. Study of the microstructure evolution of ferritic stainless ODS steels during hot working

    International Nuclear Information System (INIS)

    Karch, Abdellatif

    2014-01-01

    The production of ODS steels involves a powder consolidation step usually using the hot extrusion (HE) process. The anisotropic properties of extruded materials, especially in the ODS ferritic grades (≥wt%12Cr), need a better understanding of the metallurgical phenomena which may occur during HE and lead to the observed microstructure. The hot working behavior of these materials is of particular interest. The methodology of this work includes the microstructure analysis after interrupted hot extrusion, hot torsion and hot compression (1000-1200 C) tests of ferritic steels with 14%Cr and different amounts in Ti and Y 2 O 3 . The microstructure evolution during hot extrusion process is associated with continuous dynamic recrystallization (CDRX). It leads to the creation of new grains by the formation of low angle boundaries, and then the increase of their misorientation under plastic deformation. The investigations highlight also the role of precipitation on the kinetics of this mechanism; it remains incomplete in the presence of fine and dense nano-precipitates. After hot deformation in torsion and compression, it is noticed that both precipitates and temperature deformation have a significant impact on the deformation mechanisms and microstructure evolution. Indeed, the CDRX is dominant when temperature and amount of reinforcement are limited. However, when they are increased, limited microstructure evolution is observed. In this case, the results are interpreted through a mechanism of strain accommodation at grain boundaries, with low dislocation activity in the bulk of the grains. (author) [fr

  14. Conversion of MX Nitrides to Modified Z-Phase in 9-12%Cr Ferritic Steels

    DEFF Research Database (Denmark)

    Cipolla, Leonardo

    The 9-12%Cr ferritic steels are extensively used in modern steam power plants at service temperature up to 620°C. Currently the best perform ing ferritic creep resistance steel is the ASTM Grade 92, whose high temperature strength has recently been assessed by European Creep Collaborative Committee....... With this purpose in mind, two 12%Cr model alloys, 12CrVNbN and 12CrVN, with ultra low carbon content, were manufactured. Both model alloys consisted of Cr-, V- and Nb-nitrides only. The first model alloy, 12CrVNbN, was especially designed to quickly convert the complex V- and Nb-nitrides into modified Z...... to identify all stages of MX conversion to Z-phase particle during ageings at 600°C, 650°C and 700°C up to 10 4 hours. Transmission Electron Microscopy (TEM) and X-Ray powder Diffraction (XRD) were applied to follow the microstructural evolution of the nitrides of model alloys during ageings: morphology...

  15. Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

  16. TEM Study of the Orientation Relationship Between Cementite and Ferrite in a Bainitic Low Carbon High Strength Low Alloy Steel

    OpenAIRE

    Illescas Fernandez, Silvia; Brown, A.P.; He, K.; Fernández, Javier; Guilemany Casadamon, Josep Maria

    2005-01-01

    Two different bainitic structures are observed in a steel depending on the sample heat treatment. The different types of bainitic structures exhibit different orientation relationships between cementite and the ferrite matrix. Upper bainite presents a Pitsch orientation relationship and lower bainite presents a Bagaryatski orientation relationship. Different heat treatments of low carbon HSLA steel samples have been studied using TEM in order to find the orientation relationshi...

  17. Tensile behavior of irradiated manganese-stabilized stainless steel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  18. Characterization of 2.25Cr1Mo welded ferritic steel plate by using diffractometric and ultrasonic techniques

    International Nuclear Information System (INIS)

    Cernuschi, F.; Ghia, S.; Albertini, G.; Ceretti, M.; Rustichelli, F.

    1995-12-01

    Four different techniques (X-ray and neutron diffraction, ultrasonic birefringence and incremental hole drilling method) were applied for evaluating residual stress in a butt-welded ferritic steel palte. Measurements were carried out both before and after welding. Effects of post-welding heat treatment is also considered. A comparison between results obtained by using four different techniques is done

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

    Science.gov (United States)

    Hunt, Ryan Matthew

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

  20. On the Capability of Nonmetallic Inclusions to Act as Nuclei for Acicular Ferrite in Different Steel Grades

    Science.gov (United States)

    Loder, Denise; Michelic, Susanne Katharina; Mayerhofer, Alexander; Bernhard, Christian

    2017-08-01

    Acicular ferrite nucleates intragranularly on nonmetallic inclusions, forming a microstructure with excellent fracture toughness. The formation of acicular ferrite is strongly affected by the size, content, and composition of nonmetallic inclusions, but also by the composition of the steel matrix. The potential of inclusions in medium carbon HSLA (high-strength low-alloyed) steels has been the main focus in the literature so far. The current study evaluates the acicular ferrite capability of various inclusions types in four different steel grades with carbon contents varying between 0.04 and 0.65 wt pct. The investigated steels are produced by melting experiments on a laboratory scale and subsequent heat treatment in a High-Temperature Laser Scanning Confocal Microscope. Inclusions are exclusively formed by deoxidation and desulfurization reactions. No synthetic particles are added to the melt. The inclusion landscape is analyzed by Scanning Electron Microscopy. Final ductility of the samples is evaluated based on performed tensile tests. Inclusion types in every steel grade are assessed regarding their nucleation potential always considering the interaction with the steel composition, especially focusing on the role of manganese. The effects of (Ti,Al)Ox-, MnS-, and MgO-containing inclusions are discussed in detail.

  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. Residual stress measurement round robin on an electron beam welded joint between austenitic stainless steel 316L(N) and ferritic steel P91

    OpenAIRE

    Javadi, Y.; Smith, M.C.; Abburi Venkata, K.; Naveed, N.; Forsey, A.N.; Francis, J.A.; Ainsworth, R.A.; Truman, C.E.; Smith, D.J.; Hosseinzadeh, F.; Gungor, S.; Bouchard, P. J.; Dey, H.C.; Bhaduri, A.K.; Mahadevan, S.

    2017-01-01

    This paper is a research output of DMW-Creep project which is part of a national UK programme through the RCUK Energy programme and India's Department of Atomic Energy. The research is focussed on understanding the characteristics of welded joints between austenitic stainless steel and ferritic steel that are widely used in many nuclear power generating plants and petrochemical industries as well as conventional coal and gas-fired power systems. The members of the DMW-Creep project have under...

  3. Safe Use Limits for Advanced Ferritic Steels in Ultra-Supercritical Power Boilers.

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, RW

    2003-11-03

    In 2000, a Cooperative Research and Development Agreement (CRADA) was undertaken between the Oak Ridge National Laboratory (ORNL) and the Babcock & Wilcox Company to examine the databases for advanced ferritic steels and determine the safe limits for operation in supercritical steam power boilers. The materials of interest included the vanadium-modified 9-12% Cr steels with 1-2% Mo or W. The first task involved a review of pertinent information and the down-selection of a steel of special interest. The long-time database for 9Cr-1Mo-V steel was found to be most satisfactory for the examinations, and this steel was taken to be representative of the group. The second task involved the collection of aged and service exposed samples for metallurgical and mechanical testing. Here, aged samples to 75,000 hours, laboratory-tested samples to 83,000 hours, and service-exposed sample with up to 143,000 hours exposure were collected. The third task involved mechanical testing of exposed samples. Creep-rupture testing to long times was undertaken. Variable stress and temperature testing was included. Results were compared against the prediction of damage models. These models seemed to be adequate for life prediction. The fourth task involved the metallurgical examination of exposed specimens. Changes in microstructure were compared against published information on the evolution of microstructures in 9Cr-Mo-V steels and the results were found to be consistent with expectations. The fifth task involved a survey of steam and fireside corrosion. Data from the service-exposed tubing was examined, and a literature survey was undertaken as part of an activity in support of ultra-supercritical steam boiler technology. The corrosion study indicated some concerns about long-time fireside corrosion and suggested temperature limits were needed for corrosive coal ash conditions.

  4. Modifications in structural, cation distribution and magnetic properties of {sup 60}Co gamma irradiated Li-ferrite

    Energy Technology Data Exchange (ETDEWEB)

    Mane, Maheshkumar L. [Department of Physics, Dr. Babasaheb Ambedkar Marathwada University, Aurangabad (M.S.) 431 004 (India); Shirsath, Sagar E., E-mail: shirsathsagar@hotmail.com [Department of Physics, Dr. Babasaheb Ambedkar Marathwada University, Aurangabad (M.S.) 431 004 (India); Dhage, Vinod N.; Jadhav, K.M. [Department of Physics, Dr. Babasaheb Ambedkar Marathwada University, Aurangabad (M.S.) 431 004 (India)

    2011-09-15

    Highlights: {yields} Gamma irradiation induced defects in lithium ferrite. {yields} Modifications in structural and magnetic properties. {yields} Fe{sup 3+} changes to Fe{sup 2+} after gamma irradiation. - Abstract: Polycrystalline samples of Li{sub 0.5}Fe{sub 2.5}O{sub 4} ferrite precursor were prepared by conventional standard double sintering ceramic technique and then irradiated with three different doses of {sup 60}Co gamma rays. The crystal structure and phase orientation of the irradiated and unirradiated samples of Li{sub 0.5}Fe{sub 2.5}O{sub 4} ferrite was done by using X-ray diffraction technique at room temperature. The lattice parameter of the studied samples increased due to the formation of Fe{sup 2+} ions under the ionizing effect of gamma radiation. The strain in the materials due to the irradiation was calculated from XRD data. Scanning electron microscope (SEM) studies indicate that the irradiation causes amorphization, especially at the grain boundaries. The cation distribution was calculated from XRD data analysis. By using cation distribution structural parameters such as theoretical lattice constant, ionic radii of available sites and the oxygen parameter 'u' have been calculated. The estimated cation distribution and other structural parameters shows strong influence of gamma rays on polycrystalline Li-ferrite. The magnetic properties of irradiated and unirradiated lithium ferrite were performed by using pulse field hysteresis loop technique at room temperature. Electrical properties such as diffusion coefficient and dielectric properties were carried out with the influence of gamma irradiation. Activation energy of diffusion process decreased after irradiation. The increase of diffusion coefficient with increasing dose rate of gamma irradiation was reinforced by the increase of Fe{sup 2+} ions and the displacement of metal ions from its original sites under the effect of gamma irradiation.

  5. Effect of Niobium on the Ferrite Continuous-Cooling-Transformation (CCT) Curve of Ultrahigh-Thickness Cr-Mo Steel

    Science.gov (United States)

    Lee, Sanghoon; Na, Hyesung; Kim, Byunghoon; Kim, Dongjin; Kang, Chungyun

    2013-06-01

    Pressure vessels made for petrochemical and power plants using Cr-Mo steel must be thick (≥400 mm) and have high tensile strength (≥600 MPa). However, the tensile strength in the middle portion of the vessel is very low as a result of ferrite formation. Therefore, research was performed to study the ferrite transformation that occurs in the middle portion of high-thickness Cr-Mo steel when Nb is added to it. The ferrite-formation start time of the continuous-cooling-transformation (CCT) curve decreased with an increase in Nb content until the latter reached 0.05 pct. On cooling from the austenitizing temperature, some of the NbC present at the austenitizing temperature of 1203 K (930 °C) goes into austenite solution in the temperature range of 1173 K to 1073 K (900 °C to 800 °C). However, the ferrite curve shifted to the left for the alloy containing 0.075 pct Nb. It is postulated that the surplus NbC could act as ferrite nucleation sites despite the lower cooling rate. As a result, the hardenability improved in the order of the following Nb content: 0.05 pct, 0.025 pct, 0 pct, and 0.075 pct.

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

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 10{sup 19} n cm{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 13} n cm{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  7. Positron annihilation studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Liszkay, L.; Molnar, B.

    1988-01-01

    Several annealing studies by positron annihilation (Doppler broadening, lifetime) on neutron irradiated Cr-Mo-V reactor pressure vessel steels (Soviet type 15Kh2MFA) regarding the influences of irradiation temperature, fluence of fast neutrons as well as different impurity contents are presented and discussed. A possibility of explaining the positron annihilation data by irradiation induced carbide formation is proposed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  9. Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang, E-mail: byunts@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Baek, Jong-Hyuk [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Anderoglu, Osman; Maloy, Stuart A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Toloczko, Mychailo B. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

    2014-06-01

    The HT9 ferritic/martensitic steel with a nominal chemistry of Fe(bal.)12%Cr1%MoVW has been used as a primary core material for fast fission reactors such as FFTF because of its high resistance to radiation-induced swelling and embrittlement. Both static and dynamic fracture test results have shown that the HT9 steel can become brittle when it is exposed to high dose irradiation at a relatively low temperature (<430 °C). This article aims at a comprehensive discussion on the thermal annealing recovery of fracture toughness in the HT9 steel after irradiation up to 3148 dpa at 378504 °C. A specimen reuse technique has been established and applied to this study: the fracture specimens were tested Charpy specimens or broken halves of Charpy bars (13 × 3 × 4 mm). The post-anneal fracture test results indicated that much of the radiation-induced damage can be recovered by a simple thermal annealing schedule: the fracture toughness was incompletely recovered by 550 °C annealing, while nearly complete or complete recovery occurred after 650 °C annealing. This indicates that thermal annealing is a feasible damage mitigation technique for the reactor components made of HT9 steel. The partial recovery is probably due to the non-removable microstructural damages such as void or gas bubble formation, elemental segregation and precipitation.

  10. Thermal annealing recovery of fracture toughness in HT9 steel after irradiation to high doses

    Science.gov (United States)

    Byun, Thak Sang; Baek, Jong-Hyuk; Anderoglu, Osman; Maloy, Stuart A.; Toloczko, Mychailo B.

    2014-06-01

    The HT9 ferritic/martensitic steel with a nominal chemistry of Fe(bal.)12%Cr1%MoVW has been used as a primary core material for fast fission reactors such as FFTF because of its high resistance to radiation-induced swelling and embrittlement. Both static and dynamic fracture test results have shown that the HT9 steel can become brittle when it is exposed to high dose irradiation at a relatively low temperature (steel after irradiation up to 3148 dpa at 378504 °C. A specimen reuse technique has been established and applied to this study: the fracture specimens were tested Charpy specimens or broken halves of Charpy bars (13 × 3 × 4 mm). The post-anneal fracture test results indicated that much of the radiation-induced damage can be recovered by a simple thermal annealing schedule: the fracture toughness was incompletely recovered by 550 °C annealing, while nearly complete or complete recovery occurred after 650 °C annealing. This indicates that thermal annealing is a feasible damage mitigation technique for the reactor components made of HT9 steel. The partial recovery is probably due to the non-removable microstructural damages such as void or gas bubble formation, elemental segregation and precipitation.

  11. Design of martensitic/ferritic heat-resistant steels for application at 650 deg. C with supporting thermodynamic modelling

    Energy Technology Data Exchange (ETDEWEB)

    Knezevic, V.; Balun, J. [Max-Planck-Institut fuer Eisenforschung GmbH, 40074 Duesseldorf (Germany); Sauthoff, G. [Max-Planck-Institut fuer Eisenforschung GmbH, 40074 Duesseldorf (Germany)], E-mail: g.sauthoff@mpie.de; Inden, G.; Schneider, A. [Max-Planck-Institut fuer Eisenforschung GmbH, 40074 Duesseldorf (Germany)

    2008-03-25

    In view of developing novel heat-resisting steels for applications in conventional power plants with service temperatures of 650 deg. C, a series of martensitic/ferritic model steels with 12 wt.%Cr were studied to achieve an increased creep resistance through additional alloying with various elements for controlled precipitation of M{sub 23}C{sub 6} carbides, MX carbonitrides and intermetallic Laves phase. The alloy design relied on thermodynamic simulation calculations using Thermo-Calc and DICTRA. The mechanical testing concentrated on creep at 650 deg. C for up to 8000 h. The alloy optimization resulted in creep rupture strengths above those of the martensitic/ferritic P92 steel. The work was part of a cooperative project within the German MARCKO program.

  12. Results from Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants - Irradiation Embrittlement of RPV Steels -

    International Nuclear Information System (INIS)

    Abe, Hiroaki; Onizawa, Kunio; Katsuyama, Jinya; Murakami, Kenta; Iwai, Takeo; Iwata, Tadao; Katano, Yoshio; Sekimura, Naoto; Nagai, Yasuyoshi; Toyama, Takeshi; Tamura, Satoshi

    2012-01-01

    microstructure. The typical RPV steel, A533B steel, is bainite, which consists of colonial distribution of carbides embedded in ferrite matrix. The mechanical property and its susceptibility to irradiation were investigated systematically. It was found that the regions nearby the carbide colonies were harder and more susceptible to irradiation hardening than the ferrite matrix. Irradiation induced hardening was proportional to the square root of dose up to 1dpa under irradiations with 2.8MeV Fe 2+ ions with dose rates ranging from 10 -5 to 10 -3 dpa/s at 563K. Electrical resistivity measurement was applied to achieve indispensable insights into diffusion of solute atoms for the correlation equations which includes microstructural evolutions based on solute and defect diffusion. Trapping of vacancies by solute atoms retards vacancy annihilation and enhance solute diffusion were evident. (author)

  13. 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)

  14. Kinetics of austenite-ferrite and austenite-pearlite transformations in a 1025 carbon steel

    Science.gov (United States)

    Hawbolt, E. B.; Chau, B.; Brimacombe, J. K.

    1985-04-01

    Isothermal and continuous-cooling transformation kinetics have been measured dilatometrically for the γ → α + γ' and γ' → P reactions in a 1025 steel. The isothermal transformation of austenite for each reaction was found to fit the Avrami equation after the fraction transformed was normalized to unity at the completion of the reaction and a transformation-start time was determined. The transformation kinetics under isothermal conditions therefore were characterized in terms of the n and b parameters from the Avrami equation together with the transformation-start times. The parameter n was found to be independent of temperature over the range studied (645 to 760 ‡C) and to have values of 0.99 and 1.33 for the ferrite and pearlite reactions, respectively. This indicates that the nucleation condition is essentially constant and site saturation occurs early in the transformation process. The continuous-cooling experiments were conducted at cooling rates of 2 to 150 ‡C per second to determine the transformation-start times for the ferrite and pearlite reactions and the completion time for transformation to pearlite under CCT conditions. Both reactions were found to obey the Additivity Principle for continuous cooling provided that the incubation (pre-transformation) period was not included in the transformation time. The isothermal transformation data and CCT transformation-start times have been incorporated in a mathematical model to predict continuous-cooling transformation kinetics that agree closely with measurements made at three cooling rates.

  15. 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.

  16. Properties of hot-rolled sheets from ferritic steel with increased strength

    Science.gov (United States)

    Perlovich, Yu.; Isaenkova, M.; Dobrokhotov, P.; Stolbov, S.; Bannykh, O.; Bannykh, I.; Antsyferova, M.

    2017-10-01

    Sheets from ferritic steel 3 mm thick with increased strength after thermal hardening were studied by use of various X-ray methods and mechanical testing. Rolling of steel was carried out at 1100°C with rather great reductions per pass, so that plastic deformation of metal spread by the significant distance from the surface. The texture of sheet proved to have two sharply different layers: the inner layer of ˜40% thick with the usual rolling texture of BCC metals and the external layer with the rolling texture of FCC metals. At that, within the intermediate layer the texture is weakened. Texture formation within the external layer is conditioned by the process of dynamical deformation ageing: interstitial impurities from atmosphere block dislocations, prevent from their slip and at increased temperatures promote their collective climb. As a result, the direction of lattice rotation as well as the final rolling texture change. Due to texture layering, by impact testing of the sheet the plane of crack propagation must be changed when this crack reaches the inner layer, and then an additional energy for its further movement is required. Thermal hardening of the sheet retains the type of rolling texture, though results in some its scattering, but at the same time the breaking point of steel grows twice owing to formation of intermetallic particles.

  17. Effect of surface finishing on the oxidation behaviour of a ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Ardigo-Besnard, M.R., E-mail: maria-rosa.ardigo-besnard@u-bourgogne.fr [Laboratoire Interdisciplinaire Carnot de Bourgogne (ICB), UMR 6303 CNRS—Université de Bourgogne Franche-Comté, BP 47870, 21078 Dijon Cedex (France); Popa, I.; Heintz, O.; Chassagnon, R. [Laboratoire Interdisciplinaire Carnot de Bourgogne (ICB), UMR 6303 CNRS—Université de Bourgogne Franche-Comté, BP 47870, 21078 Dijon Cedex (France); Vilasi, M. [Institut Jean Lamour, UMR 7198 CNRS—Université de Lorraine, Parc de Saurupt, 54011 Nancy (France); Herbst, F. [Laboratoire Interdisciplinaire Carnot de Bourgogne (ICB), UMR 6303 CNRS—Université de Bourgogne Franche-Comté, BP 47870, 21078 Dijon Cedex (France); Girardon, P. [APERAM, Centre de Recherche, BP15, 62330 Isbergues (France); Chevalier, S. [Laboratoire Interdisciplinaire Carnot de Bourgogne (ICB), UMR 6303 CNRS—Université de Bourgogne Franche-Comté, BP 47870, 21078 Dijon Cedex (France)

    2017-08-01

    Highlights: • Study of surface finishing effect on the corrosion behaviour of a stainless steel. • Mirror polished samples were compared to as-rolled material. • Two oxidation mechanisms were identified depending on the surface finishing. • Before oxidation, native chemical phases are identical for both samples. • Subsurface dislocations generated by the polishing process promote Cr{sub 2}O{sub 3} formation. - Abstract: The corrosion behaviour and the oxidation mechanism of a ferritic stainless steel, K41X (AISI 441), were evaluated at 800 °C in water vapour hydrogen enriched atmosphere. Mirror polished samples were compared to as-rolled K41X material. Two different oxidation behaviours were observed depending on the surface finishing: a protective double (Cr,Mn){sub 3}O{sub 4}/Cr{sub 2}O{sub 3} scale formed on the polished samples whereas external Fe{sub 3}O{sub 4} and (Cr,Fe){sub 2}O{sub 3} oxides grew on the raw steel. Moreover, isotopic marker experiments combined with SIMS analyses revealed different growth mechanisms. The influence of surface finishing on the corrosion products and growth mechanisms was apprehended by means of X-ray photoelectron spectroscopy (XPS) and residual stress analyses using XRD at the sample surfaces before ageing.

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

    Science.gov (United States)

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

    2014-09-01

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

  19. Microstructural Evolution of Thor™ 115 Creep-Strength Enhanced Ferritic Steel

    Science.gov (United States)

    Ortolani, Matteo; D'Incau, Mirco; Ciancio, Regina; Scardi, Paolo

    2017-12-01

    A new ferritic steel branded as Thor™ 115 has been developed to enhance high-temperature resistance. The steel design combines an improved oxidation resistance with long-term microstructural stability. The new alloy, cast to different product forms such as plates and tubes, was extensively tested to assess the high-temperature time-dependent mechanical behavior (creep). The main strengthening mechanism is precipitation hardening by finely dispersed carbide and nitride phases. Information on the evolution of secondary phases and time-temperature-precipitation behavior of the alloy, essential to ensure long-term property stability, was obtained by scanning transmission electron microscopy with energy dispersive spectroscopy, and by X-ray Powder Diffraction on specimens aged up to 50,000 hours. A thermodynamic modeling supports presentation and evaluation of the experimental results. The evolution of precipitates in the new alloy confirms the retention of the strengthening by secondary phases, even after long-term exposure at high temperature. The deleterious conversion of nitrides into Z phase is shown to be in line with, or even slower than that of the comparable ASME grade 91 steel.

  20. Design of model alloys for martensitic/ferritic super heat-resistant 650 C steels

    Energy Technology Data Exchange (ETDEWEB)

    Knezevic, V.; Vilk, J.; Inden, G.; Sauthoff, G.; Agamennone, R.; Blum, W.

    2001-07-01

    The key to high creep strength of steels, besides solid solution strengthening, are fine distributions of stable precipitates which block the movement of subgrain boundaries and dislocations and delay coarsening of microstructure. The aim of the present study is to design new super heat-resistant 12%Cr ferritic steels using basic principles and concepts of physical metallurgy, to test and optimise model alloys and to investigate and clarify their behaviour under long-term creep conditions with emphasis on microstructural stability. Taking into consideration recent world-wide developments of 9-12%Cr steels with screening of available data, a series of model alloys is designed, which is supported by theoretical calculations and simulations of the expected phase transformations and precipitation processes. The alloys are prepared and tested mechanically. The effects of different types of precipitates as well as alloying elements on mechanical long-term properties are investigated. In particular the Laves phase is studied, which precipitates during service and which is to strengthen the alloys when M{sub 23}C{sub 6} precipitate particles besides finely distributed other carbides and nitrides become less effective. The effects of various austenite-forming alloying elements are also studied. (orig.)

  1. 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 2 3C 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min-Gu; Kang, Nam Hyun [Pusan National University, Busan (Korea, Republic of); Moon, Joonoh; Lee, Tae-Ho; Lee, Chang-Hoon [Korea Institute of Materials Science, Changwon (Korea, Republic of); Kim, Hyoung Chan [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-04-15

    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{sub 2}3C{sub 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.

  3. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  4. Changing in tool steels wear resistance under electron irradiation

    International Nuclear Information System (INIS)

    Braginskaya, A.E.; Manin, V.N.; Makedonskij, A.V.; Mel'nikova, N.A.; Pakchanin, L.M.; Petrenko, P.V.

    1983-01-01

    The tool steels and alloys wear resistance under dry friction after electron irradiation has been studied. Electron irradiation of a wide variety of steels is shown to increase wear resistance. In this case phase composition and lattice parameters changes are observed both in matrix and carbides. The conclusion is drawn that an appreciable increase of steel wear resistance under electron irradiation can be explained both by carbide phase volume gain and changes in it's composition and the formation of carbide phase submicroscopic heterogeneities and, possibly, complexes of defects

  5. Analysis of the Grain Size Evolution for Ferrite Formation in Fe-C-Mn Steels Using a 3D Model Under a Mixed-Mode Interface Condition

    Science.gov (United States)

    Fang, H.; Mecozzi, M. G.; Brück, E.; van der Zwaag, S.; van Dijk, N. H.

    2018-01-01

    A 3D model has been developed to predict the average ferrite grain size and grain size distribution for an austenite-to-ferrite phase transformation during continuous cooling of an Fe-C-Mn steel. Using a Voronoi construction to represent the austenite grains, the ferrite is assumed to nucleate at the grain corners and to grow as spheres. Classical nucleation theory is used to estimate the density of ferrite nuclei. By assuming a negligible partition of manganese, the moving ferrite-austenite interface is treated with a mixed-mode model in which the soft impingement of the carbon diffusion fields is considered. The ferrite volume fraction, the average ferrite grain size, and the ferrite grain size distribution are derived as a function of temperature. The results of the present model are compared with those of a published phase-field model simulating the ferritic microstructure evolution during linear cooling of an Fe-0.10C-0.49Mn (wt pct) steel. It turns out that the present model can adequately reproduce the phase-field modeling results as well as the experimental dilatometry data. The model presented here provides a versatile tool to analyze the evolution of the ferrite grain size distribution at low computational costs.

  6. 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.

  7. Bending tests and magneto-elastic analysis of ferritic stainless steel plate in a magnetic field

    International Nuclear Information System (INIS)

    Horiguchi, Katsumi; Shindo, Yasuhide

    1998-01-01

    Experimental evidence and theoretical analysis are presented for the bending of a soft ferromagnetic beam plate in a magnetic field. The experiments were conducted in the bore of a superconducting magnet at room temperature. Ferritic stainless steel SUS 430 is here used as the cantilever specimen for bending test. The experiments show the predicted increase in the deflection and strain with increasing magnetic field. The theoretical analysis is based on a classical plate bending theory for magneto-elastic interactions in a soft ferromagnetic material. Numerical calculations are carried out, and the deflection and strain are obtained for several values of magnetic field and geometrical parameter. A comparison of the deflection and strain is made between theory and experiment and the agreement is good for the magnetic field considered. (author)

  8. Phase-separation, partitioning and precipitation in MA956, an ODS ferritic stainless steel

    International Nuclear Information System (INIS)

    Read, H.G.; Hono, K.

    1996-01-01

    The behaviours of as-received and recrystallised (homogenised) MA 956, an Al-containing Cr-rich ferritic stainless steel, aged at 475 C for up to 2900 hours have been investigated. Atom probe microanalysis of the decomposition products revealed that Al did not partition significantly to the Fe-rich phase after =600 hours ageing, contrary to thermodynamic predictions. Ageing to 2900 hours, however, resulted in partitioning. Further thermodynamic analysis showed that the chemical potential of Al in the Cr-rich α' phase increased more rapidly at later stages of phase separation. The wavelength and amplitude of decomposition were found to be significantly larger in aged as-received material compared to aged homogenised material, consistent with coarsening accelerated by the enhanced solute mobilities associated with the highly-dislocated as-received material. Ti- and Si-rich precipitates were found at the α/α' interfaces at later stages of ageing. (orig.)

  9. Plastic deformation-induced phosphorus segregation to ferrite grain boundaries in an interstitial free steel

    International Nuclear Information System (INIS)

    Chen, X.-M.; Song, S.-H.

    2010-01-01

    Research highlights: → Plastic deformation causes non-equilibrium grain boundary phosphorus segregation. → Deformation induced segregation increases with increasing deformation rate. → Non-equilibrium segregation is induced by supersaturated vacancy-phosphorus complex. → Model predictions show a reasonable agreement with the observations. - Abstract: Grain boundary concentration of phosphorus in an interstitial free steel is observed by virtue of Auger electron spectroscopy after the alloy is plastically deformed to different strains under different strain rates at a high temperature in the ferrite region. The results reveal that phosphorus segregates at grain boundaries during plastic deformation. The segregation increases with increasing deformation until reaching a steady value, and at the same deformation amount it increases with increasing strain rate. Model predictions are made, which shows a reasonable agreement between the predictions and the observations.

  10. Determination of austenite vs. α-ferrite in steel by neutron and X-ray diffraction

    International Nuclear Information System (INIS)

    Als-Nielsen, J.; Clausen, K.

    1984-06-01

    Neutron and X-ray diffraction studies for determining the relative content of fcc (austenite) and bcc (α-ferrite) phases in steel samples are reported. In addition to determine the relative content of phases the diffraction method also provides information about the strain fields in the sample by the concomitant broadening of diffraction peaks. Neutron diffraction has the advantage that large sample volumes (several cc) are probed, and the effect of texture can thus be eliminated. X-ray diffraction patterns can be registered in a short time thus allowing kinetic studies of phase changes during heat treatment or mechanical treatment. In addition it is possible to probe different surface thickness by utilizing different X-ray wavelengths. Measurements of this type can be carried out on a commercial contract basis in the Solid State Physics Division at Risoe National Laboratory. (author)

  11. Aluminum and aluminum/silicon coatings on ferritic steels by CVD-FBR technology

    International Nuclear Information System (INIS)

    Perez, F.J.; Hierro, M.P.; Trilleros, J.A.; Carpintero, M.C.; Sanchez, L.; Bolivar, F.J.

    2006-01-01

    The use of chemical vapor deposition by fluidized bed reactors (CVD-FBR) offers some advantages in comparison to other coating techniques such as pack cementation, because it allows coating deposition at lower temperatures than pack cementation and at atmospheric pressure without affecting the mechanical properties of material due to heat treatments of the bulk during coating process. Aluminum and aluminum/silicon coatings have been obtained on two different ferritics steels (P-91 and P-92). The coatings were analyzed using several techniques like SEM/EDX and XRD. The results indicated that both coatings were form by Fe 2 Al 5 intermetallic compound, and in the co-deposition the Si was incorporated to the Fe 2 Al 5 structure in small amounts

  12. Neutron diffraction study on anisotropy of strain age hardening in ferritic steel

    International Nuclear Information System (INIS)

    Suzuki, Tetsuya; Yamanaka, Keisuke; Ishino, Mayuko; Shinohara, Yasuhiro; Nagai, Kensuke; Tsuru, Eiji; Xu, Pingguang

    2012-01-01

    The work-hardening characteristics of anisotropic tensile deformations and the corresponding residual strain changes of pre-strained ferritic steels without and with aging treatment were investigated by using angle dispersive neutron diffraction and electron backscatter diffraction pattern analysis. The plastic deformation along the pre-strained direction leads to evident work-hardening at the beginning stage, showing discontinuous yielding behavior. Comparably, the plastic deformation perpendicular to the pre-strained direction shows continuously yielding. The tensile and compressive residual strains were found in the and grains along the pre-strained direction, respectively. It is also found that the difference in various oriented grains after strain aging become more evident along the pre-strained direction but smaller perpendicular to the pre-strained direction, revealing a higher work hardening capability in the former case than in the latter case. (author)

  13. Characterization of Phase Transformations and Stresses During the Welding of a Ferritic Mild Steel

    Science.gov (United States)

    Dye, D.; Stone, H. J.; Watson, M.; Rogge, R. B.

    2014-04-01

    The transient stresses and phase evolution have been characterized in the quasi-steady state produced around a gas tungsten arc welding torch in a plain carbon (ASTM 1018) steel using in situ neutron diffraction. A novel method has been developed to isolate the deviatoric or plane stress state in the presence of isotropic contributions to the lattice parameter, such as thermal expansion and solute content. The stress state was found to evolve in the anticipated manner, with compressive stresses ahead of the weld and tensile stresses behind the weld, in the weld and heat-affected zone, and compression in the far field behind the weld. In particular, the region of compression in the heat-affected zone adjacent to and just behind the welding torch expected from weld models was observed. The evolution of phase fraction around the weld was also determined using the technique and the stresses obtained from the ferrite phase.

  14. Material properties of tungsten coated F82H ferritic/martensitic steel as plasma facing armor

    International Nuclear Information System (INIS)

    Yahiro, Y.; Mitsuhara, M.; Nakashima, H.; Yoshida, N.; Hirai, T.; Tokitani, M.; Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato

    2009-01-01

    Two types of plasma spray tungsten coatings on ferritic/martensitic steel F82H made by vacuum plasma spray technique (VPS) and air plasma spray technique (APS) were examined in this study to evaluate the possibility as plasma-facing armor. The VPS-W/F82H showed superior properties. The porosity of the VPS-W coatings was about 1% and most of the pores were smaller than 1-2 μm and joining of W/F82H and W/W was fairly good. Thermal load tests indicated high potential of this coating as plasma-facing armor under thermal loading. In case of APS-W/F82H, however, porosity was 6% and thermal load properties were much worse than VPS-W/F82H. It is likely that surface oxidation during plasma spray process reduced joining properties. (author)

  15. Diffusion bonding of reduced activation ferritic steel F82H for demo blanket application

    International Nuclear Information System (INIS)

    Kurasawa, T.; Tamura, M.

    1996-01-01

    A reduced activation ferritic steel, a grade F82H developed by JAERI, is a promising candidate structural material for the blanket and the first wall of DEMO reactors. In the present study, diffusion bonding of F82H has been investigated to develop the fabrication procedures of the blanket box and the first wall panel with cooling channels embedded by F82H. The parameters examined are the bonding temperature (810-1050 C), bonding pressure (2-10 MPa) and roughness of the bonding surface (0.5-12.8 μR max ), and metallurgical examination and mechanical tests of the diffusion bonded joints have been conducted. From the tests, sufficient bonding was obtained under the temperatures of 840-1 050 C (compressive stress of 3-12 MPa), and it was found that heat treatment following diffusion bonding is essential to obtain the mechanical properties similar to that of the base metal. (orig.)

  16. Improvement in the DTVG detection method as applied to cast austeno-ferritic steels

    International Nuclear Information System (INIS)

    Francois, D.

    1996-05-01

    Initially, the so-called DTVG method was developed to improve detection and (lengthwise) dimensioning of cracks in austenitic steel assembly welds. The results obtained during the study and the structural similarity between austenitic and austeno-ferritic steels led us to carry out research into adapting the method on a sample the material of which is representative of the cast steels used in PWR primary circuit bends. The method was first adapted for use on thick-wall cast austeno-ferritic steel structures and was validated for zero ultrasonic beam incidence and for a flat sample with machine-finished reflectors. A second study was carried out notably to allow for non-zero ultrasonic beam incidence and to look at the method's validity when applied to a non-flat geometry. There were three principal goals to the research; adapting the process to take into account the special case of oblique ultrasonic beam incidence (B image handling), examining the effect of non-flat geometry on the detection method, and evaluating the performance of the method on actual defects (shrinkage cavities). We began by focusing on solving the problem of oblique incidence. Having decided on automatic refracted angle determination, the problem could only be solves by locking the algorithm on a representative image of the suspect material comprising an indicator. We then used a simple geometric model to quantify the deformation of the indicators on a B-scan image due to a non-flat translator/part interface. Finally, tests were carried out on measurements acquired from flat samples containing artificial and real defects so that the overall performance of the method after development could be assessed. This work has allowed the DTVG detection method to be adapted for use with B-scan images acquired with a non-zero ultrasonic beam incidence angle. Moreover, we have been able to show that for similar geometries to those of the cast bends and for deep defects the deformation of the indicators due

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

    Energy Technology Data Exchange (ETDEWEB)

    Gazder, Azdiar A., E-mail: azdiar@uow.edu.au [Electron Microscopy Centre, University of Wollongong, New South Wales 2500 (Australia); Al-Harbi, Fayez; Spanke, Hendrik Th. [School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, New South Wales 2522 (Australia); Mitchell, David R.G. [Electron Microscopy Centre, University of Wollongong, New South Wales 2500 (Australia); Pereloma, Elena V. [Electron Microscopy Centre, University of Wollongong, New South Wales 2500 (Australia); School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, New South Wales 2522 (Australia)

    2014-12-15

    Using a combination of electron back-scattering diffraction and energy dispersive X-ray spectroscopy data, a segmentation procedure was developed to comprehensively distinguish austenite, martensite, polygonal ferrite, ferrite in granular bainite and bainitic ferrite laths in a thermo-mechanically processed low-Si, high-Al transformation-induced plasticity steel. The efficacy of the ferrite morphologies segmentation procedure was verified by transmission electron microscopy. The variation in carbon content between the ferrite in granular bainite and bainitic ferrite laths was explained on the basis of carbon partitioning during their growth. - Highlights: • Multi-condition segmentation of austenite, martensite, polygonal ferrite and ferrite in bainite. • Ferrites in granular bainite and bainitic ferrite segmented by variation in relative carbon counts. • Carbon partitioning during growth explains variation in carbon content of ferrites in bainites. • Developed EBSD image processing tools can be applied to the microstructures of a variety of alloys. • EBSD-based segmentation procedure verified by correlative TEM results.

  18. Control of Carbon Content in Steel by Introducing Proeutectoid Ferrite Transformation into Hot-Rolled Q&P Process

    Science.gov (United States)

    Li, Xiao-lei; Li, Yun-jie; Kang, Jian; Li, Cheng-ning; Yuan, Guo; Wang, Guo-dong

    2018-01-01

    A processing strategy involving primary and secondary carbon partitioning is proposed for the hot-rolled quenching and partitioning process through the introduction of proeutectoid ferrite transformation after rolling. The microstructures of the steels were characterized using scanning electron microscopy, transmission electron microscopy, electron probe microanalysis, and x-ray diffraction, and the mechanical properties were evaluated using a universal tensile machine. The blocky retained austenite that distributed along the ferrite grain boundaries was promoted based on the coupling action of the primary and secondary carbon partitioning, which enhanced the transformation-induced plasticity effect during deformation despite the high carbon concentration. A ferrite formation temperature range of 760 to 800 °C was proposed. In addition, from the perspective of industrialization, the observed `plateau trends' for the retained austenite fraction and product of strength and elongation suggest the availability of a wide processing window of 215-362 °C for controlling the coiling temperature.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  20. γ-irradiation induced zinc ferrites and their enhanced room-temperature ammonia gas sensing properties

    Science.gov (United States)

    Raut, S. D.; Awasarmol, V. V.; Ghule, B. G.; Shaikh, S. F.; Gore, S. K.; Sharma, R. P.; Pawar, P. P.; Mane, R. S.

    2018-03-01

    Zinc ferrite (ZnFe2O4) nanoparticles (NPs), synthesized using a facile and cost-effective sol-gel auto-combustion method, were irradiated with 2 and 5 kGy γ-doses using 60Co as a radioactive source. Effect of γ-irradiation on the structure, morphology, pore-size and pore-volume and room-temperature (300 K) gas sensor performance has been measured and reported. Both as-synthesized and γ-irradiated ZnFe2O4 NPs reveal remarkable gas sensor activity to ammonia in contrast to methanol, ethanol, acetone and toluene volatile organic gases. The responses of pristine, 2 and 5 kGy γ-irradiated ZnFe2O4 NPs are respectively 55%, 66% and 81% @100 ppm concentration of ammonia, signifying an importance of γ-irradiation for enhancing the sensitivity, selectivity and stability of ZnFe2O4 NPs as ammonia gas sensors. Thereby, due to increase in surface area and crystallinity on γ-doses, the γ-irradiation improves the room-temperature ammonia gas sensing performance of ZnFe2O4.

  1. Preparation of ferritic 17%Cr ODS steel by mechanical alloying from prealloyed steel powder

    Czech Academy of Sciences Publication Activity Database

    Hadraba, Hynek; Husák, Roman; Kuběna, Ivo; Bureš, R.; Fáberová, M.; Strečková, M.

    2014-01-01

    Roč. 14, č. 4 (2014), s. 222-227 ISSN 1335-8987 R&D Projects: GA ČR(CZ) GA14-25246S Institutional support: RVO:68081723 Keywords : ODS steel * mechanical alloying * hot rolling Subject RIV: JG - Metallurgy

  2. Effects of titanium on ferrite continuous cooling transformation curves of high-thickness Cr-Mo steels

    Science.gov (United States)

    Lee, Sang-Hoon; Na, Hye-Sung; Park, Gi-Deok; Kim, Byung-Hoon; Song, Sang-Woo; Kang, Chung-Yun

    2013-09-01

    The effect of Ti on the ferrite-phase transformation in the middle portion of high-thickness Cr-Mo steel vessels was studied. The phase diagrams and ferrite continuous cooling transformation (CCT) curves were calculated thermodynamically, and dilatometry tests were performed to determine the start and finish times of the ferrite transformation. When the Ti concentration was 0.015 mass%, Δ( F s - F f ) of ferrite CCT curve decreased owing to an increase in the concentration of Mn dissolved as a result of (Mn, Ti) oxide formation. When the Ti concentration was 0.03 mass% or greater, the ferrite CCT curves shifted considerably to the right along the time axis owing to an increase in Ti oxide formation and the precipitation of Ti4C2S2, both of which affect the concentration of Mn dissolved in the austenite matrix. As a result, a completely bainitic structure was obtained when the Ti concentration was 0.03 mass% or greater.

  3. Compatibility of ferritic steels with sintered Li2O pellets in a flowing-helium environment

    International Nuclear Information System (INIS)

    Chopra, O.K.; Kurasawa, T.; Smith, D.L.

    1983-01-01

    The compatibility of ferritic HT-9 alloy and Fe-9Cr-1Mo steel with Li 2 O pellets has been investigated at 823 K (550 0 C) in flowing helium containing 93 or 1 ppM H 2 O and 1 ppM H 2 . The results indicate that the alloy specimens gain weight whereas the Li 2 O pellets lose weight after exposure. There is a net loss in weight of the total reaction couple. Both steels develop an iron-rich outer scale and chromium-rich subscale. The reaction rates in helium containing 93 ppM H 2 O are greater than in helium containing 1 ppM H 2 O. The depth of internal penetration for specimens exposed in helium with 1 ppM H 2 O reaches a constant value after approx. 3.6 Ms. The specimens exposed in helium containing 93 ppM H 2 O show a gradual increase in penetration up to 7.2 Ms. For both moisture contents, the total scale thickness follows a power law and the reaction rates decrease with time. The weight losses for Li 2 O pellets follow a linear law and yield values of 12.2 and 3.8%/year in helium with 93 and 1 ppM H 2 O, respectively

  4. Development of a High-Strength Ultrafine-Grained Ferritic Steel Nanocomposite

    Science.gov (United States)

    Rahmanifard, Roohollah; Farhangi, Hasan; Novinrooz, Abdul Javad; Moniri, Samira

    2013-02-01

    This article describes the microstructural and mechanical properties of 12YWT oxide-dispersion-strengthened (ODS)-ferritic steel nanocomposite. According to the annealing results obtained from X-ray diffraction line profile analysis on mechanically alloyed powders milled for 80 hours, the hot extrusion at 1123 K (850 °C) resulted in a nearly equiaxed ultrafine structure with an ultimate tensile strength of 1470 MPa, yield strength of 1390 MPa, and total elongation of 13 pct at room temperature comparable with high-strength 14YWT ODS steel. Maximum total elongation was found at 973 K (600 °C) where fractography of the tensile specimen showed a fully ductile dimple feature compared with the splitting cracks and very fine dimpled structure observed at room temperature. The presence of very small particles on the wall of dimples at 1073 K (800 °C) with nearly chemical composition of the matrix alloy was attributed to the activation of the boundaries decohesion mechanism as a result of diffusion of solute atoms. The results of Charpy impact test also indicated significant improvement of transition temperature with respect to predecessor 12YWT because of the decreased grain size and more homogeneity of grain size distribution. Hence, this alloy represented a good compromise between the strength and Charpy impact properties.

  5. Fracture toughness in the transition region of a carbon steel and a ferritic nodular cast iron

    International Nuclear Information System (INIS)

    Nakano, Keishi; Yasunaka, Takashi

    1995-01-01

    In order to characterize the fracture toughness in the ductile-brittle transition region for thick-walled cylinders of ASME SA350 Gr.LF5 carbon steel and JIS FCD300LT ferritic nodular cast iron, elastic-plastic fracture toughness tests were carried out. The specimens were fatigue precracked compact tension (CT) specimens of 25mm in thickness. The tensile testing machines used were Instron type, electrohydraulic type and drop-weight type ones. In the static fracture toughness test on a FCD300LT cast iron, CT specimens were often fractured at somewhat higher loads after the initiation of pop-in cracks. Although the scatter of pop-in fracture toughness was small, the values of critical J-integral at the unstable brittle fracture scattered largely. In the transition region of SA350 steel, the initiation of pop-in crack was not observed, and fracture toughness scattered largely. At the propagation of the unstable crack near the transition temperature, the Weibull distribution provides good fits for the critical CTOD and the critical J-integral. This distribution can be mainly interpreted by the scatter of the distance between the precrack tip and the origin of unstable brittle fracture. (author)

  6. Formation and coalescence of strain localized regions in ferrite phase of DP600 steels under uniaxial tensile deformation

    Energy Technology Data Exchange (ETDEWEB)

    Alaie, A., E-mail: amir_alaie@yahoo.com [Department of Mechanical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of); Kadkhodapour, J. [Department of Mechanical Engineering, Shahid Rajaee Teacher Training University, Tehran (Iran, Islamic Republic of); Institute for Materials Testing, Materials Science and Strength of Materials (IMWF), University of Stuttgart, Stuttgart (Germany); Ziaei Rad, S. [Department of Mechanical Engineering, Isfahan University of Technology, Isfahan (Iran, Islamic Republic of); Asadi Asadabad, M. [Materials Research School, Isfahan (Iran, Islamic Republic of); Schmauder, S. [Institute for Materials Testing, Materials Science and Strength of Materials (IMWF), University of Stuttgart, Stuttgart (Germany)

    2015-01-19

    In this study the key factors in the creation and coalescence of strain localization regions in dual-phase steels were investigated. An in-situ tensile setup was used to follow the microscopic deformation of ferrite phase inside the microstructure of DP600 steel. The test was continued until the specimen was very close to final failure. The captured scanning electron microscopy (SEM) micrographs enabled us to directly observe the evolution of deformation bands as a contour of strain distribution in the ferrite matrix. The image processing method was used to quantify the ferrite microscopic strains; the obtained strain maps were superimposed onto the SEM micrographs. The results revealed important deformational characteristics of the microstructure at the microscopic level. It was observed that despite the formation of slip bands inside the large grains during the early stages of deformation, the large ferrite grains did not contribute to the formation of high-strain bands until the final stages of severe necking. The behavior of voids and initial defects inside the localization bands was also studied. In the final stages of deformation, cracks were observed to preferentially propagate in the direction of local deformation bands and to coalescence with each other to form the final failure lines in the microstructure. It was observed that in the final stages of deformation, the defects or voids outside the deformation bands do not contribute to the final failure mechanisms and could be considered to be of minor importance.

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

    Science.gov (United States)

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

    2017-02-01

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

  8. Steam oxidation behavior of high strength newly developed ferritic/martensitic steels at 650 C

    Energy Technology Data Exchange (ETDEWEB)

    Agueero, Alina; Gonzalez, Vanessa; Gutierrez, Marcos [Instituto Nacional de Tecnica Aeroespacial, Torrejon de Ardoz (Spain); Mayr, Peter [Massachusetts Inst. of Tech., Cambridge (United States). Dept. of Materials Science and Engineering; Spiradek-Hahn, Krystina [Austrian Institute of Technology GmbH (AIT), Seibersdorf (Austria)

    2010-07-01

    The efficiency of thermal power plants is currently limited by the strength and the oxidation resistance of the commercially available ferritic steels. The higher operating pressures and temperatures, essential to increase efficiency, impose important requirements on the materials from both the mechanical and chemical stability perspective. For instance, a creep rupture strength of 100 MPa after 100.000 hours at 650 C has been defined as the target for new steel development. Moreover, steam oxidation resistance is required as otherwise, at temperatures higher than 600 C, the resulting thick oxide scales will spall, causing blockage on bends as well as overheating of heat exchangers due to a thermal insulation effect, erosion of down-stream components and loss of cross-section in critical components such as blades. It has been shown that in general, a Cr wt. % higher than 9 is required for acceptable oxidation rates at 650 C, but such high Cr content results in a reduction of the creep strength. As an exception, several 9 wt. % steels developed by Abe which also containing Si and Mn, exhibit resistance to steam oxidation but only after having been subjected to a pre-oxidation heat treatment at 650 C for many hours. Substantial efforts are being carried out in Europe, North America and Japan attempting to design and produce steels with these properties. The steam oxidation behavior of high strength new alloys, such as CB2 and FT7 steels developed within the frame of European COST Actions 522 and 536, as well as of a NIMS developed B containing 9Cr3W3CoVNb (NPM) martensitic steel, was studied by exposing these materials to pure flowing steam in the laboratory for periods of time in excess of 10,000 h at 650 C. CB2 and FT7 have similar creep strength to P92 whereas NPM reaches 21,000 h at 100 MPa exceeding by far that of P92 according to the ECCC values. (orig.)

  9. The Studies of Irradiation Hardening of Stainless Steel Reactor Internals under Proton and Xenon Irradiation

    OpenAIRE

    Xu, Chaoliang; Zhang, Lu; Qian, Wangjie; Mei, Jinna; Liu, Xiangbing

    2016-01-01

    Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation ...

  10. THE INFLUENCE MECHANISM OF FERRITE GRAIN SIZE ON STRENGTH STRESS AT THE FATIGUE OF LOW-CARBON STEEL

    Directory of Open Access Journals (Sweden)

    I. A. Vakulenko

    2014-01-01

    Full Text Available Purpose. Explanation of the influence mechanism of ferrite grain size on the fatigue strength of low-carbon steel. Methodology. Material for research is the low-carbon steel with 0.1% of carbon contnent. The different size of ferrite grain was obtained due to varying the degree of cold plastic deformation and temperature of annealing. The estimation of grain size was conducted using methodologies of quantitative metallography. The microstructure of metal was investigated under a light microscope with increase up to 1500 times. As a fatigue response the fatigue strength of metal – a maximal value of load amplitude with endless endurance limit of specimen was used. Fatigue tests were carried out using the test machine «Saturn-10», at the symmetric cycle of alternating bend loading. Findings. On the basis of research the dependence for fatigue strength of low-carbon steel, which is based on an additive contribution from hardening of solid solution by the atoms of carbon, boundary of the ferrite grain and amount of mobile dislocations was obtained. It was established that as the grainy structure of low-carbon steel enlarges, the influence of grain size on the fatigue strength level is reduced. For the sizes of grains more than 100 mcm, basic influence on fatigue strength begins to pass to the solid solution hardening, which is determined by the state of solid solution of introduction. Originality. From the analysis of the obtained dependences it ensues that with the increase of ferrite grain size the required amount of mobile dislocations for maintenance of conditions for spreading plastic deformation becomes less dependent from the scheme of metal loading. Practical value. The obtained results present certain practical interest when developing of recommendations, directed on the increase of resource of products work from low-carbon steels in the conditions of cyclic loading. Estimation of separate contribution of the studied processes of

  11. Effect of mechanical alloying atmosphere on the microstructure and Charpy impact properties of an ODS ferritic steel

    International Nuclear Information System (INIS)

    Oksiuta, Z.; Baluc, N.

    2009-01-01

    Two types of oxide dispersion strengthened (ODS) ferritic steels, with the composition of Fe-14Cr-2W-0.3Ti-0.3Y 2 O 3 (in weight percent), have been produced by mechanically alloying elemental powders of Fe, Cr, W, and Ti with Y 2 O 3 particles either in argon atmosphere or in hydrogen atmosphere, degassing at various temperatures, and compacting the mechanically alloyed powders by hot isostatic pressing. It was found in particular that mechanical alloying in hydrogen yields a significant reduction in oxygen content in the materials, a lower dislocation density, and a strong improvement in the fast fracture properties of the ODS ferritic steels, as measured by Charpy impact tests.

  12. AFM and TEM study of cyclic slip localization in fatigued ferritic X10CrAl24 stainless steel

    Czech Academy of Sciences Publication Activity Database

    Man, Jiří; Petrenec, Martin; Obrtlík, Karel; Polák, Jaroslav

    2004-01-01

    Roč. 52, č. 19 (2004), s. 5551-5561 ISSN 1359-6454 R&D Projects: GA ČR GA106/00/D055; GA ČR GA106/01/0376; GA AV ČR IAA2041201 Institutional research plan: CEZ:AV0Z2041904 Keywords : ferritic steel * fatigue * persistent slip band Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 3.490, year: 2004

  13. Protecting against failure by brittle fracture in ferritic-steel shipping containers greater than four-in. thick

    International Nuclear Information System (INIS)

    Schwartz, M.W.

    1982-12-01

    This report presents methods for protecting against brittle fracture spent-fuel shipping containers made from ferritic-steel forgings greater than four in. thick. Both fracture arrest and fracture initiation criteria were examined as bases for establishing requirements for the design and selection of materials for shipping containers. This report also includes a discussion of the brittle-fracture sensitivity of the containers to various processes used in container fabrication. 19 figures, 3 tables

  14. The mechanism of {gamma}{sub 2} formation process in lathy ferrite microstructure of duplex stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Knezevic, V.R.; Cvijovic, Z.M.; Mihajlovic, D.V. [Belgrade Univ. (Yugoslavia). Faculty of Technology and Metallurgy

    2000-07-01

    The formation mechanism and morphology of secondary austenite ({gamma}{sub 2}) were investigated using light and scanning electron microscopy in lathy ferrite microstructure of a cast 22/7/2 copper-bearing duplex stainless steel isothermally transformed at temperatures from 800 C to 1150 C. It was established that its formation obeys the Johnson-Mehl equation and is temperature and time dependent. The precipitation sequence processes have been discussed and some aspects of precipitation behaviour are explained. (orig.)

  15. Tensile properties and hardness of two types of 11Cr-ferritic/martensitic steel after aging up to 45,000 h

    Directory of Open Access Journals (Sweden)

    Y. Yano

    2016-12-01

    Full Text Available The relationship among tensile strength, Vickers hardness and dislocation density for two types of 11Cr-ferritic/martensitic steel (PNC-FMS was investigated after aging at temperatures between 400 and 800°C up to 45,000h and after neutron irradiation. A correlation between tensile strength and Vickers hardness was expressed empirically. The linear relationship for PNC-FMS wrapper material was observed between yield stress and the square of dislocation density at RT and aging temperature according to Bailey–Hirsch relation. Therefore, it was clarified that the correlation among dislocation density, tensile strength and Vickers hardness to aging temperature was in good agreement. On the other hand, the relationship between tensile strength ratio when materials were tested at aging temperature and Larson–Miller parameter was also in excellent agreement with aging data between 400 and 700°C. It was suggested that this correlation could use quantitatively for separately evaluating irradiation effects from neutron irradiation data containing both irradiation and aging effects.

  16. Hydrogen transport through stainless steel under plasma irradiation

    Science.gov (United States)

    Airapetov, A. A.; Begrambekov, L. B.; Kaplevsky, A. S.; Sadovskiy, Ya A.

    2016-01-01

    The paper presents the results of investigation of gas exchange through stainless steel surface of the plasma chamber under irradiation with hydrogen atoms in oxygen atmosphere or oxygen contaminated hydrogen plasma. Dependence of this process on various irradiation parameters, such as the metal temperature, energy of irradiating ions, gas composition of plasma are studied. It is shown, that desorption from stainless steel is activated with the increase of the plasma chamber walls temperature and energy of irradiating ions. Hydrogen release occurs also under irradiation of the walls by helium and argon plasmas added with oxygen, however the amount of released hydrogen is several times lower than in the case of irradiation with oxygen contaminated deuterium plasma.

  17. Microstructure and mechanical properties of friction stir welded 18Cr–2Mo ferritic stainless steel thick plate

    International Nuclear Information System (INIS)

    Han, Jian; Li, Huijun; Zhu, Zhixiong; Barbaro, Frank; Jiang, Laizhu; Xu, Haigang; Ma, Li

    2014-01-01

    Highlights: • We focus on friction stir welding of 18Cr–2Mo ferritic stainless steel thick plate. • We produce high-quality joints with special tool and optimised welding parameters. • We compare microstructure and mechanical properties of steel and joint. • Friction stir welding is a method that can maintain the properties of joint. - Abstract: In this study, microstructure and mechanical properties of a friction stir welded 18Cr–2Mo ferritic stainless steel thick plate were investigated. The 5.4 mm thick plates with excellent properties were welded at a constant rotational speed and a changeable welding speed using a composite tool featuring a chosen volume fraction of cubic boron nitride (cBN) in a W–Re matrix. The high-quality welds were successfully produced with optimised welding parameters, and studied by means of optical microscopy (OM), scanning electron microscopy (SEM), electron back-scattered diffraction (EBSD) and standard hardness and impact toughness testing. The results show that microstructure and mechanical properties of the joints are affected greatly, which is mainly related to the remarkably fine-grained microstructure of equiaxed ferrite that is observed in the friction stir welded joint. Meanwhile, the ratios of low-angle grain boundary in the stir zone regions significantly increase, and the texture turns strong. Compared with the base material, mechanical properties of the joint are maintained in a comparatively high level

  18. The role of nitrogen in the preferential chromium segregation on the ferritic stainless steel (1 1 1) surface

    International Nuclear Information System (INIS)

    Yuhara, J.; Matsui, T.

    2010-01-01

    The temperature dependence on the segregation behavior of the ferritic stainless steel single crystal (1 1 1) surface morphology has been examined by scanning tunneling microscopy (STM), Auger electron spectroscopy (AES), and low energy electron diffraction (LEED). AES clearly showed the surface segregations of chromium and nitrogen upon annealing. Nanoscale triangular chromium nitride clusters were formed around 650 deg. C and were regularly aligned in a hexagonal configuration. In contrast, for the ferritic stainless steel (1 1 1) surface with low-nitrogen content, chromium and carbon were found to segregate on the surface upon annealing and Auger spectra of carbon displayed the characteristic carbide peak. For the low-nitrogen surface, LEED identified a facetted surface with (2 x 2) superstructure at 650 deg. C. High-resolution STM identified a chromium carbide film with segregated carbon atoms randomly located on the surface. The facetted (2 x 2) superstructure changed into a (3 x 3) superstructure with no faceting upon annealing at 750 deg. C. Also, segregated sulfur seems to contribute to the reconstruction or interfacial relaxation between the ferritic stainless steel (1 1 1) substrate and chromium carbide film.

  19. Mechanical properties and fracture features of low-activation ferritic-martensitic steel EK-181 at subzero temperatures

    Science.gov (United States)

    Polekhina, N. A.; Litovchenko, I. Yu.; Tyumentsev, A. N.; Kravchenko, D. A.; Chernov, V. M.; Leontyeva-Smirnova, M. V.

    2017-12-01

    The short-term strength and plastic properties of ferritic-martensitic steel EK-181, as well as the features of its plastic deformation and fracture in the temperature range from 20 to -196°C are investigated by an active tensile deformation method. A significant increase in the temperature dependence of the steel yield strength in the interval of the ductile-to-brittle transition is observed. No qualitative changes in the fracture pattern of the samples are revealed in the region of this interval. The fractograms taken after deformation at several temperatures differ only in the relative fractions of the ductile and brittle components.

  20. 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.

  1. Development of PIE techniques for irradiated LWR pressure vessel steels

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  2. Beneficial effect of Re on the long-term creep strength of high Cr ferritic heat resistant steels

    Energy Technology Data Exchange (ETDEWEB)

    Hashizume, Ryokichi; Tamura, Osam [The Kansai Electric Power Company Inc., Amagasaki (Japan); Miki, Kazuhiro; Azuma, Tsukasa [The Japan Steel Works, Ltd., Muroran (Japan). Muroran Research Lab.; Ishiguro, Tohru [The Japan Steel Works, Ltd., Muroran (Japan). Muroran Plant; Murata, Yoshinori; Morinaga, Masahiko [Nagoya Univ. (Japan). Dept. of Materials Science and Engineering

    2010-07-01

    9-12%Cr ferritic steels were designed by using the d-electrons concept for the use of steam turbine rotors operated in the USC power plants at the steam temperature of 620 C to 650 C. The crucial issue for the design is to suppress the deterioration of the long-term creep strength by alloying. First, the Re addition was found to give a beneficial effect on the creep strength of a 10%Cr-4%W steel. Then, the creep tests were performed with the six Re-free and 3.5%W ferritic steels to get an optimum Cr content in the range of 8.5% to 11.5%. As a result, it was found that an excess amount of Cr yielded a detrimental effect on the creep properties, and 9%Cr steel was the best in view of the very long-term creep strength tested in the condition of 650 C, 98 MPa. Subsequently, a series of creep tests was conducted with the steels by fixing at 9%Cr, but by varying the W content from 2% to 4% and the Re content from 0% to 0.5%. From the prolonged creep tests and creep rate tests for more than 30,000-40,000 hours, it was shown that the 9Cr-4W without Re steel had the long creep rupture life for more than 32,000 hours at 650 C 0,98 MPa and the 9Cr-4W-0.5Re steel had the longer one for more than 43,000 hours. It was the longest creep rupture life among all the tested steels. (orig.)

  3. Study of the precipitation and of the hardening microscopic mechanisms under irradiation in dilute ferritic alloys; Etude de la precipitation et des mecanismes microscopiques de durcissement sous irradiation dans des alliages ferritiques dilues

    Energy Technology Data Exchange (ETDEWEB)

    Mathon, M.H

    1995-07-01

    The copper precipitation plays a significant role in the embrittlement process of reactor vessel steels under neutron irradiation at 300 deg C. In order to understand the copper precipitation mechanisms, we have studied model ferritic binary FeCu and ternary alloys FeCuX (X=Mn,Ni, Cr, P). These materials have been either Irradiated with 2.5 MeV electrons In the 175-360 deg C temperature range or thermal aged at 500 deg C. The evolution of materials has been followed by resistivity measurements under irradiation, by small angle neutron scattering and by Vickers microhardness measurements. We have shown the similarity of copper precipitation under thermally ageing at 500 deg C and electron Irradiation at 300 deg C, in FeCu{sub 1,34%}. This result confirms that the main effect of electronic irradiation is to accelerate precipitation. Nevertheless, we have observed that irradiation induces an additional contribution to hardening attributed to point defect clusters. Concerning the ternary alloys, we observed that at 300 deg C the addition of a third element has no significant effect on the copper precipitation kinetic under irradiation but that at lower temperature manganese slows down precipitation kinetic. In order to reproduce the experimental results obtained on FeCu{sub 1,34%} by using a cluster kinetics model, we have to suppose that the precipitation is heterogeneous and controlled by interface reactions for the small size clusters. In addition, neutron or electron irradiated industrial steels have been studied by small angle neutron scattering. The results revealed the presence of nano-metric solute clusters which contain few copper atoms and which are not linked to the formation of displacement cascades. (author)

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

    Directory of Open Access Journals (Sweden)

    Anna Hojná

    2017-09-01

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

  5. Characterization of oxide nanoparticles in Al-free and Al-containing oxide dispersion strengthened ferritic steels.

    Science.gov (United States)

    Lee, Jae Hoon; Kim, Jeoung Han

    2013-09-01

    Oxide nanoparticles in oxide dispersion strengthened (ODS) ferritic steels with and without Al have been characterized by transmission electron microscopy. It is confirmed that most of the complex oxide particles consist of Y2TiO5 for 18Cr-ODS steel and YAlO3 or YAl5O12 for 18Cr5Al-ODS steel, respectivley. The addition of 5% Al in 18Cr-ODS steel leads to the formation of larger oxide particles and the reduction in their number density. For 18Cr-ODS steel, 87% of the oxide particles are coherent. The misfit strain of the coherent particles and a few semi-coherent particles is about 0.034 and 0.056, respectively. For 18Cr5Al-ODS steel, 75% of the oxide particles are semi-coherent, of which the misfit strain is 0.091 and 0.125, respectively. These results suggest that for the Al-containing ODS steel the Al addition accelerates the formation of semi-coherent oxide particles and its larger coherent and semi-coherent particles result in the larger misfit strain between the oxide particle and alloy matrix, indicating that the coherence of oxide nanoparticles in ODS steels is size-dependent.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-01

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

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

    Science.gov (United States)

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

    2017-02-01

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

  8. Impurity effects on reduced-activation ferritic steels developed for fusion applications

    International Nuclear Information System (INIS)

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

    2000-01-01

    Reduced-activation steels are being developed for fusion applications by restricting alloying elements that produce long-lived radioactive isotopes when irradiated in the fusion neutron environment. Another source of long-lived isotopes is the impurities in the steel. To examine this, three heats of reduced-activation martensitic steel were analyzed by inductively coupled plasma mass spectrometry for low-level impurities that compromise the reduced-activation characteristics: a 5-ton heat of modified F82H (F82H-Mod) for which an effort was made during production to reduce detrimental impurities, a 1-ton heat of JLF-1, and an 18-kg heat of ORNL 9Cr-2WVTa. Specimens from commercial heats of modified 9Cr-1Mo and Sandvik HT9 were also analyzed. The objective was to determine the difference in the impurity levels in the F82H-Mod and steels for which less effort was used to ensure purity. Silver, molybdenum, and niobium were found to be the tramp impurities of most importance. The F82H-Mod had the lowest levels, but in some cases the levels were not much different from the other heats. The impurity levels in the F82H-Mod produced with present technology did not achieve the low-activation limits for either shallow land burial or recycling. The results indicate the progress that has been made and what still must be done before the reduced-activation criteria can be achieved

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Boutard, J.-L., E-mail: jean-louis.boutard@cea.fr [Cabinet du Haut-Commissaire, CEA/Saclay, 91191 Gif sur Yvette Cedex (France); Badjeck, V. [LPS, UMR CNRS 8502, Building 510, Université Paris-Sud 11, 91405 Orsay Cedex (France); Barguet, L. [LAUM, UMR CNRS 6613, Building IAM – UFR Sciences, Avenue O. Messiaen, 72085 Le Mans Cedex 9 (France); Barouh, C. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Bhattacharya, A. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); CSNSM, UMR CNRS 8609, Université Paris-Sud 11, Buildings 104 and 108, 91405 Orsay Campus (France); Colignon, Y. [IM2NP, UMR CNRS 7334, Case 142, Faculté des Sciences, Campus de Saint Jérôme, Aix Marseille Université, 13397 Marseille Cedex 20 (France); Hatzoglou, C. [GPM, UMR CNRS 6634, Technopôle du Madrillet, Avenue de l’Université, BP12, 76801 Saint Etienne du Rouvray Cedex (France); Loyer-Prost, M. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Rouffié, A.L. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Sallez, N. [SIMAP, UMR CNRS 5266, INPG, Domaine Universitaire, 1130 rue de la Piscine, BP75, 38402 Saint Martin d’Hères Cedex (France); Salmon-Legagneur, H. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Schuler, T. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France)

    2014-12-15

    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.

  11. Effect of the radiation in the reference temperature T{sub 0} in ferritic steel; Efecto de la radiacion en la temperatura de referencia T{sub 0} en acero ferritico

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva O, A.; Gachuz M, M.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2004-07-01

    The present work studies the effect that produces the irradiation in ferritic steels (AISI 8620) on the reference temperature (T{sub 0}) that characterizes the tenacity to the fractures (K{sub JC}) of these materials obtaining this way a characteristic curve (Master Curve) of this steel. The approach of the 'Master curve' is based on the Astm E-1921. Following this standard the methodology of a sub size settled down in Charpy type test tubes. Due to this type of steels is used mainly in pressure vessels of the reactor in Nuclear Power plants, the fracture tenacity gives the rule at the moment for the verification of structural integrity of the pressure vessel of the reactor. (Author)

  12. Study on the activated laser welding of ferritic stainless steel with rare earth elements yttrium

    Science.gov (United States)

    Wang, Yonghui; Hu, Shengsun; Shen, Junqi

    2015-10-01

    The ferritic stainless steel SUS430 was used in this work. Based on a multi-component activating flux, composed of 50% ZrO2, 12.09 % CaCO3, 10.43 % CaO, and 27.49 % MgO, a series of modified activating fluxes with 0.5%, 1%, 2%, 5%, 10%, 15%, and 20% of rare earth (RE) element yttrium (Y) respectively were produced, and their effects on the weld penetration (WP) and corrosion resistant (CR) property were studied. Results showed that RE element Y hardly had any effects on increasing the WP. In the FeCl3 spot corrosion experiment, the corrosion rates of almost all the samples cut from welded joints turned out to be greater than the parent metal (23.51 g/m2 h). However, there was an exception that the corrosion rate of the sample with 5% Y was only 21.96 g/m2 h, which was even better than parent metal. The further Energy Dispersive Spectrometer (EDS) test showed the existence of elements Zr, Ca, O, and Y in the molten slag near the weld seam while none of them were found in the weld metal, indicating the direct transition of element from activating fluxes to the welding seam did not exist. It was known that certain composition of activating fluxes effectively restrain the loss of Cr element in the process of laser welding, and as a result, the CR of welded joints was improved.

  13. Microstructure and properties evaluations of spot-welded ferritic steel sheets via static magnetic field

    Science.gov (United States)

    Min, Ding; Yicheng, Wang

    2016-01-01

    Ferritic steel spot nuggets were produced with or without a static magnetic field. The microstructures and properties evaluations of the nuggets with or without a static magnetic field were investigated. Disordered columnar grains and some equiaxed grains among the columnar grains with a static magnetic field were discovered in this study. Based on the evaluations of the microstructure and properties, the nugget mechanisms, strengthening mechanisms, and infrared behavior of the joint were discovered. The diameter and strength of each nugget were improved with the application of a static magnetic field. The welding time and the welding force can both influence the nugget characteristics via a static magnetic field. The tensile strength of the spot joint regularly varied with magnetic field; the maximum value was 245 MPa, 11%, which was approximately 30% higher than that of the nugget without magnetic field (187 MPa, 3.8%). The magnetization force applied on the dendrite at the same time can cause the columnar dendrite to deform, break and deflect from the direction of solidification.

  14. X-ray stress measurement of ferritic steel using fourier analysis of Debye-Scherrer ring

    International Nuclear Information System (INIS)

    Fujimoto, Yohei; Sasaki, Toshihiko; Miyazaki, Toshiyuki

    2015-01-01

    In this study, X-ray stress measurements of ferritic steel based on Fourier analysis are conducted. Taira et al. developed the cosα method for X-ray stress measurements using a two-dimensional X-ray detector. Miyazaki et al. reported that the cosα method can be described more concisely by developing the Fourier series (the Fourier analysis method). The Fourier analysis method is expected to yield the stress measurement with an imperfect Debye-Scherrer ring and there is a possibility that the materials evaluation is different compared with the conventional method, that is, the sin 2 ψ method. In the Fourier analysis method, the strain measured by X-rays is developed as a Fourier series, and all the plane-stress components can be calculated from the Fourier series. In this study, the normal stress calculation was confirmed. In addition, the Fourier-analysis and cosα methods were used for X-ray stress measurements during a four-point bending test on a S45C test piece, and the effectiveness of the Fourier analysis method was confirmed. It was found that the experimental results from the Fourier analysis and cosα methods were nearly identical. In addition, the measurement accuracies of both the methods were equivalent. (author)

  15. Characterization of low-activation ferritic steel (JLF-1) weld joint by simulated heat-treatments

    International Nuclear Information System (INIS)

    Inoue, N.; Muroga, T.; Nishimura, A.; Nagasaka, T.; Motojima, O.; Uchida, S.; Yabe, H.; Oguri, K.; Nishi, Y.; Katoh, Y.; Kohyama, A.

    2000-01-01

    Characterization of a weld joint of a Fe-Cr-W ferritic steel (JLF-1) has been carried out in comparison with heat-treated specimens. The heat-treatment was carried out to simulate heating history effects of the base metal (BM), the heat-affected zone (HAZ) and the weld metal (WM) of the joint. Change in X-ray diffraction patterns and hardness of the weld joint and the heat-treated samples are compared and discussed. The results of X-ray diffractometry and the hardness measurements suggest that phase transformation should occur around the heat-treatment temperature of 820-830 deg. C, and that the transformation does not necessarily cause hardening. Although the hardness of the HAZ changes with the distance from fusion line, the internal strain and the residual stress do not change significantly throughout the HAZ. The single heat-treatment test seems insufficient to correlate directly to the HAZ of the weld joint, because repeated heating with different maximum temperatures and different cooling rates would have been applied to the HAZ

  16. 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)

  17. 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)

  18. Radiation induced phosphorus segregation in austenitic and ferritic alloys

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  19. Tensile Properties of Medium Mn Steel with a Bimodal UFG α + γ and Coarse δ-Ferrite Microstructure

    Science.gov (United States)

    Lee, Seonjong; Shin, Sunmi; Kwon, Minhyeok; Lee, Kyooyoung; De Cooman, Bruno C.

    2017-04-01

    While the tensile strength and elongation obtained for medium Mn steel would appear to make it a candidate material in applications which require formable ultra-high strength materials, many secondary aspects of the microstructure-properties relationships have not yet been given enough attention. In this contribution, the microstructural and tensile properties of medium Mn steel with a bimodal microstructure consisting of an ultra-fine grained ferrite + austenite constituent and coarse-grained delta-ferrite are therefore reviewed in detail. The tensile properties of ultra-fine-grained intercritically annealed medium Mn steel reveal a complex dependence on the intercritical annealing temperature. This dependence is related to the influence of the intercritical annealing temperature on the activation of the plasticity-enhancing mechanisms in the microstructure. The kinetics of deformation twinning and strain-induced transformation in the ultra-fine grained austenite play a prominent role in determining the strain hardening of medium Mn steel. While excellent strength-ductility combinations are obtained when deformation twinning and strain-induced transformation occur gradually and in sequence, large elongations are also observed when strain-induced transformation plasticity is not activated. In addition, the localization of plastic flow is observed to occur in samples after intercritical annealing at intermediate temperatures, suggesting that both strain hardening and strain rate sensitivity are influenced by the properties of the ultra-fine-grained austenite.

  20. Heat input effect on the microstructural transformation and mechanical properties in GTAW welds of a 409L ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Delgado, J. A.; Ambriz, R. R.; Cuenca-Alvarez, R.; Alatorre, N.; Curiel, F. F.

    2016-10-01

    Welds without filler metal and welds using a conventional austenitic stainless steel filler metal (ER308L) were performed to join a ferritic stainless steel with Gas Tungsten Arc Welding process (GTAW). Welding parameters were adjusted to obtain three different heat input values. Microstructure reveals the presence of coarse ferritic matrix and martensite laths in the Heat Affected Zone (HAZ). Dilution between filler and base metal was correlated with the presence of austenite, martensite and ferrite in the weld metal. Weld thermal cycles were measured to correlate the microstructural transformation in the HAZ. Microhardness measurements (maps and profiles) allow to identify the different zones of the welded joints (weld metal, HAZ, and base metal). Comparing the base metal with the weld metal and the HAZ, a hardness increment (∼172 HV{sub 0}.5 to ∼350 HV{sub 0}.5 and ∼310 HV{sub 0}.5, respectively) was observed, which has been attributed to the martensite formation. Tensile strength of the welded joints without filler metal increased moderately with respect to base metal. In contrast, ductility was approximately 25% higher than base metal, which provided a toughness improvement of the welded joints. (Author)

  1. Characteristics and Modification of Non-metallic Inclusions in Titanium-Stabilized AISI 409 Ferritic Stainless Steel

    Science.gov (United States)

    Kruger, Dirk; Garbers-Craig, Andrie

    2017-06-01

    This study describes an investigation into the improvement of castability, final surface quality and formability of titanium-stabilized AISI 409 ferritic stainless steel on an industrial scale. Non-metallic inclusions found in this industrially produced stainless steel were first characterized using SEM-EDS analyses through the INCA-Steel software platform. Inclusions were found to consist of a MgO·Al2O3 spinel core, which acted as heterogeneous nucleation site for titanium solubility products. Plant-scale experiments were conducted to either prevent the formation of spinel, or to modify it by calcium treatment. Modification to spherical dual-phase spinel-liquid matrix inclusions was achieved with calcium addition, which eliminated submerged entry nozzle clogging for this grade. Complete modification to homogeneous liquid calcium aluminates was achieved at high levels of dissolved aluminum. A mechanism was suggested to explain the extent of modification achieved.

  2. Liquid metal embrittlement of an austenitic 316L type and a ferritic martensitic T91 type steel by mercury

    Science.gov (United States)

    Medina-Almazán, L.; Auger, T.; Gorse, D.

    2008-06-01

    The susceptibility to liquid metal embrittlement (LME) of 316L and T91 steels by mercury has been studied at room temperature. A dedicated experimental device using center crack tension (CCT) specimens was built. We developed a specimen preparation procedure that must be rigorously applied in order to investigate the embrittling effect of Hg. The high strength ferritic-martensitic steel of type T91 is embrittled by Hg at room temperature over a large range of crosshead speeds, between 6.67 × 10 -7 and 6.67 × 10 -3 m s -1. More surprisingly, the austenitic steel of type 316L is also embrittled by Hg between 1.67 × 10 -8 and 2.5 × 10 -4 m s -1. The fracture of the T91 and 316L CCT specimens in contact with Hg occurs by shear band decohesion over the above-mentioned range of crosshead speeds.

  3. Chemical coloring on stainless steel by ultrasonic irradiation.

    Science.gov (United States)

    Cheng, Zuohui; Xue, Yongqiang; Ju, Hongbin

    2018-01-01

    To solve the problems of high temperature and non-uniformity of coloring on stainless steel, a new chemical coloring process, applying ultrasonic irradiation to the traditional chemical coloring process, was developed in this paper. The effects of ultrasonic frequency and power density (sound intensity) on chemical coloring on stainless steel were studied. The uniformity of morphology and colors was observed with the help of polarizing microscope and scanning electron microscopy (SEM), and the surface compositions were characterized by X-ray photoelectric spectroscopy (XPS), meanwhile, the wear resistance and the corrosion resistance were investigated, and the effect mechanism of ultrasonic irradiation on chemical coloring was discussed. These results show that in the process of chemical coloring on stainless steel by ultrasonic irradiation, the film composition is the same as the traditional chemical coloring, and this method can significantly enhance the uniformity, the wear and corrosion resistances of the color film and accelerate the coloring rate which makes the coloring temperature reduced to 40°C. The effects of ultrasonic irradiation on the chemical coloring can be attributed to the coloring rate accelerated and the coloring temperature reduced by thermal-effect, the uniformity of coloring film improved by dispersion-effect, and the wear and corrosion resistances of coloring film enhanced by cavitation-effect. Ultrasonic irradiation not only has an extensive application prospect for chemical coloring on stainless steel but also provides an valuable reference for other chemical coloring. Copyright © 2017 Elsevier B.V. All rights reserved.

  4. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40 %) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2–3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  5. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    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); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  6. Modeling the constitutive behavior of RAFM steels under irradiation conditions

    Science.gov (United States)

    Aktaa, J.; Petersen, C.

    2011-10-01

    A coupled viscoplastic deformation damage model will be presented which is modified to take into account irradiation induced hardening and its recovery due to inelastic deformation and/or high temperature annealing. The model allows the prediction of the constitutive behavior of RAFM steels under arbitrary creep-fatigue and irradiation loading conditions. It can be implemented in commercial finite element codes and thus be used for the lifetime assessment of fusion reactor components. The model is applied to describe the behavior of the RAFM steels, EUROFER 97 and F82H mod, observed in post irradiation examinations of the irradiation programs ARBOR I and ARBOR II. Data from their tensile and low cycle fatigue tests were used to determine the material and temperature dependent parameters of the model and to verify its prediction capability.

  7. Some considerations on the toughness properties of ferritic stainless steels - A brief review

    CSIR Research Space (South Africa)

    Van Zwieten, ACTM

    1993-02-01

    Full Text Available particles on the toughness aspects. Generally the presence of second phases such as carbides, nitrides and oxides, as well as the chromium-rich ferrite, precipitates and sigma-phase, sigma, can cause a significant decrease in the toughness of ferritic...

  8. Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

    International Nuclear Information System (INIS)

    Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

    2014-01-01

    Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H + ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix–Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix–Gao plot. The bulk hardness of the irradiated region, H 0 , estimated by the Nix–Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H 0 . Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials

  9. 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.

  10. Microstructural study of high irradiated reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, Vladimir; Petriska, Martin; Sojak, Stanislav; Veternikova, Jana [Slovak University of Technology, FEI, Bratislava (Slovakia); Krsjak, Vladimir [Institute for Energy, Joint Research Centre of the European Commission, Petten (Netherlands)

    2009-11-15

    Positron Annihilation Spectroscopy (PAS) techniques in combination with other techniques were effectively used in the testing and selection process of optimal reactor steels for use in Generation III and IV reactors or thermonuclear fusion facilities. Conventional PAS lifetime technique and pulsed low energy positron system were applied on wide spectrum of reactor steels together with other techniques viz., Transmission Electron Microscopy and Moessbauer Spectroscopy focused on the role of Nickel in the steel microstructure. Experimental experiences in this area collected over the last twenty years were very useful in the actual study by avoiding many mistakes in handling with specimens or in careful interpretation of the results. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  11. Effects of processing optimisation on microstructure, texture, grain boundary and mechanical properties of Fe–17Cr ferritic stainless steel thick plates

    Energy Technology Data Exchange (ETDEWEB)

    Han, Jian, E-mail: jh595@uowmail.edu.au [School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, Wollongong, NSW 2522 (Australia); Li, Huijun; Zhu, Zhixiong [School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, Wollongong, NSW 2522 (Australia); Jiang, Laizhu; Xu, Haigang; Ma, Li [Baoshan Iron and Steel Co., Ltd., Shanghai 200431 (China)

    2014-10-20

    The relationships between microstructure, texture, grain boundary and tensile strength, Charpy impact toughness of (Nb+Ti+V) stabilised Fe–17Cr ferritic stainless steel thick plates were investigated by means of optical microscopy, X-ray diffraction, scanning electron microscopy, electron backscatter diffraction, tensile and Charpy impact testing. The results show that for Fe–17Cr ferritic stainless steel thick plate, the addition of warm rolling procedure leads to refinement of grain size, modification of texture, and then optimisation of grain boundary, including grain boundary character distribution and grain boundary connectivity. Meanwhile, the mechanical testing results indicate that optimal transformation that warm rolling procedure brings to Fe–17Cr ferritic steel thick plate is beneficial to its mechanical properties.

  12. Annealing of a ferritic stainless steel 409 stabilized with titanium and zirconium additions

    Directory of Open Access Journals (Sweden)

    Zambrano, P.

    2011-02-01

    Full Text Available A ferritic stainless steel 409 stabilized with titanium and zirconium was subject to thermomechanical processing. It was heated at 1210 °C for one hour, followed by a 75 % hot reduction in three passes, this rolling schedule ended at 980 °C. Samples were cooled to 600 °C by water spraying followed by air-cooling. The alloy was pickled, and was reduced 80 % by cold rolling. The alloy was annealed at different temperatures for 105 s. Additional annealing treatments were carried out at temperatures of 800, 850 and 900 °C for different times. Mechanical testing and texture were made to corroborate the degree of annealing and formability. Mechanical properties and Texture analyses showed that the alloy annealed at 850 °C for 14 min was both completely recrystallized and a very good formability.

    Un acero inoxidable ferrítico 409 estabilizado con titanio y zirconio fue sujeto a procesos termomecánicos. El acero fue calentado a 1210 ºC durante una hora, seguido por un laminado en caliente del 75 % en tres pases, el proceso terminó a los 980 ºC. Las muestras fueron enfriadas hasta 600 ºC por agua atomizada seguido de enfriamiento al aire. La aleación fue decapada y laminada en frío un 80 %. Posteriormente de desarrollaron tratamientos térmicos de recocido a diferentes temperaturas por un tiempo de 105 s. Adicionalmente se desarrollaron tratamientos de recocido a temperaturas de 800, 850 y 900 ºC a diferentes tiempos. Pruebas mecánicas y textura fueron realizadas para corroborar el grado de recocido y su formalidad. El análisis de las propiedades mecánicas y la Textura mostraron que la aleación recocida a 850 ºC por 14 min (840 s fue completamente recristalizada obteniendo la mejor formabilidad.

  13. Toughness measurements of tungsten coated ferritic steels using laser induced stress pulses

    International Nuclear Information System (INIS)

    El-Awady, J.; Gupta, V.; Kim, B.; Ghoniem, N.; Sharafat, S.

    2007-01-01

    Full text of publication follows: Tungsten is a primary candidate for armor material protecting low activation ferritic steel in plasma facing components. The tungsten coatings are applied by HIPing or vacuum plasma spraying (VPS). To facilitate high helium recycling of implanted helium from the armor surface, a high porosity (10% - 30%) VPS Tungsten coating consisting of nano-sized particles was produced. Because, these pores can act as crack nucleation sites, the resistance of the coating to failure is an important factor that needs to be quantified. The failure strength of coating is typically measured by pulling on the coatings or bending the samples until failure. Such techniques introduce a significant number of uncertainties regarding the accuracy of the resultant coating strength. One of the major obstacles in such techniques is the difficulty in measuring the intrinsic mechanical properties independently form the extrinsic effects arising from material inelasticity, specimen geometry and loading configuration. To avoid such extrinsic effects we use the Laser Spallation Technique (LST) to relate the local energy release rate (i.e. coating toughness) to the coating's free surface velocity following a nano-second laser induced compression/tension stress wave in the samples. The propagation of the tension wave results in the dynamic failure of the weakest link in the coating itself or bond interface. This technique produces high strain rate loadings (10 7 sec -1 ) that will suppress all inelastic deformation accompanying the crack initiation at pores sites, thus yielding a coating toughness value representative of the intrinsic interfacial energy. This coating toughness is then used to evaluate the true failure strength of the coating through numerical analysis based on the true geometry and true loading configuration in a typical fusion reactor environment. (authors)

  14. 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.

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

    Science.gov (United States)

    Moon, Joonoh; Lee, Chang-Hoon; Lee, Tae-Ho; Jang, Min-Ho; Park, Min-Gu; Han, Heung Nam

    2014-12-01

    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.

  16. Deformation modes of proton and neutron irradiated stainless steels

    Science.gov (United States)

    Bailat, C.; Gröschel, F.; Victoria, M.

    2000-01-01

    AISI 304 and 316 stainless steels of two purity levels that have been irradiated with high energy protons up to 0.3 dpa and neutrons in a high flux reactor up to 7.5 dpa were investigated in terms of irradiation induced mechanical properties and microstructural changes. The stress-strain relationships were obtained at room temperature. The deformation, grain, twinning and irradiation defect microstructures were investigated using both transmission and scanning electron microscopy. The results are discussed in terms of deformation mechanisms linked with the radiation induced defect microstructure.

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

    Science.gov (United States)

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

    1994-03-08

    This work describes a high strength, high toughness bainitic/ferritic steel alloy comprising about 2.75% to 4.0% chromium, about 2.0% to 3.5% tungsten, about 0.10% to 0.30% vanadium, and about 0.1% to 0.15% carbon with the balance iron, wherein the percentages are by total weight of the composition, wherein the alloy having been heated to an austenitizing temperature and then cooled at a rate sufficient to produce carbide-free acicular bainite. 15 figures.

  18. The Studies of Irradiation Hardening of Stainless Steel Reactor Internals under Proton and Xenon Irradiation

    Directory of Open Access Journals (Sweden)

    Chaoliang Xu

    2016-06-01

    Full Text Available Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa, whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

  19. Resistance to abrasive wear and mechanical properties of ferritic-martensitic and ferritic-austenitic structures of the steels C 22, C 45, and X 2 CrNiMoN 22 5 3

    International Nuclear Information System (INIS)

    Rosenheinrich, M.

    1989-01-01

    Two-phase, ferritic-martensitic structures were produced with the help of the unalloyed carbon steels C 22 and C 45, and ferritic-austenitic structures were produced by thermal and/or thermo-mechanical treatment with the help of the high-alloy steel X2CrNiMoN 22 5 3. The phase portion, distribution, form and size were varied. The structures were analyzed quantitatively and described with the help of structure factors such as duplex and dispersion parameters. The mechanical characteristics of structures in tensile, impact and rolling tests were studied. The abrasive wear resistance of the structures against flint was determined in a tribological system in accordance with the abrasive-paper method. The mechanical characteristics were influenced by the structure factors. An optimum of wear resistance was exhibited by ferritic-martensitic structures with an island-shaped coarse and hard martensite stored in a ferritic matrix. The abrasive wear resistance of the ferrite-austenite structure increased as the percentage of austenite increased, although the structure hardness decreased. (orig.) With 62 figs., 28 tabs [de

  20. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

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

  1. Damage rate and spectrum effects in ferritic steel ΔNDTT data

    International Nuclear Information System (INIS)

    Simons, R.L.

    1986-09-01

    A model was developed for irradiation-induced hardening in pressure vessel steels by cascade-induced nucleation and free defect diffusion of copper to coherent copper clusters. The model was used as a functional framework for fitting data on ΔNDTT as a function of damage dose and dose rate. Several defect cross sections having different neutron energy dependences were tried. The cross section for Frenkel pair production, based on resistivity measurements at 4 K, was found to give a comparable or slightly better fit to the data than dpa. Damage rate is an important parameter to consider in correlation of ΔNDTT data and damage rate can affect the nucleation as well as the growth of the copper precipitation

  2. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Onizawa, K. [Japan Atomic Energy Agency, 2-4 Shirakata-Shirane, Tokai, Naka-gun, Ibaraki 319-1195 (Japan); Suzuki, M. [Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki-gun, Ibaraki 311-1393 (Japan)

    2013-11-15

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 10{sup 23} n m{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 17} n m{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ′-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  3. Overload effects on a ferritic-baintic steel and a cast aluminium alloy: two very different behaviours

    Energy Technology Data Exchange (ETDEWEB)

    Saintier, N. [Arts et Metiers Paris Tech, I2M, UMR CNRS, Universite Bordeaux 1, Talene Cedex (France); El Dsoki, C.; Kaufmann, H.; Sonsino, C.M. [Fraunhofer-Institute for Structural Durability and System Reliability LBF, Darmstadt (Germany); Dumas, C. [RENAULT, Technocentre, Guyancourt Cedex (France); Voellmecke, F.J. [BORBET GmbH, Hallenberg-Hesborn (Germany); Palin-Luc, T.; Bidonard, H.

    2011-10-15

    Load controlled fatigue tests were performed up to 10{sup 7} cycles on flat notched specimens (K{sub t} = 2.5) under constant amplitude and variable amplitude loadings with and without periodical overloads. Two materials are studied: a ferritic-bainitic steel (HE400M steel) and a cast aluminium alloy (AlSi7Mg0.3). These materials have a very different cyclic behaviour: the steel exhibits cyclic strain softening whereas the Al alloy shows cyclic strain hardening. The fatigue tests show that, for the steel, periodical overload applications reduce significantly the fatigue life for fully reversed load ratio (R{sub {sigma}} = -1), while they have no influence under pulsating loading (R{sub {sigma}} = 0). For the Al alloy overloads have an effect (fatigue life decreasing) only for variable amplitude loadings. The detrimental effect of overloads on the steel is due to ratcheting at the notch root which evolution is overload's dependent. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  4. Impact Strength of Austenitic and Ferritic-Austenitic Cr-Ni Stainless Cast Steel in -40 and +20°C Temperature

    Directory of Open Access Journals (Sweden)

    Kalandyk B.

    2014-10-01

    Full Text Available Studies described in this paper relate to common grades of cast corrosion resistant Cr-Ni steel with different matrix. The test materials were subjected to heat treatment, which consisted in the solution annealing at 1060°C followed by cooling in water. The conducted investigations, besides the microstructural characteristics of selected cast steel grades, included the evaluation of hardness, toughness (at a temperature of -40 and +20oC and type of fracture obtained after breaking the specimens on a Charpy impact testing machine. Based on the results of the measured volume fraction of ferrite, it has been found that the content of this phase in cast austenitic steel is 1.9%, while in the two-phase ferritic-austenitic grades it ranges from 50 to 58%. It has been demonstrated that within the scope of conducted studies, the cast steel of an austenitic structure is characterised by higher impact strength than the two-phase ferritic-austenitic (F-A grade. The changing appearance of the fractures of the specimens reflected the impact strength values obtained in the tested materials. Fractures of the cast austenitic Cr-Ni steel obtained in these studies were of a ductile character, while fractures of the cast ferritic-austenitic grade were mostly of a mixed character with the predominance of brittle phase and well visible cleavage planes.

  5. The tensile and fatigue properties of type 1.4914 ferritic steel for fusion reactor applications

    International Nuclear Information System (INIS)

    Marmy, P.; Victoria, M.; Ruan, Y.

    1989-08-01

    Martensitic steels have received considerable attention as structural materials in fusion reactor applications. In present designs, fusion reactors are expected to operate in a cyclic mode, thus producing cyclic thermal stresses in the first wall. Due to its thermal expansion coefficient and very low swelling rate, 1.4914 martensitic steel is a suitable candidate for the first wall with high neutron loadings. This paper presents the preirradiation results obtained with subsize-specimens designed to be irradiated with a proton beam in the PIREX facility at the Paul Scherrer Institute (PSI) of Wuerenlingen. Both tensile and low cycle fatigue tests were performed in vacuum in the region from 300 K to 870 K (720 K in the case of fatigue tests). Tensile tests on the subsize specimens (0.33 mm thick) compared well to those on bulk specimens, showing a minimum in ductility at around 620 K. The fatigue tests, performed on tubular specimens (3.4 mm external diameter, 0.35 mm wall thickness) showed substantial softening setting in at a low number of cycles. The initial microstructure observed in transmission microscopy consists of fine martensite laths. As cyclic deformation proceeds, dislocation cells form, that gradually replace the martensitic laths. (author) 19 figs., 5 tabs., 16 refs

  6. Standard test method for conducting drop-weight test to determine nil-ductility transition temperature of ferritic steels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2006-01-01

    1.1 This test method covers the determination of the nil-ductility transition (NDT) temperature of ferritic steels, 5/8 in. (15.9 mm) and thicker. 1.2 This test method may be used whenever the inquiry, contract, order, or specification states that the steels are subject to fracture toughness requirements as determined by the drop-weight test. 1.3 The values stated in inch-pound units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  7. Measurement of internal stresses in a welded assembly made of 20 MND 53 ferritic steel. Stress relaxation

    International Nuclear Information System (INIS)

    Baron, J.L.

    1981-05-01

    An assembly consisting of two flanged plates welded end to end was realized, the plates being made of 20 MND 53 ferritic steel. This grade of steel is used to fabricate PWR reactor vessels. At each important stage in the realization of the assembly, internal stresses were measured by X-ray diffraction techniques. These measurements showed that the finished realization manifested compression stresses in the main parts of its structure and large amplitude tensile stresses in the central weld seam. A heat treatment was performed (615 0 C - 17 hours). Stresses levels almost every where in the assembly dropped to practically zero. Although stress relaxation kinetics are probably slower in voluminous parts, this result and most of the results published in literature on this subject, indicate that such treatment notably diminishes internal welding stress even in thick parts [fr

  8. Final report for the year 2001 on experimental and theoretical investigations of irradiation effects on physical and mechanical properties of iron and RAFM steels

    DEFF Research Database (Denmark)

    Singh, Bachu Narain

    2003-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties have been studied both experimentally and theoretically. Specimens of pure iron and RAFM (reduced activation ferritic-martensic) steels were irradiated todifferent dose levels and at different irradiation...... temperatures. The resulting microstructure was characterized using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. Mechanical properties weredetermined by uniaxial tensile testing. Dislocation-loop interaction, formation of rafts of loops, radiation...... hardening and formation of “cleared channels” were studied using different computational techniques. Experiments have shown that nano-voids areformed both in pure iron and F82H steel already at 50°C. In pure iron, the formation of nano-voids is detected already at a dose level of ~10-3 dpa. Also in iron...

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

    Science.gov (United States)

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

    2017-06-03

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

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

    Directory of Open Access Journals (Sweden)

    Suvi Papula

    2017-06-01

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

  11. Factors Affecting Impact Toughness in Stabilized Intermediate Purity 21Cr Ferritic Stainless Steels and Their Simulated Heat-Affected Zones

    Science.gov (United States)

    Anttila, Severi; Alatarvas, Tuomas; Porter, David A.

    2017-12-01

    The correlation between simulated weld heat-affected zone microstructures and toughness parameters has been investigated in four intermediate purity 21Cr ferritic stainless steels stabilized with titanium and niobium either separately or in combination. Extensive Charpy V impact toughness testing was carried out followed by metallography including particle analysis using electron microscopy. The results confirmed that the grain size and the number density of particle clusters rich in titanium nitride and carbide with an equivalent circular diameter of 2 µm or more are statistically the most critical factors influencing the ductile-to-brittle transition temperature. Other inclusions and particle clusters, as well as grain boundary precipitates, are shown to be relatively harmless. Stabilization with niobium avoids large titanium-rich inclusions and also suppresses excessive grain growth in the heat-affected zone when reasonable heat inputs are used. Thus, in order to maximize the limited heat-affected zone impact toughness of 21Cr ferritic stainless steels containing 380 to 450 mass ppm of interstitials, the stabilization should be either titanium free or the levels of titanium and nitrogen should be moderated.

  12. Large Scale Screening of Low Cost Ferritic Steel Designs For Advanced Ultra Supercritical Boiler Using First Principles Methods

    Energy Technology Data Exchange (ETDEWEB)

    Ouyang, Lizhi [Tennessee State Univ. Nashville, TN (United States)

    2016-11-29

    Advanced Ultra Supercritical Boiler (AUSC) requires materials that can operate in corrosive environment at temperature and pressure as high as 760°C (or 1400°F) and 5000psi, respectively, while at the same time maintain good ductility at low temperature. We develop automated simulation software tools to enable fast large scale screening studies of candidate designs. While direct evaluation of creep rupture strength and ductility are currently not feasible, properties such as energy, elastic constants, surface energy, interface energy, and stack fault energy can be used to assess their relative ductility and creeping strength. We implemented software to automate the complex calculations to minimize human inputs in the tedious screening studies which involve model structures generation, settings for first principles calculations, results analysis and reporting. The software developed in the project and library of computed mechanical properties of phases found in ferritic steels, many are complex solid solutions estimated for the first time, will certainly help the development of low cost ferritic steel for AUSC.

  13. Effect of ferrite on the precipitation of σ phase in cast austenitic stainless steel used for primary coolant pipes of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yongqiang; Li, Na, E-mail: wangyongqiang1124@163.com [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology, Beijing (China)

    2017-11-15

    The effect of ferrite phase on the precipitation of σ phase in a Z3CN20.09M cast austenitic stainless steel (CASS) used for primary coolant pipes of pressurized water reactor (PWR) nuclear power plants was investigated by using isothermal heat-treatment, optical microscopy (OM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) techniques. The influence of different morphologies and volume fractions of ferrite in the σ phase formation mechanism was discussed. The amount of σ phase precipitated in all specimens with different microstructures increased with increasing of aging time, however, the precipitation rate is significant different. The formation of σ phase in specimens with the coarsest ferrite and the dispersively smallest ferrite is slowest. The lowest level Cr content in ferrite and fewest α/γ interfaces in specimen are the main reasons for the slowest σ precipitation due to they are unfavorable for the kinetics and thermodynamics of phase transformation respectively. By contraries, the fastest formation of σ phase takes place in specimens with narrow and long ferrite due to the most α/γ interfaces and higher Cr content in ferrite which are beneficial for preferential nucleation and formation thermodynamics of σ phase. (author)

  14. 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

  15. Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

  16. Studies on oxidation and deuterium permeation behavior of a low temperature α-Al{sub 2}O{sub 3}-forming Fe−Cr−Al ferritic steel

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yu-Ping [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Zhao, Si-Xiang [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Liu, Feng; Li, Xiao-Chun; Zhao, Ming-Zhong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Wang, Jing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Lu, Tao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Hong, Suk-Ho [National Fusion Research Institute, 169-148 Gwahangno, Yusung-Gu, Daejeon, 305-333 (Korea, Republic of); Zhou, Hai-Shan, E-mail: haishanzhou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Luo, Guang-Nan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei, 230031 (China); Hefei Center for Physical Science and Technology, Hefei, 230031 (China); Hefei Science Center of Chinese Academy of Science, Hefei, 230027 (China)

    2016-08-15

    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{sub 2}O{sub 3} was detected in the oxide film by XPS, and HRTEM showed that Al{sub 2}O{sub 3} in the α phase was found. The formation of α-Al{sub 2}O{sub 3} layer at a relatively low temperature may result from the formation of Cr{sub 2}O{sub 3} nuclei.

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

    International Nuclear Information System (INIS)

    Dalton, J.H.

    1978-01-01

    The paper reviews the effects of fast-neutron irradiation on the tensile properties of austenitic stainless steels at irradiation temperatures of less than 400 degrees Celcius, using as an example, work carried out at Pelindaba on an AISI 316 type steel produced in South Africa. Damage produced in these steels at higher irradiation temperatures and fluences is also briefly discussed. The paper concludes with a discussion of some methods of overcoming or decreasing the effects of irradiation damage [af

  18. Irradiation hardening of Mod.9Cr-1Mo steel

    International Nuclear Information System (INIS)

    Ryu, Woo-Seog; Kim, Sung-Ho; Choo, Kee-Nam; Kim, Do-Sik

    2009-01-01

    An irradiation test of Mod.9Cr-1Mo steel was carried out in the OR5 test hole of HANARO of a 30 MW thermal power at 390±10degC up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E > 1.0 MeV). The dpa of the irradiated specimens was evaluated to be 0.034 - 0.07. Tensile and impact tests of the irradiated Mod.9Cr-1Mo were done in the hot cell of the IMEF. The change of the tensile strength by irradiation was similar to the change of the yield strength. The increase of the yield and tensile strengths was up to 18% and 10% respectively. The elongation reduction of the weldment was up to 65%. (author)

  19. Oxidation behavior of ferritic-martensitic and ODS steels in supercritical water

    Science.gov (United States)

    Bischoff, Jeremy

    water corroded much faster than those in steam (1.5 to 2 times faster). Additionally, during these corrosion tests a marker experiment was performed with the deposition of micrometric palladium markers on the surface of some samples prior to oxidation. The markers were found at the outer-inner layer interface, consistent with a corrosion mechanism of outward migration of iron to form the outer layer and inward migration of oxygen to form the inner layer. The discrepancy between the SCW and steam environments suggests that the outward migration of iron may be the rate-limiting step. A detailed study of the oxide advancement was performed using the TEM by analyzing the inner and diffusion layer structure. Energy-filtered TEM images were acquired to analyze the micrometric and nanometric distribution of elements in these layers. Such images from the inner layer revealed the presence of localized chromium enrichment regions associated with the presence of pores. Additionally, an iron-chromium nanometric segregation was observed and may be associated with the mixture of Fe3O4 and FeCr2O4. In the diffusion layer, small nanometric chromium-rich oxide particles were seen within metal grains. The (Fe,Cr)3O4 spinel oxide has an inverse spinel structure as Fe3O4 but becomes normal spinel as FeCr 2O4, thus the structure changes depending on the chromium content. Additionally, the spinel structure was analyzed using the ligand theory and showed that chromium does not migrate and that the main diffusing species is the Fe2+ ion. Calculations of the amount of iron leaving the inner layer showed that this amount accounted for the amount of iron necessary to form the outer layer, thus no dissolution of oxide in SCW is observed. Additionally, the differences in oxidation behavior in steam and SCW suggest that the rate-limiting step for the corrosion of ferritic-martensitic steels is the iron outward migration. The iron migration is driven by the gradient in the Fe2+/Fe 3+ ratio and is

  20. Investigation of microstructure and mechanical properties of low dose neutron irradiated HT-9 steel

    International Nuclear Information System (INIS)

    Sarkar, A.; Alsabbagh, A.H.; Murty, K.L.

    2014-01-01

    Highlights: • Neutron irradiation has been carried out on HT-9 steel. • Microstructure of the irradiated HT-9 steel has been investigated using XRD. • There is an increase in dislocation density in the irradiated sample. • Tensile tests have been carried out to determine the changes in mechanical properties due to irradiation. • Yield stress and strain rate sensitivity increased due to irradiation. - Abstract: HT-9 steel samples have been irradiated with fast neutrons (E > 0.1 MeV) to a low dose (1.2 × 10 −3 dpa). Microstructure of the unirradiated and irradiated samples has been characterized by X-ray diffraction line profile analysis using different model-based approaches. The domain size and density of dislocations of the irradiated steel have been estimated. Different types of tensile tests have been carried out at room temperature to assess the changes in mechanical properties of HT-9 steel due to neutron irradiation

  1. Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under gamma-rays irradiation

    International Nuclear Information System (INIS)

    Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

    2014-01-01

    Seawater was injected into reactor cores of Units 1, 2, and 3 in the Fukushima Daiichi nuclear power station as an urgent coolant. It is considered that the injected seawater causes corrosion of steels of the reactor pressure vessel and primary containment vessel. To investigate the effects of gamma-rays irradiation on weight loss in carbon steel and low-alloy steel, corrosion tests were performed in diluted seawater at 50°C under gamma-rays irradiation. Specimens were irradiated with dose rates of 4.4 kGy/h and 0.2 kGy/h. To evaluate the effects of hydrazine (N 2 H 4 ) on the reduction of oxygen and hydrogen peroxide, N 2 H 4 was added to the diluted seawater. In the diluted seawater without N 2 H 4 , weight loss in the steels irradiated with 0.2 kGy/h was similar to that in the unirradiated steels, and weight loss in the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those in the unirradiated steels. Weight loss in the steels irradiated in the diluted seawater containing N 2 H 4 was similar to that in the diluted seawater without N 2 H 4 . When N 2 was introduced into the gas phase in the flasks during gamma-rays irradiation, weight loss in the steels decreased. (author)

  2. Microchemical evolution of neutron-irradiated stainless steel

    International Nuclear Information System (INIS)

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

    1980-04-01

    The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations

  3. Effects of Mn and Al on the Intragranular Acicular Ferrite Formation in Rare Earth Treated C-Mn Steel

    Science.gov (United States)

    Song, Mingming; Song, Bo; Yang, Zhanbing; Zhang, Shenghua; Hu, Chunlin

    2017-07-01

    The influence of Al, Mn and rare earth (RE) on microstructure of C-Mn steel was investigated. The capacities of different RE inclusions to induce intragranular acicular ferrite (AF) formation were compared. Result shows that RE treatment could make C-Mn steel from large amounts of intragranular AF. Al killed is detrimental to the formation of intragranular AF in RE-treated C-Mn steel. An upper bainite structure would replace the AF when Al content increased to 0.027 mass %. The optimal Mn content to form AF is about 0.75-1.31 mass %. The effective RE inclusion which could induce AF nucleation is La2O2S. When patches of MnS are attached on the surface of La2O2S inclusion, AF nucleation capacity of RE-containing inclusion could enlarge obviously. The existence of manganese-depleted zone and low lattice misfit would be the main reason of La-containing inclusion inducing AF nucleation in C-Mn steel.

  4. 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

  5. Phase identification in neutron irradiated stainless steels

    International Nuclear Information System (INIS)

    Lee, E.H.; Rowcliffe, A.F.

    1980-01-01

    Techniques used for the identification of phases which develop in AISI 316 stainless steel which has been modified by the addition of Ti are described. Five major phases were identified in the alloy containing 0.2 wt % Ti after irradation by 7 x 10 22 n.cm -2 at temperatures ranging from 400 to 650 0 C. Once identification was established from diffraction pattern measurements, subsequent identification could be made by observation of the characteristic shape of each phase combined with the observation of certain characteristic features of te x-ray spectrum of each phase. This combination permitted rapid identification of large numbers of particles necessary for the elucidation of the role of phase instabilities in void swelling

  6. In-line x-ray phase-contrast tomography and diffraction-contrast tomography study of the ferrite-cementite microstructure in steel

    NARCIS (Netherlands)

    Kostenko, A.; Sharma, H.; Gözde Dere, E.; King, A.; Ludwig, W.; Van Oel, W.; Offerman, S.E.; Stallinga, S.; Van Vliet, L.J.

    2011-01-01

    This work presents the development of a non-destructive imaging technique for the investigation of the microstructure of cementite grains embedded in a ferrite matrix of medium-carbon steel. The measurements were carried out at the material science beamline of the European Synchrotron Radiation

  7. Investigation of microstructural changes in ferritic stainless steels caused by high-temperature deformation. Technical progress report, July 1, 1982-July 1, 1983

    International Nuclear Information System (INIS)

    Weertman, J.R.

    1983-07-01

    The research effort of this grant has been directed into two areas: a study of the high temperature mechanical behavior of a ferritic stainless steel, Fe9CrlMo modified by the addition of small amounts of V and Nb, and an investigation by small angle neutron scattering of changes in its microstructure produced by service at elevated temperature

  8. Recommendations for protecting against failure by brittle fracture: Category II and III ferritic steel shipping containers with wall thickness greater than four inches

    International Nuclear Information System (INIS)

    Schwartz, M.W.; Fischer, L.E.

    1996-08-01

    This report provides criteria for selecting ferritic steels that would prevent brittle fracture in Category II and III shipping containers with wall thickness greater than 4 inches. These methods are extensions of those previously used for Category II and III containers less than 4 inches thick and Category I containers more than 4 inches thick

  9. The Kinetics of Dislocation Loop Formation in Ferritic Alloys Through the Aggregation of Irradiation Induced Defects

    Science.gov (United States)

    Kohnert, Aaron Anthony

    The mechanical properties of materials are often degraded over time by exposure to irradiation environments, a phenomenon that has hindered the development of multiple nuclear reactor design concepts. Such property changes are the result of microstructural changes induced by the collision of high energy particles with the atoms in a material. The lattice defects generated in these recoil events migrate and interact to form extended damage structures. This study has used theoretical models based on the mean field chemical reaction rate theory to analyze the aggregation of isolated lattice defects into larger microstructural features that are responsible for long term property changes, focusing on the development of black dot damage in ferritic iron based alloys. The purpose of such endeavors is two-fold. Primarily, such models explain and quantify the processes through which these microstructures form. Additionally, models provide insight into the behavior and properties of the point defects and defect clusters which drive general microstructural evolution processes. The modeling effort presented in this work has focused on physical fidelity, drawing from a variety of sources of information to characterize the unobservable defect generation and agglomeration processes that give rise to the observable features reported in experimental data. As such, the models are based not solely on isolated point defect creation, as is the case with many older rate theory approaches, but instead on realistic estimates of the defect cluster population produced in high energy cascade damage events. Experimental assessments of the microstructural changes evident in transmission electron microscopy studies provide a means to measure the efficacy of the kinetic models. Using common assumptions of the mobility of defect clusters generated in cascade damage conditions, an unphysically high density of damage features develops at the temperatures of interest with a temperature dependence

  10. Microstructural examination of commercial ferritic alloys at 299 DPA

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1995-11-01

    Microstructures and density change measurements are reported for Martensitic commercial steels HT-9 and Modified 9Cr-lMo (T9) and oxide dispersion strengthened ferritic alloys MA956 and NU957 following irradiation in the FFTF/MOTA at 420 degrees C to 200 DPA. Swelling as determined by density change remains below 2% for all conditions. Microstructures are found to be stable except in recrystallized grains of MA957, which are fabrication artifacts, with only minor swelling in the Martensitic steels and α' precipitation in alloys with 12% or more chromium. These results further demonstrate the high swelling resistance and microstructural stability of the ferritic alloy class

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

    International Nuclear Information System (INIS)

    Vincent, E.

    2006-12-01

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

  12. 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

  13. Neutron metrology in the HFR. Irradiation of Low Activation Steel Specimens R285-04 (ILAS). Evaluation report

    International Nuclear Information System (INIS)

    Ketema, D.J.

    1998-06-01

    ECN is working on the assessment of low temperature irradiation hardening and embrittlement of ferritic/martensitic alloys, developed for fusion application. The irradiation of specimen holder R285-04 (ILAS) in the HFR in Petten, Netherlands, loaded with tensile specimens manufactured from four types of stainless steel alloys, is part of this programme. The R285-04 assembly was irradiated in channel 3 of a TRIO type facility in HFR core-position D2 up to a target dose level of approximately 2.5 dpa (displacements per atom) in stainless steel at a nominal temperature of about 325C. This report presents the final metrology results obtained from activation monitor sets situated in one of the specimen channels inside the specimen holder, including detailed information concerning an estimation of the fluence dose received by each specimen separately and its temperature during irradiation. The total number of displacements per atom (dpa), the generated helium content and the activity values after irradiation for several waiting times are also given for each specimen. Additionally the metrology results were cross-checked with calculations by means of the KENO-Va Monte Carlo code, giving a very good agreement on the centre-line of the specimen holder, and larger deviations on the top and bottom positions of the assembly. The main results of the thermal and fast neutron fluence measurements, indicate that the obtained damage levels in the steel specimens loaded in this specimen holder vary from 2.1 to 3.5 dpa. Detailed data are presented. 28 refs

  14. Simulation of the hot rolling and accelerated cooling of a C-Mn ferrite-bainite strip steel

    Science.gov (United States)

    Debray, B.; Teracher, P.; Jonas, J. J.

    1995-01-01

    By means of torsion testing, the microstructures and mechanical properties produced in a 0.14 Pct C-1.18 Pct Mn steel were investigated over a wide range of hot-rolling conditions, cooling rates, and simulated coiling temperatures. The austenite grain size present before accelerated cooling was varied from 10 to 150 μm by applying strains of 0 to 0.8 at temperatures of 850 °C to 1050 °C. Two cooling rates, 55 °C/s and 90 °C/s, were used. Cooling was interrupted at temperatures ranging from 550 °C to 300 °C. Optical microscopy and transmission electron microscopy (TEM) were employed to investigate the microstructures. The mechanical properties were studied by means of tensile testing. When a fine austenite grain size was present before cooling and a high cooling rate (90 °C/s) was used, the microstructure was composed of ferrite plus bainite and a mixture of ferrite and cementite, which may have formed by an interphase mechanism. The use of a lower cooling rate (55 °C/s) led to the presence of ferrite and fine pearlite. In both cases, the cooling interruption temperature and the amount of prior strain had little influence on the mechanical properties. Reheating at 1050 °C, which led to the presence of very coarse austenite, resulted in a stronger influence of the interruption temperature. A method developed at Institut de Recherche Sidérurgique (IRSID, St. Germain-en-Laye, France) for deducing the Continuous-Cooling-Transformation (CCT) diagrams from the cooling data was adapted to the present apparatus and used successfully to interpret the observed influence of the process parameters.

  15. Gas atomized precursor alloy powder for oxide dispersion strengthened ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Rieken, Joel [Iowa State Univ., Ames, IA (United States)

    2011-12-13

    Gas atomization reaction synthesis (GARS) was employed as a simplified method for producing precursor powders for oxide dispersion strengthened (ODS) ferritic stainless steels (e.g., Fe-Cr-Y-(Ti,Hf)-O), departing from the conventional mechanical alloying (MA) process. During GARS processing a reactive atomization gas (i.e., Ar-O2) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 150 nm) metastable Cr-enriched oxide layer that was used as a vehicle for solid-state transport of O into the consolidated microstructure. In an attempt to better understand the kinetics of this GARS reaction, theoretical cooling curves for the atomized droplets were calculated and used to establish an oxidation model for this process. Subsequent elevated temperature heat treatments, which were derived from Rhines pack measurements using an internal oxidation model, were used to promote thermodynamically driven O exchange reactions between trapped films of the initial Cr-enriched surface oxide and internal Y-enriched intermetallic precipitates. This novel microstructural evolution process resulted in the successful formation of nano-metric Y-enriched dispersoids, as confirmed using high energy X-ray diffraction and transmission electron microscopy (TEM), equivalent to conventional ODS alloys from MA powders. The thermal stability of these Y-enriched dispersoids was evaluated using high temperature (1200°C) annealing treatments ranging from 2.5 to 1,000 hrs of exposure. In a further departure from current ODS practice, replacing Ti with additions of Hf appeared to improve the Y-enriched dispersoid thermal stability by means of crystal structure modification. Additionally, the spatial distribution of the dispersoids was found to depend strongly on the original rapidly solidified microstructure. To exploit this, ODS microstructures were engineered from

  16. Characterization of thick plasma spray tungsten coating on ferritic/martensitic steel F82H for high heat flux armor

    International Nuclear Information System (INIS)

    Yahiro, Y.; Mitsuhara, M.; Tokunakga, K.; Yoshida, N.; Hirai, T.; Ezato, K.; Suzuki, S.; Akiba, M.; Nakashima, H.

    2009-01-01

    Two types of plasma spray tungsten coatings on ferritic/martensitic steel F82H made by vacuum plasma spray technique (VPS) and air plasma spray technique (APS) were examined in this study to evaluate the possibility as plasma-facing armor. The VPS-W/F82H showed superior properties. The porosity of the VPS-W coatings was about 0.6% and most of the pores were smaller than 1-2 μm and joining of W/F82H and W/W was fairly good. Thermal load tests indicated high potential of this coating as plasma-facing armor under thermal loading. In case of APS-W/F82H, however, porosity was 6% and thermal load properties were much worse than VPS-W/F82H. It is likely that surface oxidation during plasma spray process reduced joining properties. Remarkably, both coatings created soft ferrite interlayer after proper heat treatments probably due to high residual stress at the interfaces after the production. This indicates the potential function of the interlayer as stress relieve and possible high performance of such coating component under thermal loads.

  17. Characterization of thick plasma spray tungsten coating on ferritic/martensitic steel F82H for high heat flux armor

    Science.gov (United States)

    Yahiro, Y.; Mitsuhara, M.; Tokunakga, K.; Yoshida, N.; Hirai, T.; Ezato, K.; Suzuki, S.; Akiba, M.; Nakashima, H.

    2009-04-01

    Two types of plasma spray tungsten coatings on ferritic/martensitic steel F82H made by vacuum plasma spray technique (VPS) and air plasma spray technique (APS) were examined in this study to evaluate the possibility as plasma-facing armor. The VPS-W/F82H showed superior properties. The porosity of the VPS-W coatings was about 0.6% and most of the pores were smaller than 1-2 μm and joining of W/F82H and W/W was fairly good. Thermal load tests indicated high potential of this coating as plasma-facing armor under thermal loading. In case of APS-W/F82H, however, porosity was 6% and thermal load properties were much worse than VPS-W/F82H. It is likely that surface oxidation during plasma spray process reduced joining properties. Remarkably, both coatings created soft ferrite interlayer after proper heat treatments probably due to high residual stress at the interfaces after the production. This indicates the potential function of the interlayer as stress relieve and possible high performance of such coating component under thermal loads.

  18. Effect of Normalizing Temperature on Fracture Characteristic of Tensile and Impact Tested Creep Strength-Enhanced Ferritic P92 Steel

    Science.gov (United States)

    Saini, N.; Pandey, C.; Mahapatra, M. M.

    2017-11-01

    The high-temperature Cr-Mo creep strength-enhanced ferritic (CSEF) steels are mainly used in nuclear and thermal power plants. In the present investigation, a systematic study on fracture surface morphologies of tensile and impact tested specimens and mechanical properties of cast and forged (C&F) P92 steel was performed for various heat treatment conditions. The heat treatment was carried out in normalizing temperature range of 950-1150 °C and then tempered to a fixed tempering temperature of 760 °C. The effect of varying normalizing temperatures before and after tempering on microstructure evolution, tensile properties, Vicker's hardness and Charpy toughness was studied. The normalizing temperature before and after tempering was having a noticeable effect on mechanical properties of as-received P92 steel. The fracture surface of impact and tensile tested samples was also studied for various normalizing temperatures with or without tempering. Fracture surface morphology was affected by the presence of secondary phase carbide particles. The fraction area of cleavage facets on the tensile fracture surface was found to be increased with an increase in the normalizing temperature. The fractured tensile specimens were characterized by transgranular ductile dimples, tear ridges and transgranular cleavage facets for various heat treatments. The fracture mode of impact tested samples was more complex. It showed both quasi-cleavage facets and ductile dimple tearing for various normalizing temperatures.

  19. 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.

  20. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  1. Analysis of the Plasticity-Enhancing Mechanisms in 12 pctMn Austeno-ferritic Steel by In Situ Neutron Diffraction

    Science.gov (United States)

    Lee, Sangwon; Woo, Wanchuck; De Cooman, Bruno C.

    2014-12-01

    The tensile behavior of ductile ultra-high strength Fe-12 pctMn-0.3 pctC-2 pctAl austeno-ferritic steel was studied by in situ neutron diffraction measurement of the elastic lattice strains, dislocation density, stacking fault probability, and strain-induced transformation kinetics. Micro-yielding was observed in austenite, and the plastic deformation of ferrite remained very limited throughout the deformation. The analysis identified three contributions to the strain hardening: twinning-induced plasticity, transformation-induced plasticity, and the accumulation of a high density of geometrically necessary dislocations accommodating the strain mismatch at the phase boundaries.

  2. Mechanical properties of phases in austeno-ferritic duplex stainless steel-Surface stresses studied by X-ray diffraction

    International Nuclear Information System (INIS)

    Dakhlaoui, Rim; Braham, Chedly; Baczmanski, Andrzej

    2007-01-01

    In this work the parameters characterizing the individual elastoplastic mechanical behaviour of each phase in austeno-ferritic duplex stainless steels are determined by using X-ray diffraction during a uniaxial tensile test. The interpretation of the experimental data is based on the diffraction elastic constants calculated by the self-consistent model taking the anisotropy of the studied materials into account. The elastoplastic model is used to predict the evolution of the internal stresses during loading, and to identify the critical resolved shear stresses and strain hardening parameters of the material. The effect of the chemical composition on the individual elastoplastic behaviour of the studied phases is established by comparing results from three different samples. Finally, the X-ray diffraction results are compared with those previously obtained by using neutron radiation

  3. Compliance variations in the fatigue thresold regime of a low alloy ferritic steel under closure-free testing conditions

    International Nuclear Information System (INIS)

    Vaidya, W.V.

    1991-01-01

    Compliance variations in the threshold regime of a high strength ferritic steel tested under closure-free conditions at room temperature and in air are reported. In contrast to the Paris regime, and irrespective of whether the data during load shedding, at threshold or after postthreshold load increase are considered, it is found that comparatively compliance varies inconsistently in the threshold regime. Therefore, a 1:1 correlation between the averaged optical crack length and that inferred from compliance was not observed. This discrepancy is analyzed. The variations in compliance are utilized to infer the crack front behavior, and the results are discussed in terms of the microstructural impedance. (orig.) With 22 figs., 2 appendices [de

  4. AFM and TEM study of cyclic slip localization in fatigued ferritic X10CrAl24 stainless steel

    International Nuclear Information System (INIS)

    Man, J.; Petrenec, M.; Obrtlik, K.; Polak, J.

    2004-01-01

    Atomic force microscopy and high resolution scanning electron microscopy were applied to the study of surface relief evolution at emerging persistent slip bands (PSBs) in individual grains of ferritic X10CrAl24 stainless steel cycled with constant plastic strain amplitude. Only the combination of both methods can reveal the true shape and fine details of extrusions and intrusions. Quantitative data on the changes of the surface topography of persistent slip markings and on the kinetics of extrusion growth during the fatigue life were obtained. Transmission electron microscopy of surface foils revealed PSBs with the typical, well-known ladder structure. Experimental data on cyclic slip localization in PSBs are compared with those in fcc metals and discussed in terms of vacancy models of surface relief evolution and fatigue crack initiation

  5. High heat loading properties of vacuum plasma spray tungsten coatings on reduced activation ferritic/martensitic steel

    Science.gov (United States)

    Tokunaga, K.; Hotta, T.; Araki, K.; Miyamoto, Y.; Fujiwara, T.; Hasegawa, M.; Nakamura, K.; Ezato, K.; Suzuki, S.; Enoeda, M.; Akiba, M.; Nagasaka, T.; Kasada, R.; Kimura, A.

    2013-07-01

    High density W coatings on reduced activation ferritic martensitic steel (RAF/M) have been produced by Vacuum Plasma Spraying technique (VPS) and heat flux experiments on them have been carried out to evaluate their possibility as a plasma-facing armor in a fusion device. In addition, quantitative analyses of temperature profile and thermal stress have been carried out using the finite element analysis (FEA) to evaluate its thermal properties. No cracks or exfoliation has been formed by steady state and cyclic heat loading experiments under heat loading at 700 °C of surface temperature. In addition, stress distribution and maximum stress between interface of VPS-W and RAF/M have been obtained by FEA. On the other hand, exfoliation has occurred at interlayer of VPS-W coatings near the interface between VPS-W and RAF/M at 1300 °C of surface temperature by cyclic heat loading.

  6. 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.

  7. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  8. Effect of cold-working on steam corrosion (5500C, 70 atm) of five ferritic steels with different chromium contents

    International Nuclear Information System (INIS)

    Leistikow, S.; Thenen, A.v.; Pott, E.

    Five different ferritic chromium steels were tested at 550 0 , 70 atm, in superheated steam to evaluate how steam corrosion depends on chromium content (0.8--18.7 percent) and the amount of cold work (0--90 percent). Oxidation kinetics and oxide morphology were examined by gravimetric, metallographic, and microanalytical methods. Results showed for all alloys (greater than or equal to 6.7 percent Cr)--in addition to the known effect of higher corrosion resistance by increasing chromium content--an increasing corrosion resistance with increasing cold work. A gain of 70--75 percent (related to the metal loss of the undeformed material) was found after 1000-h exposure in case of 80 percent cold-deformed alloys containing more than 11 percent Cr. The relationship between corrosion, chromium content, and cold deformation was approximated by a negative exponential function. Microanalysis revealed high-Cr local areas of the highly deformed oxidized 18.7 percent Cr alloy. The effect of the bulk deforming pretreatments upon corrosion rate is explained by improved diffusivity of Cr in the defect structure, preferential oxidation, and formation of a solid solution (FeFe/sub 2-x/ Cr/sub x/O 4 ) of spinel type as oxide subscale. Protective scales with low defect concentrations and reduced ion diffusivity were formed which are controlling the corrosion reaction. Thus, especially in case of highly deformed materials with medium chromium content the oxidation changed over from a parabolic to a logarithmic rate law. The experiment showed that corrosion of ferritic steels with medium chromium content can be predicted if the Cr content as well as the cold deforming pretreatment are taken into account. (U.S.)

  9. On the S-phase formation and the balanced plasma nitriding of austenitic-ferritic super duplex stainless steel

    Science.gov (United States)

    de Oliveira, Willian R.; Kurelo, Bruna C. E. S.; Ditzel, Dair G.; Serbena, Francisco C.; Foerster, Carlos E.; de Souza, Gelson B.

    2018-03-01

    The different physical responses of austenite (γ) and ferrite (α) iron structures upon nitriding result in technical challenges to the uniform modification of α-γ materials, as the super duplex stainless steel (SDSS). The effects of voltage (7-10 kV), frequency and pulse width on the nitrogen plasma immersion ion implantation of SDSS (α ∼ 56%, γ ∼ 44%) were investigated, correlated with structural, morphological and mechanical analyses. By controlling the treatment power, temperatures ranged from 292 °C to 401 °C. Despite the overall increase in hardness for any of the employed parameters (from ∼6 GPa to ∼15 GPa), the structure of individual grains was strikingly dissimilar at the same temperatures, depending on the energetic conditions of implantation. Modified-α grains containing iron nitrides (ε-Fe2-3N, γ‧ -Fe4N) presented intense brittleness, whereas the expanded phase γN (S-phase) laid principally in modified-γ grains, exhibiting ductile-like deformation features and thicker layers. The γN was the dominant phase in both α-γ grains at ∼401 °C, providing them with balanced structure and mechanical behavior. These phenomena corroborate with γN as mediator of the process, through a mechanism involving the nitrogen-promoted ferrite to austenite conversion and nitrides dissolution at high temperatures. An approximately linear correlation of the γN content with respect to the ion energy per pulse was demonstrated, which properly embodies limiting effects to the treatment. This can be a parameter for the α-γ steel surface modification, consisting in a better adjustment to obtain more precise control along with temperature.

  10. Heavy-section steel irradiation program: Embrittlement issues

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1991-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

  11. 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.

  12. Comparison of high temperature steam oxidation behavior of Zircaloy-4 versus austenitic and ferritic steels under light water reactor safety aspects

    International Nuclear Information System (INIS)

    Leistikow, S.; Schanz, G.; Zurek, Z.

    1985-12-01

    A comparative study of the oxidation behavior of Zy-4 versus steel No. 1.4914 and steel No. 1.4970 was performed in high temperature steam. Reactor typical tube sections of all three materials were exposed on both sides to superheated steam at temperatures ranging from 600 to 1300 0 C for up to 6 h. The specimens were evaluated by gravimetry, metallography, and other methods. The results are presented in terms of weight gain, corresponding metal (wall) penetration and consumption as function of time and temperature. Concerning the corrosion resistance the ranking position of Zy-4 was between the austenitic and the ferritic steel. Because of the chosen wall dimensions Zy-4 and the austenitic steel behaved similarly in that the faster oxidation of the thicker Zy-4 cladding consumed the total wall thickness in a time equivalent to the slower oxidation of the thinner austenitic steel cladding. The ferritic steel cladding however was faster consumed because of the lower oxidation resistance and the thinner wall thickness compared to the austenitic steel. So besides oxide scale formation, oxygen diffusion into the bulk of the metal forming various oxygen-containing phases were evaluated - also in respect to their influence on mechanical cladding properties and the dimensional changes. (orig./HP) [de

  13. Results of crack-arrest tests on irradiated a 508 class 3 steel

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  14. Fractographic and microstructural aspects of fracture toughness testing in irradiated 304 stainless steel

    International Nuclear Information System (INIS)

    Cullen, W.H.; Hiser, A.L.; Hawthorne, J.R.; Abramczyk, G.A.; Caskey, G.R.

    1987-01-01

    Fracture toughness and Charpy impact test results on 304 stainless steel baseplate, weld and heat-affected zone (HAZ) tested at 25 0 C and 125 0 C are correlated with the microstructural and fractographic features observed in these materials. Specimens were collected from several sections of 12.7 mm (0.5 in.) wall thickness piping removed from a process system, and were characterized by different material chemistries and thermomechanical histories. As a result, mechanical properties vary over a considerable range from one pipe section to another. The presence of delta ferrite in some of the samples caused significant degradations in the toughness properties for certain crack orientations. Decreases in Charpy impact energies occur in the same material for different crack orientations. Materials irradiated showed 40% decreases in Charpy impact energy, but little change in fracture morphology. An increase in the test temperature resulted in an expected increase in Charpy energies for all materials. Fractographic features did not change appreciably with respect to the 100 0 C increase in test temperature. In unirradiated specimens, a test temperature increase caused lower J/sub Ic/ and J-R curve values with tearing modules values increased. The latter is due to the large decreases in tensile strength with increasing test temperature. The weld metals tend to have the highest tearing resistance, while the HAZ's tend to have the lowest. 30 figs., 3 tabs

  15. 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)

  16. Swelling and irradiation creep of neutron irradiated 316Ti and 15-15Ti steels

    International Nuclear Information System (INIS)

    Maillard, A.; Touron, H.; Seran, J.L.; Chalony, A.

    1992-01-01

    The global behavior, the swelling and irradiation creep resistances of cold worked 316Ti and 15-15Ti, two variants of austenitic steels in use as core component materials of the French fast reactors, are compared. The 15-15Ti leads to a significant improvement due to an increase in the incubation dose swelling. The same phenomena observed on 316Ti are found on 15-15Ti. All species without fuel like samples, wrappers or empty clad swell and creep less than fuel pin cladding irradiated in the same conditions. To explain the swelling difference, as for 316Ti, thermal gradient is also invoked but the irradiation creep difference is not yet clearly understood. To predict the behavior of clads it is indispensable to study the species themselves and to use specific rules. All results confirm the good behavior of 15-15Ti, the best behavior being obtained with the 1% Si doped version irradiated up to 115 dpa

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

    International Nuclear Information System (INIS)

    Sheng Zhongqi; Xiao Hong; Peng Feng; Ti Zhongxin

    1994-04-01

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

  18. Effect of W and N on mechanical properties of reduced activation ferritic/martensitic EUROFER-based steel grades

    Science.gov (United States)

    Puype, A.; Malerba, L.; De Wispelaere, N.; Petrov, R.; Sietsma, J.

    2018-04-01

    The C, N, and W content in EUROFER97, a 9CrWVTa reduced-activation ferritic/martensitic (RAFM) steel, was varied to obtain an experimental assessment of the main effects of the compositional variation on the mechanical properties and microstructural characteristics of six different experimental grades. Light optical microscopy (LOM) and electron back-scattered diffraction (EBSD) revealed in almost all cases a fine tempered lath martensite structure. Analyses of transmission electron micrographs, together with inductively coupled plasma mass spectrometry (ICPMS) and energy-dispersive x-ray spectroscopy (EDS) data, shows the precipitation state and spatial distribution of MxCy (M = Cr, W and Fe) and MX (M = V and/or Ta, X = C or N) carbonitrides within the matrix. The mechanical characterization of the six different steel grades was carried out by means of A50 tensile testing and Charpy tests on standard specimens (55 × 10 × 10 mm3). Lowering the carbon content and keeping the nitrogen content higher than 0.02 wt%, leads to a reduction of the ductile-to-brittle-transition-temperature (DBTT) in comparison with EUROFER97-2. The addition of tungsten further reduces the DBTT to - 94 °C, while maintaining good tensile strength and elongation.

  19. Precipitation and impact toughness of Nb–V stabilised 18Cr–2Mo ferritic stainless steel during isothermal aging

    International Nuclear Information System (INIS)

    Han, Jian; Li, Huijun; Barbaro, Frank; Jiang, Laizhu; Zhu, Zhixiong; Xu, Haigang; Ma, Li

    2014-01-01

    The effect of isothermal aging on precipitation behaviour and Charpy impact toughness of Nb–V stabilised 18Cr–2Mo ferritic stainless steel was investigated by means of Thermo-Calc prediction, scanning electron microscopy, transmission electron microscopy, X-ray diffraction and Charpy impact toughness testing. The results show that, niobium, vanadium carbides and nitrides, Fe 2 Nb (Laves phase) and Cr 23 C 6 formed after 2 h aging at 800 °C, and the equilibrium solvus temperature of Fe 2 Nb phase increases to above 750 °C, higher than the calculated temperature (730 °C) using Thermo-Calc. After isothermal aging at 750–950 °C, 2 h aging resulted in a decrease in toughness due to the formation of precipitation, especially (Nb,V)(C,N) and Fe 2 Nb. When isothermally aged at 800 °C for up to 24 h, the coarsening rate of Fe 2 Nb particle is much higher than that of (Nb,V)(C,N), and the impact toughness of the steel is dependent on quantity and sizes of (Nb,V)(C,N) and Fe 2 Nb particles

  20. Influence of Texture on Impact Toughness of Ferritic Fe-20Cr-5Al Oxide Dispersion Strengthened Steel

    Science.gov (United States)

    Sánchez-Gutiérrez, Javier; Chao, Jesus; Vivas, Javier; Galvez, Francisco; Capdevila, Carlos

    2017-01-01

    Fe-based oxide dispersion strengthened (ODS) steels are oriented to applications where high operating temperatures and good corrosion resistance is paramount. However, their use is compromised by their fracture toughness, which is lower than other competing ferritic-martenstic steels. In addition, the route required in manufacturing these alloys generates texture in the material, which induces a strong anisotropy in properties. The V-notched Charpy tests carried out on these alloys, to evaluate their impact toughness, reveal that delaminations do not follow the path that would be expected. There are many hypotheses about what triggers these delaminations, but the most accepted is that the joint action of particles in the grain boundaries, texture induced in the manufacturing process, and the actual microstructure of these alloys are responsible. In this paper we focused on the actual role of crystallographic texture on impact toughness in these materials. A finite elements simulation is carried out to solely analyze the role of texture and eliminate other factors, such as grain boundaries and the dispersed particles. The work allows us to conclude that crystallographic texture plays an important role in the distribution of stresses in the Charpy specimens. The observed delaminations might be explained on the basis that the crack in the grain, causing the delamination, is directly related to the shear stresses τ12 on both sides of the grain boundary, while the main crack propagation is a consequence of the normal stress to the crack. PMID:28773104

  1. Determination of the fracture thoughness curve within the ductile brittle transition region in ferritic steel AISI4140

    International Nuclear Information System (INIS)

    Hernandez, R.; Orozco, E.

    1996-01-01

    The aim of this work is to show the validity in the employment of small test tubes (1/2 T) in order to determine the fracture thoughness in ferritic steels that experience the beginning of cracking by cleavage, to elastic instability, and/or elasto-plastic. It was calculated the change of fracture thoughness in the ductile brittle transition region like function of the temperature employing statistic methods for steel to the annealed carbon of the type AISI4140. The testings were carried out within an interval of temperatures, where the cracking by cleavage and/or pop-in occurs. The thoughness curve of the cracking in the transition region was determined, in small test tubes 1/2 T, and in standard test tubes, 1T. It was calculated the beginning of instability of the integral J, J IC , and was converted to its equivalent in K JC units based in the ASTM Standard rev. 6-12-95 (ref. 6). (Author)

  2. Effect of the thermal ageing on the tensile and impact properties of a 18%Cr ODS ferritic steel

    Energy Technology Data Exchange (ETDEWEB)

    Rouffié, A.L., E-mail: anne-laure.rouffie@cea.fr [CEA, DEN, DANS, DMN, SRMA, Bât 453, F-91191 Gif-sur-Yvette (France); Mines ParisTech, Centre des Matériaux P.M. Fourt, UMR CNRS 7633, BP 87, 91003 Evry (France); Crépin, J.; Sennour, M. [Mines ParisTech, Centre des Matériaux P.M. Fourt, UMR CNRS 7633, BP 87, 91003 Evry (France); Tanguy, B. [CEA, DEN, DANS, DMN, SEMI, Bât 625, F-91191 Gif-sur-Yvette (France); Pineau, A. [Mines ParisTech, Centre des Matériaux P.M. Fourt, UMR CNRS 7633, BP 87, 91003 Evry (France); Hamon, D.; Wident, P.; Vincent, S. [CEA, DEN, DANS, DMN, SRMA, Bât 453, F-91191 Gif-sur-Yvette (France); Garat, V. [AREVA NP, 10 rue J. Récamier, 69006 Lyon (France); Fournier, B. [Manoir Industries – Petrochem and Nuclear, Metallurgy Dpt., 12 rue des Ardennes, BP 8401 Pitres, 27108 Val de Reuil Cedex (France)

    2014-02-01

    The effects of the thermal ageing at 400 °C, 500 °C and 600 °C during 5000 h on the mechanical properties of a 18%Cr ODS ferritic steel are investigated. A hardening effect is observed after ageing at 400 °C and 500 °C, probably due to the presence of chromium rich α′ particles as suggested by the literature. The impact resistance and the ductility of the material are strongly lowered by the ageing at 600 °C. This embrittlement is characterized on the fracture surfaces by the presence of cleavage facets on the whole range of testing temperatures. The intermetallic σ phase is found to be responsible for the occurrence of cleavage fracture on the material aged at 600 °C, and thus for the significant embrittlement of this material. M{sub 23}C{sub 6} carbides are also observed before and after thermal ageing. The lattice parameters of the σ phase and the M{sub 23}C{sub 6} carbides observed in this 18%Cr ODS steel aged at 600 °C during 5000 h are measured.

  3. Texture development study during the primary recrystallization of ferritic steels by using X ray and electron backscattering diffraction

    International Nuclear Information System (INIS)

    Loew, Marjorie

    2006-01-01

    X ray and electron backscattering diffraction, in distinct levels, were applied to evaluate microstructural changes in two low carbon ferritic steels (2 per cent Si and ABNT 1006), observing the texture development in cold lamination step (skin-pass) and in the subsequent annealing at 760 deg C. In these two steels, results showed that after the skin-pass and annealing in the conditions of the present work, the observed phenomenon is the primary recrystallization. By applying skin-pass dislocations were introduced mostly in low Taylor factor grains as they are prone to be more deformed. Nucleation and grain growth were observed in high density dislocation cell regions. Silicon presence delayed the recovery favoring the sub-boundaries increase. It was not observed the abnormal grain growth, even in the presence of Gross grains. CSL boundaries did not guarantee the grains growth. Growing nuclei gave rise to grains with distinct orientations, showing that the grain growth was not dependent on the previous presence of grains with the developed orientation. This fact demonstrates that the abnormal grain growth is not necessarily related to the Gross grains. (author)

  4. Influence of Texture on Impact Toughness of Ferritic Fe-20Cr-5Al Oxide Dispersion Strengthened Steel.

    Science.gov (United States)

    Sánchez-Gutiérrez, Javier; Chao, Jesus; Vivas, Javier; Galvez, Francisco; Capdevila, Carlos

    2017-07-03

    Fe-based oxide dispersion strengthened (ODS) steels are oriented to applications where high operating temperatures and good corrosion resistance is paramount. However, their use is compromised by their fracture toughness, which is lower than other competing ferritic-martenstic steels. In addition, the route required in manufacturing these alloys generates texture in the material, which induces a strong anisotropy in properties. The V-notched Charpy tests carried out on these alloys, to evaluate their impact toughness, reveal that delaminations do not follow the path that would be expected. There are many hypotheses about what triggers these delaminations, but the most accepted is that the joint action of particles in the grain boundaries, texture induced in the manufacturing process, and the actual microstructure of these alloys are responsible. In this paper we focused on the actual role of crystallographic texture on impact toughness in these materials. A finite elements simulation is carried out to solely analyze the role of texture and eliminate other factors, such as grain boundaries and the dispersed particles. The work allows us to conclude that crystallographic texture plays an important role in the distribution of stresses in the Charpy specimens. The observed delaminations might be explained on the basis that the crack in the grain, causing the delamination, is directly related to the shear stresses τ 12 on both sides of the grain boundary, while the main crack propagation is a consequence of the normal stress to the crack.

  5. Aluminide slurry coatings for protection of ferritic steel in molten nitrate corrosion for concentrated solar power technology

    Science.gov (United States)

    Audigié, Pauline; Bizien, Nicolas; Baráibar, Ignacio; Rodríguez, Sergio; Pastor, Ana; Hernández, Marta; Agüero, Alina

    2017-06-01

    Molten nitrates can be employed as heat storage fluids in solar concentration power plants. However molten nitrates are corrosive and if operating temperatures are raised to increase efficiencies, the corrosion rates will also increase. High temperature corrosion resistant coatings based on Al have demonstrated excellent results in other sectors such as gas turbines. Aluminide slurry coated and uncoated P92 steel specimens were exposed to the so called Solar Salt (industrial grade), a binary eutectic mixture of 60 % NaNO3 - 40 % KNO3, in air for 2000 hours at 550°C and 580°C in order to analyze their behavior as candidates to be used in future solar concentration power plants employing molten nitrates as heat transfer fluids. Coated ferritic steels constitute a lower cost technology than Ni based alloy. Two different coating morphologies resulting from two heat treatment performed at 700 and 1050°C after slurry application were tested. The coated systems exhibited excellent corrosion resistance at both temperatures, whereas uncoated P92 showed significant mass loss from the beginning of the test. The coatings showed very slow reaction with the molten Solar Salt. In contrast, uncoated P92 developed a stratified, unprotected Fe, Cr oxide with low adherence which shows oscillating Cr content as a function of coating depth. NaFeO2 was also found at the oxide surface as well as within the Fe, Cr oxide.

  6. Stress–strain relationship between ferrite and martensite in a dual-phase steel studied by in situ neutron diffraction and crystal plasticity theories

    International Nuclear Information System (INIS)

    Woo, W.; Em, V.T.; Kim, E.-Y.; Han, S.H.; Han, Y.S.; Choi, S.-H.

    2012-01-01

    The stress–strain relationship between ferrite and martensite phases in the commercial dual-phase DP980 steel was studied using in situ neutron diffraction and the crystal plasticity finite element method (CPFEM). The phase identification method based on the image quality of electron backscatter diffraction and a filtering process was used to obtain information concerning individual crystallographic orientations for ferrite and martensite phases in DP980 steel. The (2 0 0) and (2 1 1) lattice strains of ferrite and martensite phases were measured along the loading and transverse directions as a function of macroscopic stress using in situ neutron diffraction. A CPFEM based on representative volume elements (RVE) was applied to determine the microscopic hardening parameters for each phase by fitting the measured macroscopic stress and measured (2 0 0) and (2 1 1) lattice strains. The microscopic hardening parameters for each phase successfully captured the influence of the crystallographic orientation of the ferrite phase on the localization of shear strain and the behavior of ductile failure in RVE of the unit cell during uniaxial tension.

  7. Nanostructure evolution in ODS steels under ion irradiation

    Directory of Open Access Journals (Sweden)

    S. Rogozhkin

    2016-12-01

    In this work, we carried out atom probe tomography (APT and transmission electron microscopy (TEM studies of three different ODS steels produced by mechanical alloying: ODS Eurofer, 13.5Cr ODS and 13.5Cr-0.3Ti ODS. These materials were investigated after irradiation with Fe (5.6MeV or Ti (4.8MeV ions up to 1015ion/cm2 and part of them up to 3×1015ion/cm2. In all cases, areas for TEM investigation were cut at a depth of ∼ 1.3µm from the irradiated surface corresponding to the peak of the radiation damage dose. It was shown that after irradiation at RT and at 300°С the number density of oxide particles in all the samples grew up. Meanwhile, the fraction of small particles in the size distribution has increased. APT revealed an essential increase in nanoclusters number and a change of their chemical composition at the same depth. The nanostructure was the most stable in 13.5Cr-0.3Ti ODS irradiated at 300°С: the increase of the fraction of small oxides was minimal and no change of nanocluster chemical composition was detected.

  8. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  9. Transformation behaviour of the ferritic-martensitic steels with 8-14 % chromium

    International Nuclear Information System (INIS)

    Schirra, M.; Finkler, H.

    2002-06-01

    Comprehensive development work on martensitic steels belonging to the so-called 12% Cr-steel group have been performed at the Institute for Materials Research of Forschungszentrum Karlsruhe in order to meet the various requirements in nuclear and conventional energy technology. The transformation characteristics of 62 different grades of steel and heats have been determined and continuous cooling transformation (CCT) diagrams have been prepared. The diagrams are described in a chronological sequence by subjects because the change in chemical composition can be correlated only partly with the transformation behaviour in cases where several alloying elements are simultaneously subjected to changes. In the introduction the basic difference is shown between isothermal and CCT diagrams and the transformation behaviour, respectively, by the example of the Nb-free steel 1.4922 (X20CrMoV 12 1) and the Nb-containing steel 1.4914 (X18CrMoVNb 12 1). (orig.)

  10. 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].

  11. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    International Nuclear Information System (INIS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

  12. Transport of carbon in 316 steels submitted to neutron irradiation

    International Nuclear Information System (INIS)

    Rouault, J.; Galland, L.; Cytermann, R.; Colin, M.

    1983-04-01

    The carburization of fast reactor cladding material may affect its mechanical properties and give rise to severe embrittlement. Carbon profiles were determined by EMPA in various irradiated carburized clads. All clads were in 316 type steels. An effective diffusion coefficient (Dsub(eff)) has been calculated for each profile. The set of Dsub(eff)) is shown in an Arrhenius Diagram. The experimental dispersion on Dsub(eff)) calculated values is due to the non-applicability of the model to a few profiles. The analysis is then made on the remaining Dsub(eff)). These values constitue a good coherent set of points. A comparison is then drawn between this set of points and: - true diffusion coefficient of carbon in the gamma phase, - effective diffusion coefficients of carbon derived from out-of-pile simulation experiments. Activation energy of Dsub(eff) coefficient (in pile and out-of-pile) is small compared too the activation of carbon diffusion in austenite. Dsub(eff) values are quite insensitive to surface concentration in the range 0,9 - 4%. Diffusion time is shown to have a great influence on Dsub(eff): Dsub(eff) decreases as time increases. A Dsub(eff) value for simple evaluations of carburization intensity in irradiated 316 steels is recommended [fr

  13. Study of Ferrite During Refinement of Prior Austenite Grains in Microalloyed Steel Continuous Casting

    Science.gov (United States)

    Liu, Jiang; Wen, Guanghua; Tang, Ping

    2017-12-01

    The formation of coarse prior austenite grain is a key factor to promote transverse crack, and the susceptibility to the transverse crack can be reduced by refining the austenite grain size. In the present study, the high-temperature confocal laser scanning microscope (CLSM) was used to simulate two types of double phase-transformation technologies. The distribution and morphology of ferrites under different cooling conditions were analyzed, and the effects of ferrite distribution and morphology on the double phase-transformation technologies were explored to obtain the suitable double phase-change technology for the continuous casting process. The results indicate that, under the thermal cycle TH0 [the specimens were cooled down to 913 K (640 °C) at a cooling rate of 5.0 K/s (5.0 °C/s)], the width of prior austenite grain boundaries was thick, and the dislocation density at grain boundaries was high. It had strong inhibition effect on crack propagation; under the thermal cycle TH1 [the specimens were cooled down to 1073 K (800 °C) at a cooling rate of 5.0 K/s (5.0 °C/s) and then to 913 K (640 °C) at a cooling rate of 1.0 K/s (1.0 °C/s)], the width of prior austenite grain boundary was thin, and the dislocation density at grain boundaries was low. It was beneficial to crack propagation. After the first phase change, the developed film-like ferrite along the austenite grain boundaries improved the nucleation conditions of new austenitic grains and removed the inhibition effect of the prior austenite grain boundaries on the austenite grain size.

  14. Numerical simulation of hydrogen-assisted crack initiation in austenitic-ferritic duplex steels; Numerische Simulation der wasserstoffunterstuetzten Rissbildung in austentisch-ferritischen Duplexstaehlen

    Energy Technology Data Exchange (ETDEWEB)

    Mente, Tobias

    2015-07-01

    Duplex stainless steels have been used for a long time in the offshore industry, since they have higher strength than conventional austenitic stainless steels and they exhibit a better ductility as well as an improved corrosion resistance in harsh environments compared to ferritic stainless steels. However, despite these good properties the literature shows some failure cases of duplex stainless steels in which hydrogen plays a crucial role for the cause of the damage. Numerical simulations can give a significant contribution in clarifying the damage mechanisms. Because they help to interpret experimental results as well as help to transfer results from laboratory tests to component tests and vice versa. So far, most numerical simulations of hydrogen-assisted material damage in duplex stainless steels were performed at the macroscopic scale. However, duplex stainless steels consist of approximately equal portions of austenite and δ-ferrite. Both phases have different mechanical properties as well as hydrogen transport properties. Thus, the sensitivity for hydrogen-assisted damage is different in both phases, too. Therefore, the objective of this research was to develop a numerical model of a duplex stainless steel microstructure enabling simulation of hydrogen transport, mechanical stresses and strains as well as crack initiation and propagation in both phases. Additionally, modern X-ray diffraction experiments were used in order to evaluate the influence of hydrogen on the phase specific mechanical properties. For the numerical simulation of the hydrogen transport it was shown, that hydrogen diffusion strongly depends on the alignment of austenite and δ-ferrite in the duplex stainless steel microstructure. Also, it was proven that the hydrogen transport is mainly realized by the ferritic phase and hydrogen is trapped in the austenitic phase. The numerical analysis of phase specific mechanical stresses and strains revealed that if the duplex stainless steel is

  15. Effect of the damage by radiation on the reference temperature T{sub 0} of ferritic steel; Efecto del dano por radiacion en la temperatura de referencia T{sub 0} de acero ferritico

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva O, A

    2004-07-01

    Presently work studies the effect that produces the irradiation in ferritic steels, on the reference temperature T{sub 0} (intrinsic characteristic of the fracture tenacity in the area of ductile-fragile transition), applying the approach of the Master curve that is based on the norm Astm E-1921. For it it was elaborated a methodology and procedure for test tubes type Charpy according to the standard before mentioned. Due to the ferritic steels are used mainly in pressure vessels to the reactor (RPV) of nuclear power plants; in the samples it was simulated the effect of the damage for irradiation through a thermal treatment that induced the precipitation of the carbides and sulfurs in the limits of grain (one of the modifications suffered in the irradiated materials); it was made a comparison later with material samples in initial state (without thermal treatment), used as witness sample, by means of assays of fracture mechanics, specifically flexion in three points; this way with it to observe the effect of the damage for irradiation in the reference temperature (T{sub 0}). This temperature (T{sub 0}) it is a very important parameter in the mechanical property of the material called fracture tenacity; which at the moment gives the rule for the verification of structural integrity of the RPV. As a result of this it was observed an increase in the reference temperature in the material in fragilezed state with respect to the initial state of 31.75 C. They were carried out metallographic analysis and fractographs of the assayed surface finding carbide inclusions and sulfurs that in theory of the Master Curve they are initiators of cracks and of a possible catastrophic flaw of the material. At the moment the Division of Scientific Investigation of the ININ is carrying out activities in the Nucleo electric Central of Laguna Verde (CNLV) related with the program of surveillance of the materials of the vessel of the unit 2, as well as projects of structural integrity

  16. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    In order to investigate the role of He in cavity nucleation in neutron irradiated iron and steel, pure iron and Eurofer-97 steel have been He implanted and neutron irradiated in a systematic way at different temperatures, to different He and neutron doses and with different He implantation rates....

  17. Microstructural Characteristic of Dissimilar Welded Components (AISI 430 Ferritic-AISI 304 Austenitic Stainless Steels) by CO2 Laser Beam Welding (LBW)

    OpenAIRE

    CALIGULU, Ugur; DIKBAS, Halil; TASKIN, Mustafa

    2012-01-01

    In this study, microstructural characteristic of dissimilar welded components (AISI 430 ferritic-AISI 304 austenitic stainless steels) by CO2 laser beam welding (LBW) was investigated. Laser beam welding experiments were carried out under argon and helium atmospheres at 2000 and 2500 W heat inputs and 100-200-300 cm/min. welding speeds. The microstructures of the welded joints and the heat affected zones (HAZ) were examined by optical microscopy, SEM, EDS and XRD analysis. The tensile strengt...

  18. Microstructural Characteristic of Dissimilar Welded Components (AISI 430 Ferritic-AISI 304 Austenitic Stainless Steels) by CO2 Laser Beam Welding (LBW)

    OpenAIRE

    ÇALIGÜLÜ, Uğur; CALIGULU, Ugur; DIKBAS, Halil; TASKIN, Mustafa

    2010-01-01

    In this study, microstructural characteristic of dissimilar welded components (AISI 430 ferritic-AISI 304 austenitic stainless steels) by CO2 laser beam welding (LBW) was investigated. Laser beam welding experiments were carried out under argon and helium atmospheres at 2000 and 2500 W heat inputs and 100-200-300 cm/min. welding speeds. The microstructures of the welded joints and the heat affected zones (HAZ) were examined by optical microscopy, SEM, EDS and XRD analysis. The tensile strengt...

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

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

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

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

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

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

  1. Modelling of the effect of precipitates on work-hardening, ductility and impact behaviour of ferritic-martensitic Cr steels

    Science.gov (United States)

    Preininger, D.

    2002-12-01

    The effect of precipitates on work-hardening, tensile ductility and impact behaviour of carbon and high nitrogen martensitic 7-12Cr as well as particle strengthened ODS-(9-13)Cr steels have been analysed by models. A minimum of work-hardening and uniform strain generally appears around 600 °C at onset of dislocation recovery. Pronounced precipitation by increasing nitrogen and carbon content or additionally of fine Y 2O 3-particles distinctly increases work-hardening and uniform ductility. These, however, decrease with increasing strengthening but do not reach a visible level above 1500 MPa for ODS-steels at 20 °C. Minima of total elongation and fracture strain additionally appear in carbon and nitrogen martensitic steels around 300 °C where dynamic strain ageing occurs. Fracture strain and ductile upper shelf energy of Charpy tests in accordance with model predictions also decrease with increasing yield strength more strongly for ODS-steels due to their enhanced work-hardening and localized deformation. The strength-induced increase of ductile-to-brittle transition temperatures of ODS-steels is comparable to that observed by irradiation defect strengthening.

  2. Effects of neutron irradiation on tensile and creep properties of stainless steels

    International Nuclear Information System (INIS)

    Miyaji, Noriko; Abe, Yasuhiro; Asayama, Tai; Aoto, Kazumi; Ukai, Shigeharu

    1997-01-01

    In order to investigate the effects of neutron irradiation on the creep and tensile properties of stainless steels, post-irradiation tests were made on the specimens of FBR grade type 316 stainless steel (316FR) and type 304 stainless steel. The post-irradiation tensile tests showed that the fracture elongation of both 316FR and type 304 stainless steel decreased and the 0.2% proof strength increased by irradiation. These phenomena are related to the point defect accumulation due to neutron irradiation. The post-irradiation creep test of 316FR demonstrated that the time to rupture decreased to between 1/3 and 1/5 of the unirradiated one, and this reduction is smaller than that of type 304 stainless steels under the same irradiation and test conditions. The creep property degradation of type 304 stainless steel due to the irradiation is caused by accumulation of helium bubbles at the grain boundaries. As for 316FR, it is considered that beyond the neutron exposure level of 0.3dpa a growth of phosphide caused a decrease in solution hardening and accumulation of helium bubbles at the grain boundaries. It is concluded that the reduction ratio of time to rupture for both 316FR and type 304 stainless steels after irradiation became larger than 1/30, which is the lower limit of the reduction ratio for the 'Monju' FBR. (author)

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

    International Nuclear Information System (INIS)

    Lloyd, G.J.

    1982-04-01

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

  4. Effects of fast neutron irradiation on the fracture behavior of stainless steel

    International Nuclear Information System (INIS)

    Huang, F.H.; Fish, R.L.

    1982-03-01

    In designing against premature fracture, the characteristics of materials must be measured and design criteria developed. The reduction in ductility for irradiated stainless steels has been observed, but little work has been conducted on evaluating the effects of notches on these materials. A reduction in notch ductility has been investigated in Charpy-V impact tests of irradiated Type 304 and Type 316 stainless steel; in irradiated Type 304 stainless steel, notch effects were not observed at 232 and 317 0 C, but as the test temperature was increased from 538 to 593 0 C, the material irradiated to a fluence of 3 X 10 22 n/cm 2 exhibited a notch weakening. Recently, similar experiments were performed on irradiated 20% cold worked Type 316 stainless steel to determine the effects of irradiation on the fracture behavior of this alloy

  5. Effect of strain rate and notch geometry on tensile properties and fracture mechanism of creep strength enhanced ferritic P91 steel

    Science.gov (United States)

    Pandey, Chandan; Mahapatra, M. M.; Kumar, Pradeep; Saini, N.

    2018-01-01

    Creep strength enhanced ferritic (CSEF) P91 steel were subjected to room temperature tensile test for quasi-static (less than 10-1/s) strain rate by using the Instron Vertical Tensile Testing Machine. Effect of different type of notch geometry, notch depth and angle on mechanical properties were also considered for different strain rate. In quasi-static rates, the P91 steel showed a positive strain rate sensitivity. On the basis of tensile data, fracture toughness of P91 steel was also calculated numerically. For 1 mm notch depth (constant strain rate), notch strength and fracture toughness were found to be increased with increase in notch angle from 45° to 60° while the maximum value attained in U-type notch. Notch angle and notch depth has found a minute effect on P91 steel strength and fracture toughness. The fracture surface morphology was studied by field emission scanning electron microscopy (FESEM).

  6. Effect of Cooling Rate on Precipitation Behavior and Micromechanical Properties of Ferrite in V-N Alloyed Steel During a Simulated Thermomechanical Process

    Science.gov (United States)

    Zhang, Jing; Wang, Fu-Ming; Yang, Zhan-Bing; Li, Chang-Rong

    2017-12-01

    The effect of the cooling rate after finish deformation at 1223 K (950 °C) on the microstructural evolution, V(C,N) precipitation, and micromechanical properties of ferrite in high-N V-alloyed building steel was comparatively investigated using a Gleeble-1500 thermomechanical simulator. Metallographic analysis shows that polygonal ferrite (PF) and pearlite (P) were dominant microconstituents at cooling rates ranging from 0.5 K/s to 3 K/s (0.5 °C/s to 3 °C/s). As the cooling rate increased within this range, the average ferrite grain size decreased from 6.1 ± 0.30 to 4.4 ± 0.25 μm. Besides, the sheet spacing of interphase precipitated V(C,N) particles decreased from 64.0 to 78.7 to 21.9 to 24.5 nm, and the average size of randomly precipitated particles was refined from 8.2 ± 3.24 to 6.3 ± 2.18 nm. The number density of precipitates with a size below and above 10 nm decreased, and the total number density decreased from 2482 ± 430 to 1699 ± 142 μm-2. Moreover, high-resolution transmission electron microscopy (HRTEM) observation revealed that there exists a coherent interface between the nanoscaled V(C,N) particle and the ferrite matrix. This interface lowered the nucleation energy barrier and promoted the V(C,N) particle precipitation in the ferrite matrix. Nanoindentation measurements indicated that the ferrite phase became softer, and the corresponding value of nanohardness and Young's modulus decreased as the cooling rate increased, which was caused predominantly by the decrease in precipitation hardening due to the lower number density of V(C,N) precipitates.

  7. Interpretation of dynamic tensile behavior by austenite stability in ferrite-austenite duplex lightweight steels.

    Science.gov (United States)

    Park, Jaeyeong; Jo, Min Cheol; Jeong, Hyeok Jae; Sohn, Seok Su; Kwak, Jai-Hyun; Kim, Hyoung Seop; Lee, Sunghak

    2017-11-16

    Phenomena occurring in duplex lightweight steels under dynamic loading are hardly investigated, although its understanding is essentially needed in applications of automotive steels. In this study, quasi-static and dynamic tensile properties of duplex lightweight steels were investigated by focusing on how TRIP and TWIP mechanisms were varied under the quasi-static and dynamic loading conditions. As the annealing temperature increased, the grain size and volume fraction of austenite increased, thereby gradually decreasing austenite stability. The strain-hardening rate curves displayed a multiple-stage strain-hardening behavior, which was closely related with deformation mechanisms. Under the dynamic loading, the temperature rise due to adiabatic heating raised the austenite stability, which resulted in the reduction in the TRIP amount. Though the 950 °C-annealed specimen having the lowest austenite stability showed the very low ductility and strength under the quasi-static loading, it exhibited the tensile elongation up to 54% as well as high strain-hardening rate and tensile strength (1038 MPa) due to appropriate austenite stability under dynamic loading. Since dynamic properties of the present duplex lightweight steels show the excellent strength-ductility combination as well as continuously high strain hardening, they can be sufficiently applied to automotive steel sheets demanded for stronger vehicle bodies and safety enhancement.

  8. 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.

  9. Austenite – ferrite transformation temperature regression equations for low carbon steels with cooling rate account

    Science.gov (United States)

    Záhumenský, P.; Kohútek, I.; Semeňák, J.

    2017-12-01

    The austenite-ferrite transformation temperatures evaluated by dilatometry using thermo-mechanical simulator Gleeble 1500D are investigated in this paper. The effect of cooling rates 1, 5, 10 and 15°C/s on the upper and lower critical transformation temperatures was evaluated for 30 specimens of six material groups. Considering the cooling rate from dilatometry tests and chemical composition (C ≤ 0.2%, Mn ≤ 2%, Si ≤ 0.26%) of particular specimens, the regression equations for both transformation temperatures were derived. These relations have to be satisfied to avoid the crack formation during continuous casting, as well as to provide the hot rolling control. The proposed regression equations are compared with 32 similar ones adopted from 1961 to 2017 and exhibit a good conformity and accuracy.

  10. Testing of Ni-plated ferritic steel interconnect in SOFC stacks

    DEFF Research Database (Denmark)

    Nielsen, K.A.; Dinesen, A.R.; Korcakova, L.

    2006-01-01

    Stack tests were run at 850 °C for periods from 80 hours to 1,150 hours to develop contacting procedures and at the same time evaluate the performance of a 5 μm electroplated nickel coating on a ferritic Fe22Cr interconnect. The metallic nickel coating reacted relatively quickly during the initial...... of the protective scale on the cathode side was susceptible to pitting-type corrosion patterns, which may limit the life expectancy to less than 2,000 hours for the 200 μm thick interconnect tested. The initial area-specific resistances (ASR) at the interconnect/cathode current collector interface...

  11. 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)

  12. Evaluation of ferritic alloy Fe-2 1/4Cr-1Mo after neutron irradiation: Microstructural development

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1986-10-01

    As part of a program to provide a data base on the bainitic alloy Fe-2-1/4-1Mo for fusion energy applications, microstructural examinations are reported for nine specimen conditions for 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510 0 C. Void swelling is found following irradiation at 400 0 C to 480 0 C. Concurrently dislocation structure and precipitation developed. Peak void swelling, void density, dislocation density and precipitate number density formed at the lowest temperature, approximately 400 0 C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior

  13. Nanoindentation on an oxide dispersion strengthened steel and a ferritic/martensitic steel implanted with He ions

    Science.gov (United States)

    Yang, Yitao; Kang, Suk Hoon; Zhang, Chonghong; Jang, Jinsung

    2014-12-01

    ODS steel MA956 and F/M steel T92 were implanted with 30 keV He ions to fluences of 3.0 × 1014 (0.013 at.%/0.0046 dpa), 3.0 × 1015 (0.13 at.%/0.046 dpa), 3.0 × 1016 (1.3 at.%/0.46 dpa) and 1.0 × 1017 ions/cm2 (4.5 at.%/1.5 dpa) at room temperature. Nanoindentation and TEM were used to investigate the nanohardness and microstructure change induced by He ion implantation. TEM results showed that He bubbles and a damage zone (∼250 nm) were observed in both materials at He concentration of 0.13 at.%, small cracks or connected bubbles in surface near region formed at He concentration of 4.5 at.%. Nanoindentation results showed that evident hardness increase was observed at the depth of 38 nm. The hardness peak at 38 nm shifted to 58 nm at He concentration of 4.5 at.%, which could be associated with the formation of small cracks or connected bubbles in surface near region. The damage layer was thin and close to surface, a method, proposed by Hosemann basing on the "rule of mixtures" model, was used to estimate the hardening effects from defects and He in this layer. The estimated results showed that the hardness increased rapidly with damage at low damage level, and started to increase slowly and presented a saturation trend at the damage level higher than ∼0.2 dpa. From the hardening fraction, significant hardening occurred for T92 compared with that for MA956, which indicated that ODS steel MA956 was better than F/M steel T92 in hardening resistance induced by He at room temperature.

  14. Microstructural observation of ion-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. The effect of nitrogen in sintered atmosphere of the ferritic stainless steels AISI 430L P/M; Efecto del nitrogeno en la atmosfera de sinterizacion del acero inoxiable ferritico AISI 430L P/M

    Energy Technology Data Exchange (ETDEWEB)

    Corpas, F. A.; Ruiz-Roman, J. M.; Codina, S.; Iglesias, F. J.

    2005-07-01

    In this paper, we have studied the nitrogen effects different sintering atmospheres (nitrogen-hydrogen, and dissociate ammonia) on ferritic stainless steels (430L), fabricated by powder metallurgy process. We have carried out a study of the physical (density, porosity and dimensional variation) and mechanical properties (hardness, tensile strength, and lengthening) of the ferritic stainless steels sintered in the afore-mentioned atmospheres, as well as of their behaviour in pitting corrosion. We have studied, also the microstructure of the steels, which depends on the atmosphere used for sintering. (Author) 13 refs.

  16. Effects of activating fluxes on the weld penetration and corrosion resistant property of laser welded joint of ferritic stainless steel

    Science.gov (United States)

    Wang, Yonghui; Hu, Shengsun; Shen, Junqi

    2015-10-01

    This study was based on the ferritic stainless steel SUS430. Under the parallel welding conditions, the critical penetration power values (CPPV) of 3mm steel plates with different surface-coating activating fluxes were tested. Results showed that, after coating with activating fluxes, such as ZrO2, CaCO3, CaF2 and CaO, the CPPV could reduce 100~250 W, which indicating the increases of the weld penetrations (WP). Nevertheless, the variation range of WP with or without activating fluxes was less than 16.7%. Compared with single-component ones, a multi-component activating flux composed of 50% ZrO2, 12.09% CaCO3, 10.43% CaO, and 27.49% MgO was testified to be much more efficient, the WP of which was about 2.3-fold of that without any activating fluxes. Furthermore, a FeCl3 spot corrosion experiment was carried out with samples cut from weld zone to test the effects of different activating fluxes on the corrosion resistant (CR) property of the laser welded joints. It was found that all kinds of activating fluxes could improve the CR of the welded joints. And, it was interesting to find that the effect of the mixed activating fluxes was inferior to those single-component ones. Among all the activating fluxes, the single-component of CaCO3 seemed to be the best in resisting corrosion. By means of Energy Dispersive Spectrometer (EDS) testing, it was found that the use of activating fluxes could effectively restrain the loss of Cr element of weld zone in the process of laser welding, thus greatly improving the CR of welded joints.

  17. Short crack growth and fatigue life in austenitic-ferritic duplex stainless steel

    Czech Academy of Sciences Publication Activity Database

    Polák, Jaroslav; Zezulka, Petr

    28 2005, č. 10 (2005), s. 923-935 ISSN 8756-758X R&D Projects: GA ČR(CZ) GA106/01/0376 Institutional research plan: CEZ:AV0Z20410507 Keywords : crack growth * crack initiation * duplex steel Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 0.673, year: 2004

  18. The microstructure effect on the fracture toughness of ferritic Ni-alloyed steels

    Energy Technology Data Exchange (ETDEWEB)

    Scheid, Adriano, E-mail: scheid@ufpr.br [Programa de Pos-Graduação em Engenharia Mecânica, PGMec, Universidade Federal do Paraná, Av. Cel. Francisco H. dos Santos, 210, Curitiba (Brazil); Félix, Lorenzo Marzari; Martinazzi, Douglas; Renck, Tiago; Fortis Kwietniewski, Carlos Eduardo [Programa de Pos-Graduação em Engenharia de Minas, Metalurgia e Materiais, PPGE3M, Universidade Federal do Rio Grande do Sul, Av. Bento Gonçalves, 9500, Porto Alegre (Brazil)

    2016-04-20

    Production of oil and gas in the Brazilian pre-salt faces several technical challenges and one of them that is a major concern is the presence of CO{sub 2} in high concentration. The aim of this work is to evaluate the fracture toughness of two nickel-containing steels as an alternative material to manufacture low-temperature toughness improved CO{sub 2} transporting pipelines for Enhanced oil recovery (EOR). Optical and scanning electron microscopies were employed to characterize the steels microstructures. Electron back-scattered diffraction was used to estimate the effective grain size and the density of high-angle grain boundaries. Fracture toughness was determined by the use of the crack tip opening displacement methodology. The results indicated that for the as-rolled condition the large islands of the microconstituent M/A in the 5{sup 1/2} Ni steel had a detrimental effect on fracture toughness at −100 °C, while finer M/A particles and lower effective grain size with higher density of high-angle grain boundaries in the 9 Ni steel turned its fracture toughness practically temperature independent. Additionally, heat treatment (quenching and tempering) has the potential to dissolve the M/A hard particles and consequently improve fracture toughness at low temperature.

  19. In situ neutron diffraction during tensile deformation of a ferrite-cementite steel

    Czech Academy of Sciences Publication Activity Database

    Tomota, Y.; Lukáš, Petr; Neov, Dimitar; Harjo, S.; Abe, YR.

    2003-01-01

    Roč. 51, č. 3 (2003), s. 805-817 ISSN 1359-6454 R&D Projects: GA ČR GA202/03/0891 Institutional research plan: CEZ:AV0Z1048901 Keywords : steel * neutron diffraction * stress-strain relationship measurement Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 3.059, year: 2003

  20. Evaluation of Perovskite Overlay Coatings on Ferritic Stainless Steels for SOFC Interconnect Applications

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Z Gary; Xia, Gordon; Maupin, Gary D.; Stevenson, Jeffry W.

    2006-08-02

    Conductive oxide coatings are used to improve electrical performance and surface stability of metallic interconnects, as well as to mitigate or prevent chromium poisoning in solid oxide fuel cells (SOFCs). To further understand materials suitability and shed light on mass transport, two conductive perovskites, were taken as examples and applied as dense coatings via radio frequency (rf)-sputtering on three stainless steels.

  1. Probing Formability Improvement of Ultra-thin Ferritic Stainless Steel Bipolar Plate of PEMFC in Non-conventional Forming Process

    Science.gov (United States)

    Bong, Hyuk Jong; Barlat, Frédéric; Lee, Myoung-Gyu

    2016-08-01

    Formability increase in non-conventional forming profiles programmed in the servo-press was investigated using finite element analysis. As an application, forming experiment on a 0.15-mm-thick ferritic stainless steel sheet for a bipolar plate, a primary component of a proton exchange membrane fuel cell, was conducted. Four different forming profiles were considered to investigate the effects of forming profiles on formability and shape accuracy. The four motions included conventional V motion, holding motion, W motion, and oscillating motion. Among the four motions, the holding motion, in which the slide was held for a certain period at the bottom dead point, led to the best formability. Finite element simulations were conducted to validate the experimental results and to probe the formability improvement in the non-conventional forming profiles. A creep model to address stress relaxation effect along with tool elastic recovery was implemented using a user-material subroutine, CREEP in ABAQUS finite element software. The stress relaxation and variable contact conditions during the holding and oscillating profiles were found to be the main mechanism of formability improvement.

  2. Characterisation of As-deformed microstructure of ODS NI-Base superalloy and ODS ferritic steel prior to directional recrystallisation

    International Nuclear Information System (INIS)

    Baloch, M.M.; Memon, S.A.

    2007-01-01

    The materials studied are unusual in the sense that they have been prepared from mechanically alloyed procedures, including compaction and hot extrusion. It was felt necessary to characterise the initial microstructure thoroughly prior to directional recrystallisation of the alloys. Following consolidation by hot extrusion, dispersion strengthened superalloys appear to display a very fine sub-micron grain size consisting of both dislocation free recrystallised material and un- recrystallised regions of high dislocation density. It is found that there is a very fine dislocation cell structure in the ODS (Oxide Dispersion Strengthened) Ferritic stainless Steel prior to recrystallisation treatment, which shows that alloy is in old-deformed condition after mechanical alloying, extrusion I hot-working. This is in contrast to the mechanically alloyed Nickel Base Superalloy, which have consistently been found to be in primary recrystallisation state following extrusion. In order to understand the recrystallisation behaviour of the two mechanically illoyed materials with commercial designations MA6000 and MA956, a measurement of the orientation relationship between adjacent grains in the as- deformed ODS alloys has also been carried out using Transmission Electron microscope. (author)

  3. Effects of hydrogen and loading mode on the fracture toughness of a reduced activation ferritic stainless steel

    International Nuclear Information System (INIS)

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

    1993-01-01

    The full spectrum of fracture toughness (J integrals), including pure mode I, different mixed mode I/III and pure mode III, will be examined for a ferritic/martensitic stainless steel with 0.1C-8Cr-2W-0.2V-0.04Ta-Fe (by wt%), designated as F-82H. The J integrals of pure mode I (J Ic ) and mixed mode I/III (J mixed ) are determined with single specimen method using standard compact tension specimens and modified compact tension specimens, respectively. The pure mode III integral is measured with multiple specimen method using 'triple-pantleg' specimens. The pure mode III integral is measured with multiple specimen method using 'triple-pantleg' specimens. Effects of hydrogen on the J integrals of pure mode I and mixed mode I/III are also going to be studied. 9 ppm H (about 500 appm) is pre-charged into specimens cathodically. The details of experimental procedure were described in this report. The preliminary results showed that addition of mode III stress (shear stress) to mode I loading had a significant negative effect on the fracture toughness of F-82H. The complete results, analysis and conclusion will be reported in next report. The results would be important to fusion reactor design

  4. Structure-Property-Fracture Mechanism Correlation in Heat-Affected Zone of X100 Ferrite-Bainite Pipeline Steel

    Science.gov (United States)

    Li, Xueda; Ma, Xiaoping; Subramanian, S. V.; Misra, R. D. K.; Shang, Chengjia

    2015-03-01

    Structural performance of a weld joint primarily depends on the microstructural characteristics of heat-affected zone (HAZ). In this regard, the HAZ in X100 ferrite-bainite pipeline steel was studied by separating the HAZ into intercritically reheated coarse-grained (ICCG) HAZ containing and non-containing regions. These two regions were individually evaluated for Charpy impact toughness and characterized by electron back-scattered diffraction (EBSD). Low toughness of ~50 J was obtained when the notch of impact specimen encountered ICCGHAZ and high toughness of ~180 J when the notch did not contain ICCGHAZ. Fracture surface was ~60 pct brittle in the absence of ICCGHAZ, and 95 pct brittle (excluding shear lip) in the presence of ICCGHAZ in the impact tested samples. The underlying reason is the microstructure of ICCGHAZ consisted of granular bainite and upper bainite with necklace-type martensite-austenite (M-A) constituent along grain boundaries. The presence of necklace-type M-A constituent notably increases the susceptibility of cleavage microcrack nucleation. ICCGHAZ was found to be both the initiation site of the whole fracture and cleavage facet initiation site during brittle fracture propagation stage. Furthermore, the study of secondary microcracks beneath CGHAZ and ICCGHAZ through EBSD suggested that the fracture mechanism changes from nucleation-controlled in CGHAZ to propagation-controlled in ICCGHAZ because of the presence of necklace-type M-A constituent in ICCGHAZ. Both fracture mechanisms contribute to the poor toughness of the sample contained ICCGHAZ.

  5. Precipitation behavior and martensite lath coarsening during tempering of T/P92 ferritic heat-resistant steel

    Science.gov (United States)

    Xu, Lin-qing; Zhang, Dan-tian; Liu, Yong-chang; Ning, Bao-qun; Qiao, Zhi-xia; Yan, Ze-sheng; Li, Hui-jun

    2014-05-01

    Tempering is an important process for T/P92 ferritic heat-resistant steel from the viewpoint of microstructure control, as it facilitates the formation of final tempered martensite under serving conditions. In this study, we have gained deeper insights on the mechanism underlying the microstructural evolution during tempering treatment, including the precipitation of carbides and the coarsening of martensite laths, as systematically analyzed by optical microscopy, transmission electron microscopy, and high-resolution transmission electron microscopy. The chemical composition of the precipitates was analyzed using energy dispersive X-ray spectroscopy. Results indicate the formation of M3C (cementite) precipitates under normalized conditions. However, they tend to dissolve within a short time of tempering, owing to their low thermal stability. This phenomenon was substantiated by X-ray diffraction analysis. Besides, we could observe the precipitation of fine carbonitrides (MX) along the dislocations. The mechanism of carbon diffusion controlled growth of M23C6 can be expressed by the Zener's equation. The movement of Y-junctions was determined to be the fundamental mechanism underlying the martensite lath coarsening process. Vickers hardness was estimated to determine their mechanical properties. Based on the comprehensive analysis of both the micro-structural evolution and hardness variation, the process of tempering can be separated into three steps.

  6. Effect of processing on microstructural features and mechanical properties of a reduced activation ferritic/martensitic EUROFER steel grade

    Science.gov (United States)

    Puype, A.; Malerba, L.; De Wispelaere, N.; Petrov, R.; Sietsma, J.

    2017-10-01

    The microstructure of a 9Cr-1W-0.22V-0.09Ta-0.11C reduced activation ferritic/martensitic (RAFM) steel has been investigated after thermo-mechanical rolling with subsequent annealing for 30 min at temperatures of 880 °C, 920 °C, 980 °C and 1050 °C, followed by water quenching. Scanning and transmission electron microscopy investigations and electron backscattered diffraction (EBSD) measurements were performed to determine the microstructural features after the different thermal treatments. Additionally, the microstructure and the mechanical properties of the materials were studied after tempering at 750 °C for 2 h. This study aims to understand microstructural processes that occur in the material during thermo-mechanical treatment and to assess the effect of the microstructure on its strength and toughness, with a view on improving its mechanical performance. Microstructural analysis together with the data from mechanical tests identified the beneficial effect of grain refinement obtained with adequate processing on the ductile-to-brittle transition temperature (DBTT) and on the delay of strength degradation at elevated temperatures.

  7. Strain-based plastic instability acceptance criteria for ferritic steel safety class 1 nuclear components under level D service loads

    Directory of Open Access Journals (Sweden)

    Ji-Su Kim

    2015-04-01

    Full Text Available This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1 a section average approach and (2 a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

  8. Initiation of Stress Corrosion Cracking of 26Cr-1Mo Ferritic Stainless Steels in Hot Chloride Solution

    International Nuclear Information System (INIS)

    Kwon, H. S.; Hehemann, R. F.

    1987-01-01

    Elongation measurements of 26Cr-1Mo ferritic stainless steels undergoing stress corrosion in boiling LiCl solution allow the induction period to be distinguished from the propagation period of cracks by the deviation of elongation from the logarithmic creep law. Localised corrosion cells are activated exclusively at slip steps by loading and developed into corrosion trenches. No cracks have developed from the corrosion trenches until the induction period is exceeded. The induction period is regarded as a time for localised corrosion cells to achieve a critical degree of occlusion for crack initiation. The repassivation rate of exposed metal by creep or emergence of slip steps decreases as the load increases and is very sensitive to the microstructural changes that affect slip tep height. The greater susceptibility to stress corrosion cracking of either prestrained or grain coarsened 26Cr-1Mo alloy compared with that of mill annealed material results from a significant reduction of repassivation rate associated with the increased slip step height. The angular titanium carbonitrides particles dispersed in Ti-stabilized 26Cr-1Mo alloy have a detrimental effect on the resistance to stress corrosion cracking

  9. Microstructural evolution in a ferritic-martensitic stainless steel and its relation to high-temperature deformation and rupture models

    Energy Technology Data Exchange (ETDEWEB)

    DiMelfi, R.J.; Gruber, E.E.; Kramer, J.M.

    1991-01-01

    The ferritic-martensitic stainless steel HT-9 exhibits an anomalously high creep strength in comparison to its high-temperature flow strength from tensile tests performed at moderate rates. A constitutive relation describing its high-temperature tensile behavior over a wide range of conditions has been developed. When applied to creep conditions the model predicts deformation rates orders of magnitude higher than observed. To account for the observed creep strength, a fine distribution of precipitates is postulated to evolve over time during creep. The precipitate density is calculated at each temperature and stress to give the observed creep rate. The apparent precipitation kinetics thereby extracted from this analysis is used in a model for the rupture-time kinetics that compares favorably with observation. Properly austenitized and tempered material was aged over times comparable to creep conditions, and in a way consistent with the precipitation kinetics from the model. Microstructural observations support the postulates and results of the model system. 16 refs., 10 figs.

  10. 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)

  11. Critical Heat Flux Test with the Ferritic Martensitic Steel Mock-ups for the DEMO Blanket First Wall

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Chang, Doo Hee; Song, Woo Sob; Hong, Bong Geun

    2009-01-01

    Korea has proposed and designed a DEMO concept considering a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). In these concepts, Ferritic Martensite Steel (FMS) is used as the structural material. The blanket FW of these concepts is an important component which faces the plasma directly and therefore, it is subjected to high heat and neutron loads. The FW is composed of the FMS as a structural material and an armor material such as tungsten and beryllium. Fabrication technology have been being developed especially for the joining between an armor material and FMS and more the Critical Heat Flux (CHF) should be investigated for design and safety aspect. In the present study, three FMS mock-ups without armor material were fabricated with a HIP (Hot Isostatic Pressing), which was developed similarly to the development of the ITER blanket FW in Korea. And they were tested in the high heat flux (HHF) test facility

  12. 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).

  13. Strain-based plastic instability acceptance criteria for ferritic steel safety class 1 nuclear components under level D

    International Nuclear Information System (INIS)

    Kim, Ji Su; Lee, Han Sang; Kim, Yun Jae; Kim, Jong Sung; Kim, Jin Won

    2015-01-01

    This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the over conservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

  14. Influence of Powder Outgassing Conditions on the Chemical, Microstructural, and Mechanical Properties of a 14 wt% Cr Ferritic ODS Steel

    Science.gov (United States)

    Sornin, D.; Giroux, P.-F.; Rigal, E.; Fabregue, D.; Soulas, R.; Hamon, D.

    2017-11-01

    Oxide dispersion-strengthened ferritic stainless steels are foreseen as fuel cladding tube materials for the new generation of sodium fast nuclear reactors. Those materials, which exhibit remarkable creep properties at high temperature, are reinforced by a dense precipitation of nanometric oxides. This precipitation is obtained by mechanical alloying of a powder and subsequent consolidation. Before consolidation, to obtain a fully dense material, the powder is vacuumed to outgas trapped gases and species adsorbed at the surface of the powder particles. This operation is commonly done at moderate to high temperature to evacuate as much as possible volatile species. This paper focuses on the influence of outgassing conditions on some properties of the further consolidated materials. Chemical composition and microstructural characterization of different materials obtained from various outgassing cycles are compared. Finally, impact toughness of those materials is evaluated by using Charpy testing. This study shows a significant influence of the outgassing conditions on the mechanical properties of the consolidated material. However, microstructure and oxygen contents seem poorly impacted by the various outgassing conditions.

  15. Fe2+ ion irradiated JRQ steel investigated by nanoindentation and slow-positron Doppler broadening spectroscopy

    Science.gov (United States)

    Pecko, Stanislav; Heintze, Cornelia; Bergner, Frank; Anwand, Wolfgang; Slugeň, Vladimír

    2018-01-01

    A model reactor pressure vessel (RPV) steel, known as JRQ, was manufactured in Japan for IAEA neutron embrittlement research studies in late 80 s. This model alloy belongs to the commercially used steel of A533B-1 type and shows relatively large changes in mechanical properties after a neutron irradiation due to considerable copper content (0.15 wt%). In order to simulate neutron irradiation and investigate the hardening effect, studied specimens of JRQ steel were exposed to Fe2+ ion irradiation in five different exposures calculated using the SRIM code. The ion energy of 5 MeV, temperature at 300 °C and the flux of 1.0 × 1011 cm-2 s-1 were the same during the irradiations. The hardening was investigated and observed by means of nanoindentation technique and a defect profile of irradiated steels was measured by Slow-positron Doppler broadening spectroscopy (DBS). The observed increasing trend of nanohardness as a function of fluence is in good agreement with the trend observed on the basis of Vickers hardness measured for neutron-irradiated JRQ. This confirms that Cu precipitation is most likely responsible for the observed irradiation hardening and that neutron-irradiation-induced damage can be simulated using ion irradiation in the present case. We have also excluded open volume (vacancy type) defects in the crystal lattice of JRQ steel from a responsibility for the damage arising by the Fe2+ ion irradiation.

  16. ACICULAR FERRITE

    Directory of Open Access Journals (Sweden)

    BOLSHAKOV V. I.

    2015-09-01

    Full Text Available Intermediate austenite transformation develops in the temperature between the regions pearlitic and martensitic transformation [4]. Under continuous cooling steel at speeds below the critical value, but higher than those necessary for the decomposition of austenite by the diffusion mechanism, the formation of a mixture of different types of structures whose identification is not always unambiguous. This resulted in a different classification systems of microstructures of low-carbon steel after accelerated cooling and the absence of a common terminology relating to the products of austenite decomposition [3; 5 – 11]. In modern terminology, all of the intermediate transformation product classifications based on the differentiation of the following features – the morphology of bainite ferrite component (rack or plate, the presence of iron carbide precipitates, their distribution and morphology, as well as the presence or absence of residual austenite or martensite-austenite mixture. Identification of the products of the intermediate conversion not morphology ferrite component, and other characteristics by light microscopy is extremely difficult, and in some instances impossible due to the limited resolution of the light microscope, so for these purposes should be to use the method of transmission electron microscopy of thin foils. Electron microscopy studies show that low-carbon steels lamellar morphology of intermediate products decomposition of austenite is extremely rare, which is confirmed by foreign authors [2; 7; 9; 10].

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

    Science.gov (United States)

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

    2017-06-01

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

  18. Effect of zirconium addition on the microstructure and mechanical properties of 15Cr-ODS ferritic Steels consolidated by hot isostatic pressing

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Haijian, E-mail: haijianxu@eis.hokudai.ac.jp [Key Laboratory for Anisotropy and Texture of Materials, Ministry of Education, Northeastern University, Shenyang 110819 (China); Material Science and Engineering, Faculty of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Lu, Zheng; Wang, Dongmei; Liu, Chunming [Key Laboratory for Anisotropy and Texture of Materials, Ministry of Education, Northeastern University, Shenyang 110819 (China)

    2017-01-15

    The influence of Zr addition on the microstructure and mechanical properties of mechanically alloyed (MA) ODS ferritic steels were studied in this work. The microstructure characteristics included the grain size, oxide particles number densities, size distributions, crystal structures and compositions. TEM foils measurements were complemented by studies of alloys on carbon extraction replica and focus ion beam (FIB) foils. The tensile properties were carried out at different temperatures. The microstructure and mechanical properties were analyzed and compared with nominal compositions (wt.%): Fe-15Cr-2W-0.3Y{sub 2}O{sub 3} and Fe-15Cr −2W-0.3Zr-0.3Y{sub 2}O{sub 3}. The experimental revealed that the addition of Zr increased the volume fraction of the smallest and equiaxed ferritic grains, number density of nano-oxide particles and decreased the average size of oxide particles within the ferritic matrix, promoting the formation of fine trigonal δ-phase Y{sub 4}Zr{sub 3}O{sub 12} nano-oxides and leading to the enhancement of the mechanical properties of the ODS steels.

  19. Degradation of physical and mechanical properties of steel G-91 under low-dose neutron irradiation

    Science.gov (United States)

    Kislitsin, Sergey; Dikov, Alexey; Maksimkin, Oleg; Merezhko, Mikhail; Rofman, Oleg; Turubarova, Lyudmila; Gorlachev, Igor; Sil'nagina, Nadezhda

    2017-12-01

    Changes in the structure and physicomechanical properties of steel G-91 were studied after low-dose neutron irradiation. The irradiation was carried out in the "wet" channel of the WWR-K research nuclear reactor of INP, Almaty, Kazakhstan, to the fast neutron fluencies 8.6×1019 n/cm2 at a temperature of steel G-91. Microstructural changes manifested themselves in a growth of dislocation density and appearance of radiation defects (black dots). The most significant consequence of low-dose irradiation during a long period (up to a year and a half) is severe corrosion, which leads to embrittlement of steel G-91.

  20. Effect of residual stresses on individual phase mechanical properties of austeno-ferritic duplex stainless steel

    International Nuclear Information System (INIS)

    Dakhlaoui, R.; Baczmanski, A.; Braham, C.; Wronski, S.; Wierzbanowski, K.; Oliver, E.C.

    2006-01-01

    The mechanical properties of both phases in duplex stainless steel have been studied in situ using neutron diffraction during mechanical loading. Important differences in the evolution of lattice strains are observed between tests carried out in tension and compression. An elastoplastic self-consistent model is used to predict the evolution of internal stresses during loading and to identify critical resolved shear stresses and strain hardening parameters of the material. The differences between tensile and compressive behaviours of the phases are explained when the initial stresses are taken into account in model calculations. The yield stresses in each phase of the studied steel have been experimentally determined and successfully compared with the results of the elastoplastic self-consistent model

  1. The Effect of Temperature on the Chromizing Process for Ferritic-Martensitic Steel

    Science.gov (United States)

    Alia, F. F.; Kurniawan, T.; Ani, M. H. B.; Nandiyanto, A. B. D.

    2017-10-01

    The formation of protective Cr2O3 layer was usually retarded in the high temperature steam oxidation of boiler tube materials. This condition makes the oxidation rate higher than that in dry condition. Therefore in this work, chromizing process is introduced to diffuse chromium on the surface of boiler steel so that it can act as a reservoir for the formation of Cr2O3 layer. The chromizing process was conducted on T91 steel by exposing it into alumina crucible. The crucible was exposed at different temperature (600°C-1050°C) under argon environment in the crucible that contains the chromizing mixture powder of masteralloy Cr, activator NH4Cl and filler Al2O3. It was found that Cr diffusion was happened at higher temperature and it formed Cr carbide on the surface. It also clarified that this chromizing process can prevent the retardation of Cr2O3 layer.

  2. Microstructure stability and creep behaviour of advanced high chromium ferritic steels

    Czech Academy of Sciences Publication Activity Database

    Sklenička, Václav; Kuchařová, Květa; Kudrman, J.; Svoboda, Milan; Kloc, Luboš

    43 2005, č. 1 (2005), s. 20-33 ISSN 0023-432X R&D Projects: GA ČR(CZ) GA106/02/0608; GA AV ČR(CZ) IAA2041101; GA AV ČR(CZ) 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : 9-12%Cr steels * microstructure stability * creep behaviour * nonsteady creep loading Subject RIV: JG - Metallurgy Impact factor: 0.973, year: 2005

  3. Comparison of low stress creep properties of ferritic and austenitic creep resistant steels

    Czech Academy of Sciences Publication Activity Database

    Kloc, Luboš; Sklenička, Václav; Ventruba, J.

    319-321, - (2001), s. 774-778 ISSN 0921-5093. [International Conference on Strength of Materials /12./. Monterey, CA, USA, 27.08.2000-01.09.2000] R&D Projects: GA AV ČR IAA2041702; GA MŠk OC 522.40 Institutional research plan: CEZ:AV0Z2041904 Keywords : viscous creep * power-law creep * creep-resistant steel Subject RIV: JG - Metallurgy Impact factor: 0.978, year: 2001

  4. Stress corrosion cracking of ferritic reactor pressure vessel steels under boiling water reactor conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.

    2001-01-01

    The stress corrosion cracking (SCC) behaviour of low-alloy reactor pressure vessel (RPV) steels in oxygenated high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both, RPV structural integrity and safety, has been a subject of controversial discussions for many years. The SCC crack growth behaviour of different RPV steels under simulated BWR/NWC conditions was therefore characterized by constant load and ripple load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. Modern high-temperature water loops, online crack growth monitoring (DCPD) and fractographical analysis by scanning electron microscopy were used to quantify the cracking response. It is concluded that there is no susceptibility to sustained SCC crack growth at temperatures around 288 C under purely static loading, as long as small-scale-yielding conditions prevail at the crack tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water chemistry conditions (EPRI Action Level 3) and/or for highly stressed specimens, either loaded near to K IJ or with a high degree of plasticity in the remaining ligament. The conservative character of the 'BWR VIP 60 Disposition Lines 1 and 2' for SCC crack growth in low-alloy steels has been confirmed by this study for 288 C and RPV base material. Preliminary results indicate, that these disposition lines may be significantly or slightly exceeded (even in steels with low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 - 250 C) in RPV materials, which show a distinct susceptibility to Dynamic Strain Ageing (DSA). (orig.)

  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. Application of electron backscatter diffraction (EBSD) to fracture studies of ferritic steels.

    Science.gov (United States)

    Davies, P A; Novovic, M; Randle, V; Bowen, P

    2002-03-01

    The application of electron backscatter diffraction (EBSD) to fracture studies has provided a new method for investigating the crystallography of fracture surfaces. The crystallographic indices of cleavage planes can be measured both directly from the fracture surface and indirectly from metallographic sections perpendicular to the plane of the adjoining fracture surfaces. The results of direct individual cleavage facet plane orientation measurements are presented for carbon-manganese (C-Mn) and low-alloy Mn-Mo-Ni (similar to ASTM A553 type-B). Pressure vessel steel weld metals, obtained from fracture surfaces of Charpy impact test specimens fractured at various test temperatures and for an ultra-low carbon steel (Fe-0.002C- 0.058P) fractured at -196 degrees C by impact. In addition to the direct measurement from the fracture surface, cleavage facet orientation measurements for the ultra-low carbon steel were complemented by the results obtained from the metallographic sections. Fractographic observations revealed that cleavage fracture is accommodated by a microvoid coalescence fracture micromechanism, which was induced by decohesion of second phase particles (inclusions). The correlation between the direct and indirect methodologies shows that the cleavage facet planes are dominated by the [001] plane orientations, and indicated that even when information concerning the full five degrees of freedom is inaccessible, the cleavage facet plane could still be determined. Finally, the advantages and disadvantages of direct orientation measurements from the fracture surface and indirectly by a destructive sectioning technique are discussed.

  7. A simplified LBB evaluation procedure for austenitic and ferritic steel piping

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, R.M.; Wichman, K.R.

    1997-04-01

    The NRC previously has approved application of LBB analysis as a means to demonstrate that the probability of pipe rupture was extremely low so that dynamic loads associated with postulated pipe break could be excluded from the design basis (1). The purpose of this work was to: (1) define simplified procedures that can be used by the NRC to compute allowable lengths for circumferential throughwall cracks and assess margin against pipe fracture, and (2) verify the accuracy of the simplified procedures by comparison with available experimental data for piping having circumferential throughwall flaws. The development of the procedures was performed using techniques similar to those employed to develop ASME Code flaw evaluation procedures. The procedures described in this report are applicable to pipe and pipe fittings with: (1) wrought austenitic steel (Ni-Cr-Fe alloy) having a specified minimum yield strength less than 45 ksi, and gas metal-arc, submerged arc and shielded metal-arc austentic welds, and (2) seamless or welded wrought carbon steel having a minimum yield strength not greater than 40 ksi, and associated weld materials. The procedures can be used for cast austenitic steel when adequate information is available to place the cast material toughness into one of the categories identified later in this report for austenitic wrought and weld materials.

  8. Strength and toughness of ferritic structural steels between 20deg C and 400deg C

    International Nuclear Information System (INIS)

    Memhard, D.

    1989-05-01

    The objective of the study was to quantitativley describe within the temperature range of 20deg C and 400deg C the strength and toughness behaviour of construction steels with diverse characteristics in dependence of stress rate and temperature and furthermore to evaluate the time dependence of material behaviour by comparison of experiments under tension load and constant loads. To this end tension tests at expansion rates of 10 -2 to 10 -7 s -1 of the steels 15 MnNi 6 3 and MnMoNi 5 5 and on screw steel 21 CrMoV 5 7, relaxation and short term creep tests were conducted. Parallel to this fracture mechanics tests on CT-specimens with a tension rate of 10 -1 to 10 -4 mm/s and creep test under constant load were conducted. The three basic test enabled a comprehensive description of the material behaviour under uniaxial tensile load. In the field of ductile material failure where the crack propagation is accompanied by noticeable plastic deformation of the cracked structure, the temperature and tension rate dependence of the flow behaviour on the crack resistance behaviour is noticeable. In contrast to crack resistance behaviour in path-controlled experimental procedures time-dependent plastic deformations in short term creep tests can no longer be neglected. (orig./MM) With 129 figs., 4 tabs., 154 refs [de

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. The Effects of One and Double Heat Treatment Cycles on the Microstructure and Mechanical Properties of a Ferritic-Bainitic Dual Phase Steel

    Science.gov (United States)

    Piri, Reza; Ghasemi, Behrooz; Yousefpour, Mardali

    2018-03-01

    In this study, samples with ferritic-bainitic dual phase structures consisting of 62 pct bainite were obtained from the AISI 4140 steel by applying one and double heat treatment cycles. Microstructural investigations by electron and optical microscopy indicated that the sample heat treated through double cycle benefited from finer ferrite and bainite grains. Additionally, results obtained from mechanical tests implied that the double-cycle heat-treated sample not only has a higher tensile strength as well as ultimate strength but also benefits from a higher ductility along with a higher impact energy than the one-cycle heat-treated sample. Moreover, fractography results showed that the type of fracture in both samples is a combination of the brittle and the ductile fracture. Besides, the ratio of the ductile fracture is higher for the double-cycle heat-treated sample than for the one-cycle sample, due to the lower aggregation of sulfur at grain boundaries.

  11. Analysis of ferritic stainless steel tube applied in radiation furnaces; Analise de tubos de aco inoxidavel ferritico para aplicacao em fornos de radiacao

    Energy Technology Data Exchange (ETDEWEB)

    Porto, P.C.R.; Spim, J.A. [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Centro de Tecnologia. Lab. de Fundicao], e-mail: spim@ufrgs.br; Santos, C.A. [Pontificia Universidade Catolica do Rio Grande do Sul (PUC-RS), Porto Alegre, RS (Brazil). Fac. de Engenharia. Programa de Pos-Graduacao em Engenharia e Tecnologia de Materiais (PGETEMA)

    2006-07-01

    The objective of this work was to evaluate the change in mechanic properties and phase transformations of ferritic stainless steel tube, ASTM 268 Gr 446, applied in high temperature conditions. The work has used tubes from radiation furnaces of the PETROBRAS Xisto Industrialization Unit. The samples used for comparison were obtained from new tubes and tubes already used in furnaces. The test analyses were optical metallography, scanning electron microscopy, energy dispersive spectrometer, hardness and microhardness test and tension test. Results have shown that the new tubes presented a ferritic matrix and in old tubes were observed a great quantity of sigma phase and carbides. Along with the thickness of the tubes it was verified that the inside region presented an increase of sulfate and the outside region an increase of carbides. (author)

  12. 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.

  13. Final report for the year 2001 on experimental and theoretical investigations of irradiation effects on physical and mechanical properties of iron and RAFM steels

    International Nuclear Information System (INIS)

    Singh, B.N.

    2003-08-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties have been studied both experimentally and theoretically. Specimens of pure iron and RAFM (reduced activation ferritic-martensic) steels were irradiated to different dose levels and at different irradiation temperatures. The resulting microstructure was characterized using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. Mechanical properties were determined by uniaxial tensile testing. Dislocation-loop interaction, formation of rafts of loops, radiation hardening and formation of 'cleared channels' were studied using different computational techniques. Experiments have shown that nano-voids are formed both in pure iron and F82H steel already at 50 deg. C. In pure iron, the formation of nano-voids is detected already at a dose level of ∼10 -3 dpa. Also in iron, self-interstitial atoms were found to accumulate in the form of glissile and sessile loops; at higher dose levels, these loops led to formation of rafts of loops. Irradiation led to an increase in the yield strength, a sudden drop in the yield stress, and, at higher doses, the initiation of plastic instability immediately beyond the upper yield point. Experimental as well as the results of computer simulations are found to be consistent with the cascade induced source hardening model

  14. Final report for the year 2001 on experimental and theoretical investigations of irradiation effects on physical and mechanical properties of iron and RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N

    2003-08-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties have been studied both experimentally and theoretically. Specimens of pure iron and RAFM (reduced activation ferritic-martensic) steels were irradiated to different dose levels and at different irradiation temperatures. The resulting microstructure was characterized using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. Mechanical properties were determined by uniaxial tensile testing. Dislocation-loop interaction, formation of rafts of loops, radiation hardening and formation of 'cleared channels' were studied using different computational techniques. Experiments have shown that nano-voids are formed both in pure iron and F82H steel already at 50 deg. C. In pure iron, the formation of nano-voids is detected already at a dose level of {approx}10{sup -3} dpa. Also in iron, self-interstitial atoms were found to accumulate in the form of glissile and sessile loops; at higher dose levels, these loops led to formation of rafts of loops. Irradiation led to an increase in the yield strength, a sudden drop in the yield stress, and, at higher doses, the initiation of plastic instability immediately beyond the upper yield point. Experimental as well as the results of computer simulations are found to be consistent with the cascade induced source hardening model.

  15. Intergranular stress corrosion cracking of ion irradiated 304L stainless steel in PWR environment

    OpenAIRE

    Gupta, Jyoti

    2016-01-01

    IASCC is irradiation – assisted enhancement of intergranular stress corrosion cracking susceptibility of austenitic stainless steel. It is a complex degrading phenomenon which can have a significant influence on maintenance time and cost of PWRs’ core internals and hence, is an issue of concern. Recent studies have proposed using ion irradiation (to be specific, proton irradiation) as an alternative of neutron irradiation to improve the current understanding of the mechanism. The objective of...

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

    International Nuclear Information System (INIS)

    Noh, Sanghoon; Kim, Tae Kyu

    2014-01-01

    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, 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

  17. Fatigue behaviour and crack growth of ferritic steel under environmental conditions

    International Nuclear Information System (INIS)

    Herter, K.H.; Schuler, X.; Weissenberg, T.

    2012-01-01

    The assessment of fatigue and cyclic crack growth behaviour of safety relevant components is of importance for the ageing management with regard to safety and reliability. For cyclic stress evaluation different codes and standards provide fatigue analysis procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. For the fatigue design curves used as a limiting criteria the influence of different factors like e.g. environment, surface finish and temperature must be taken into consideration in an appropriate way. Fatigue tests were performed in the low cycle fatigue (LCF) und high cycle fatigue (HCF) regime with low alloy steels as well as with Nb- and Ti-stabilized German austenitic stainless steels in air and high temperature (HT) boiling water reactor environment to extend the state of knowledge of environmentally assisted fatigue (EAF) as it can occur in boiling water reactor (BWR) plants. Using the reactor pressure vessel (RPV) steel 22NiMoCr3-7 experimental data were developed to verify the influence of BWR coolant environment (high purity water as well as sulphate containing water with 90 ppb SO 4 at a test temperature of 240 C and an oxygen content of 400 ppb) on the fatigue life and to extend the basis for a reliable estimation of the remaining service life of reactor components. Corresponding experiments in air were performed to establish reference data to determine the environmental correction factor F en accounting for the environment. The experimental results are compared with international available mean data curves, the new design curves and on the basis of the environmental factor F en . Furthermore the behaviour of steel 22NiMoCr3-7 in oxygenated high temperature water under transient loading conditions was investigated with respect to crack initiation and cyclic crack growth. In this process the stress state of the specimen and the chemical composition of the high

  18. Effect of microstructure on the oxidation behaviour of a low alloy ferritic steel

    International Nuclear Information System (INIS)

    Raman, R.K.S.; Khanna, A.S.; Gnanamoorthy, J.B.

    1991-01-01

    The influence of various metallurgical parameters such as grain size, cold working and heat-treatment on the oxidation behaviour of 2 1/4Cr-1Mo steel has been studied. Optical microscopy and other electron optical techniques such as SEM/EDX, EPMA and XRD have been used to study the morphology and composition of the scales as well as the scale/metal interface. Preferential oxidation along grain boundaries, changes in the range of oxidation as a result of prior cold working, and the role of secondary phases due to various heat-treatments on the oxidation behaviour have been investigated. (author)

  19. Microstructural investigation, using small-angle neutron scattering (SANS), of Optifer steel after low dose neutron irradiation and subsequent high temperature tempering

    International Nuclear Information System (INIS)

    Coppola, R.; Lindau, R.; Magnani, M.; May, R.P.; Moeslang, A.; Valli, M.

    2007-01-01

    The microstructural effect of low dose neutron irradiation and subsequent high temperature tempering in the reduced activation ferritic/martensitic steel Optifer (9.3 Cr, 0.1 C, 0.50 Mn, 0.26 V, 0.96 W, 0.66 Ta, Fe bal wt%) has been studied using small-angle neutron scattering (SANS). The investigated Optifer samples had been neutron irradiated, at 250 o C, to dose levels of 0.8 dpa and 2.4 dpa. Some of them underwent 2 h tempering at 770 o C after the irradiation. The SANS measurements were carried out at the D22 instrument of the High Flux Reactor at the Institut Max von Laue - Paul Langevin, Grenoble, France. The differences observed in nuclear and magnetic SANS cross-sections after subtraction of the reference sample from the irradiated one suggest that the irradiation and the subsequent post-irradiation tempering produce the growth of non-magnetic defects, tentatively identified as microvoids

  20. Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

    Science.gov (United States)

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

    2017-11-01

    The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

  1. Experimental and Computational Investigation of Structural Integrity of Dissimilar Metal Weld Between Ferritic and Austenitic Steel

    Science.gov (United States)

    Santosh, R.; Das, G.; Kumar, S.; Singh, P. K.; Ghosh, M.

    2018-03-01

    The structural integrity of dissimilar metal welded (DMW) joint consisting of low-alloy steel and 304LN austenitic stainless steel was examined by evaluating mechanical properties and metallurgical characteristics. INCONEL 82 and 182 were used as buttering and filler materials, respectively. Experimental findings were substantiated through thermomechanical simulation of the weld. During simulation, the effect of thermal state and stress distribution was pondered based on the real-time nuclear power plant environment. The simulation results were co-related with mechanical and microstructural characteristics. Material properties were varied significantly at different fusion boundaries across the weld line and associated with complex microstructure. During in-situ deformation testing in a scanning electron microscope, failure occurred through the buttering material. This indicated that microstructure and material properties synergistically contributed to altering the strength of DMW joints. Simulation results also depicted that the stress was maximum within the buttering material and made its weakest zone across the welded joint during service exposure. Various factors for the failure of dissimilar metal weld were analyzed. It was found that the use of IN 82 alloy as the buttering material provided a significant improvement in the joint strength and became a promising material for the fabrication of DMW joint.

  2. Effect of nano-size oxide particle dispersion and δ-ferrite proportion on creep strength of 9Cr-ODS steel

    International Nuclear Information System (INIS)

    Ohtsuka, Satoshi; Kaito, Takeji; Kim, Sawoong; Inoue, Masaki; Asayama, Tai; Ohnuma, Masato; Suzuki, Junichi

    2009-01-01

    The effects of chemical compositions on the microstructure and high-temperature creep strength of 9Cr-ODS steel was discussed in the light of quantitative data of δ-ferrite proportion and nano-size oxide particle dispersion, which were evaluated by dilatometric analysis and small angle neutron/X-ray scattering analysis, respectively. These quantitative data are well consistent with the conventional data obtained by transmission electron microscope. Both data indicate that the important microstructural feature for creep strength improvement of the 9Cr-ODS steel is the number density of nano-size oxide particles, and ferrite/martensite (F/M) duplex structure is favorable for high population nano-size oxide particle dispersion. Both optimization of excess oxygen concentration and control of the F/M duplex structure are promising technique for nano-structure control of 9Cr-ODS steel. Tungsten solid solution strengthening appears to be small compared with oxide dispersion strengthening enhanced by duplex microstructure formation. (author)

  3. Cyclic oxidation of stainless steel ferritic AISI 409, AISI 439 and AISI 441; Oxidacao ciclica dos acos inoxidaveis ferriticos AISI 409, AISI 439 e AISI 441

    Energy Technology Data Exchange (ETDEWEB)

    Salgado, Maria de Fatima; Santos, Diego Machado dos; Oliveira, Givanilson Brito de, E-mail: fatima.salgado@pq.cnpq.br [Universidade Estadual do Maranhao (CESC/UEMA), Caxias, MA (Brazil). Centro de Estudos Superiores; Rodrigues, Samara Clotildes Saraiva; Brandim, Ayrton de Sa [Instituto Federal do Piaui (PPGEM/IFPI), PI (Brazil); Lins, Vanessa de Freitas Cunha [Universidade Federal de Minas Gerais (IFMG), MG (Brazil)

    2014-07-01

    Stainless steels have many industrial applications. The cyclic oxidation of ferritic stainless steels technical and scientific importance presents, because they are less susceptible to peeling the austenitic alloys. For the purpose of investigating the behavior of these steels under thermal cycling, cyclic oxidation of AISI 409, AISI 441 and AISI 439 was carried out in a tubular furnace under two different conditions: oxidation by dipping the steel in the synthetic condensate for 10h and without oxidation immersion in the condensate, for up to 1500h at 300° C temperature. Using techniques: SEM, EDS and XRD revealed a microstructure with increased oxidation in the samples were immersed in the condensate. The oxide film remained intact during oxidation for steels 439 and 441 409 The Steel immersed in the condensate was rupture of the film after the 20th cycle of oxidation. The chemical characterization of the films allowed the identification of elements: Chromium, Iron, Aluminium and Silicon To a great extent, Cr{sub 2}O{sub 3}. (author)

  4. Evolution of microstructure at hot band annealing of ferritic FeSi steels

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Jürgen, E-mail: juergen.schneider@t-online.de [Institute of Metal Forming, Technische Universität Bergakademie Freiberg, Bernhard-von Cotta-Str. 4, D-09596 Freiberg (Germany); Stahlzentrum Freiberg e.V., Leipziger Straße 34, D-09599 Freiberg (Germany); Li, Guangqiang [State Key Lab. of Refractories and Metallurgy, Wuhan University of Science and Technology, No. 947 Heping Avenue, Qingshan District, Wuhan 430081 (China); Franke, Armin [Stahlzentrum Freiberg e.V., Leipziger Straße 34, D-09599 Freiberg (Germany); Zhou, Bowen [State Key Lab. of Refractories and Metallurgy, Wuhan University of Science and Technology, No. 947 Heping Avenue, Qingshan District, Wuhan 430081 (China)

    2017-02-15

    The magnetic properties of the finally fabricated nonoriented FeSi steels critically depend on the microstructure and on the occurring crystallographic texture. The fabrication route comprises hot rolling, coiling and cooling, hot band annealing before cold rolling (optional), cold rolling and the final thermal treatment. As well known there is an interplay between the microstructure and texture during the various processing steps. For that reason, it is of interest to know more on the evolution of the microstructure at hot band annealing of hot band prepared in different ways. In this paper we will summarize our recent results on the evolution of microstructure during thermal annealing of hot band: thermal treatment following immediately the last pass of hot rolling or a hot band annealing as a separate processing step before cold rolling.

  5. Improved oxidation resistance of ferritic steels with LSM coating for high temperature electrochemical applications

    DEFF Research Database (Denmark)

    Palcut, Marián; Mikkelsen, Lars; Neufeld, Kai

    2012-01-01

    The effect of single layer La0.85Sr0.15MnO3−δ (LSM) coatings on high temperature oxidation behaviour of four commercial chromia-forming steels, Crofer 22 APU, Crofer 22 H, E-Brite and AL 29-4C, is studied. The samples were oxidized for 140–1000 h at 1123 K in flowing simulated ambient air (air + 1......% H2O) and oxygen and corrosion kinetics monitored by mass increase of the materials over time. The oxide scale microstructure and chemical composition are investigated by scanning electron microscopy/energy-dispersive spectroscopy. The kinetic data obey a parabolic rate law. The results show...... that the LSM coating acts as an oxygen transport barrier that can significantly reduce the corrosion rate....

  6. Characterization of nano-sized oxides in Fe-12Cr oxide-dispersion-strengthened ferritic steel using small-angle neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Han, Young-Soo; Mao, Xiaodong; Jang, Jinsung; Kim, Tae-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-04-01

    The ferritic ODS steel was manufactured by hot isostatic pressing and heat treatment. The nano-sized microstructures such as yttrium oxides and Cr oxides were quantitatively analyzed by small-angle neutron scattering (SANS). The effects of the fabrication conditions on the nano-sized microstructure were investigated in relation to the quantitative analysis results obtained by SANS. The ratio between magnetic and nuclear scattering components was calculated, and the characteristics of the nano-sized yttrium oxides are discussed based on the SANS analysis results. (orig.)

  7. Characterization of nano-sized oxides in Fe-12Cr oxide-dispersion-strengthened ferritic steel using small-angle neutron scattering

    International Nuclear Information System (INIS)

    Han, Young-Soo; Mao, Xiaodong; Jang, Jinsung; Kim, Tae-Kyu

    2015-01-01

    The ferritic ODS steel was manufactured by hot isostatic pressing and heat treatment. The nano-sized microstructures such as yttrium oxides and Cr oxides were quantitatively analyzed by small-angle neutron scattering (SANS). The effects of the fabrication conditions on the nano-sized microstructure were investigated in relation to the quantitative analysis results obtained by SANS. The ratio between magnetic and nuclear scattering components was calculated, and the characteristics of the nano-sized yttrium oxides are discussed based on the SANS analysis results. (orig.)

  8. Irradiation Effects in Fortiweld Steel Containing Different Boron Isotopes

    International Nuclear Information System (INIS)

    Grounes, M.

    1967-07-01

    Tensile specimens and miniature impact specimens of the low alloyed pressure vessel steel Fortiweld have been irradiated at 265 deg C in R2 to two neutron doses, 6.5 x 10 18 n/cm 2 (> 1 MeV) and 4 x 10 19 n/cm 2 (thermal) and also 9.0 x 10 18 n/cm 2 (> 1 MeV) and 6 x 10 19 n/cm 2 (thermal). Material from three laboratory melts, in which the boron consisted of 10 B, 11 B and natural boron respectively, were investigated. The results both of tensile tests and impact tests with miniature impact specimens show that the 10 B-alloyed material was changed more and the 11 B-alloyed material was changed less than the material containing natural boron. At the higher neutron dose the increase in yield strength (0.2 % offset yield strength) was 11 kg/mm in the 10 B containing material compared to 5 kg/mm in the 11 B-containing material. The decrease in total elongation was 5 and 0 percentage units respectively. The transition temperature was increased 190 deg C at the higher neutron dose in the 10 B-alloyed material, 40 deg C in the 11 B-alloyed material and 80 deg C in the material containing natural boron

  9. Environmental Cracking and Irradiation Resistant Stainless Steels by Additive Manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B.

    2018-03-30

    Metal additive manufacturing (AM), or metal 3D printing is an emergent advanced manufacturing method that can create near net shape geometries directly from computer models. This technology can provide the capability to rapidly fabricate complex parts that may be required to enhance the integrity of reactor internals components. Such opportunities may be observed during a plant refueling outage and AM parts can be rapidly custom designed, manufactured and deployed within the outage interval. Additive manufacturing of stainless steel (SS) components can add business benefits on fast delivery on repair hardware, installation tooling, new design prototypes tests, etc. For the nuclear industry, the supply chain is always an issue for reactor service. AM can provide through-life supply chain (40-60 years) for high-value low-volume components. In the meantime, the capability of generating complex geometries and functional gradient materials will improve the performance, reduce the overall component cost, plant asset management cost and increase the plant reliability by the improvement in materials performance in nuclear environments. While extensive work has been conducted regarding additively manufacturing of austenitic SS parts, most efforts focused only on basic attributes such as porosity, residual stress, basic tensile properties, along with components yield and process monitoring. Little work has been done to define and evaluate the material requirements for nuclear applications. Technical gaps exist, which limit this technology adoption in the nuclear industry, which includes high manufacturing cost, unknown risks, limited nuclear related data, lack of specification and qualification methods, and no prior business experience. The main objective of this program was to generate research data to address all these technical gaps and establish a commercial practice to use AM technology in the nuclear power industry. The detailed objectives are listed as follows: (1

  10. Effect of proton irradiation on the magnetic properties of manganese ferrite.

    Science.gov (United States)

    Hyun, Sung Wook; Hong, Sun Chun; Kim, Sam Jin; Kim, Chul Sung

    2011-07-01

    Cubic-spinel MnFe2O4 magnetic nanoparticles (NPs) were prepared, with an average particle size of about 4 nm determined from a high-resolution transmission electron microscope. When the NPs were proton-irradiated, the lattice constants decreased with increasing proton irradiation. Before the proton irradiation, the NPs exhibited 36.2 +/- 0.1 emu/g magnetization (M(S)) and 11.1 +/- 0.1 Oe coercivity (H(C)). After the irradiation of the samples with 5 and 10 pC/microm2 doses, the M(S) changed to 35.6 and 35.1 +/- 0.1 emu/g, and the H(C) to 11.3 and 12.9 +/- 0.1 Oe, respectively. The room-temperature Mössbauer spectra of the NPs showed superparamagnetic characteristics, with the single-absorption line of two sites and a large relaxation frequency. During the proton irradiation, the relaxation frequency decreased to 156.02 and 134.29 +/- 0.01 Gamma/ħ from the unirradiated sample's 164.02 +/- 0.01 Gamma/ħ. It is suggested that the proton irradiation induced the increase in the anisotropy energy of the MnFe2O4 NPs. Moreover, from the external-field-induced Mössbauer spectra at 4.2 K, an increase in the canted angle of the hyperfine field between sites A (tetrahedral) and B (octahedral) was observed with proton irradiation.

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

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

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

  12. First-principles study on influence of molybdenum on acicular ferrite formation on TiC particles in microallyed steels

    Science.gov (United States)

    Hua, Guomin; Li, Changsheng; Cheng, Xiaonong; Zhao, Xinluo; Feng, Quan; Li, Zhijie; Li, Dongyang; Szpunar, Jerzy A.

    2018-01-01

    In this study, influences of molybdenum on acicular ferrite formation on precipitated TiC particles are investigated from thermodynamic and kinetic respects. In thermodynamics, Segregation of Mo towards austenite/TiC interface releases the interfacial energy and induces phase transformation from austenite to acicular ferrite on the precipitated TiC particles. The Phase transformation can be achieved by displacive deformation along uniaxial Bain path. In addition, the segregation of Mo atom will also lead to the enhanced stability of ferrite in comparison with austenite no matter at low temperature or at high temperature. In kinetics, the Mo solute in acicular ferrite can effectively suppress the diffusion of carbon atoms, which ensures that orientation relationship between acicular ferrite and austenitized matrix can be satisfied during the diffusionless phase transformation. In contrast to ineffectiveness of TiC particles, the alloying Mo element can facilitate the formation of acicular ferrite on precipitated TiC particles, which is attributed to the above thermodynamic and kinetic reasons. Furthermore, Interfacial toughness and ductility of as-formed acicular ferrite/TiC interface can be improved simultaneously by segregation of Mo atom.

  13. Effects of neutron irradiation on resistivity of reactor pressure vessel steel

    Science.gov (United States)

    Li, Chengliang; Shu, Guogang; Liu, Yi; Huang, Yili; Chen, Jun; Duan, Yuangang; Liu, Wei

    2018-02-01

    The embrittlement of reactor pressure vessel (RPV) steel owing to fast-neutron irradiation is one of its primary failure mechanisms. In this work, neutron irradiation tests were performed on an RPV steel at a high temperature (565 K) using a neutron irradiation test reactor. In addition, resistivity measurements were performed on the RPV steel both before and after irradiation in a hot laboratory using the four-probe method. The results showed that the resistivity of the RPV steel exhibits nonlinear behaviour with respect to the radiation fluence and that the nonlinearity becomes more pronounced with an increase in the radiation fluence. For instance, when the radiation fluence is 0.1540 dpa and the excitation current is increased from 0.2 mA to 200 mA, the resistivity of the RPV steel decreases by as much as 67.12%. During irradiation embrittlement, the resistivity increases with the fluence. When the radiation fluence is greater than 0.116 dpa, the increase in the resistivity accelerates. When the radiation fluence is less than 0.116 dpa and when an excitation current of 2 mA or 20 mA is used, the relationship between the resistivity and the radiation fluence for the RPV steel is a quadratic one, whereas that between the rate of change in the resistivity and the radiation fluence is a linear one. Thus, the resistivity of RPV steel can be used to characterise its degree of irradiation embrittlement, and resistivity measurements can be employed as a nondestructive evaluation technique for monitoring the degree of irradiation damage experienced by in-service RPV steel.

  14. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  15. Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

    Energy Technology Data Exchange (ETDEWEB)

    Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

    2015-12-15

    Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

  16. Effect of Heat Input on Microstructure Evolution and Mechanical Properties in the Weld Heat-Affected Zone of 9Cr-2W-VTa Reduced Activation Ferritic-Martensitic Steel for Fusion Reactor

    Science.gov (United States)

    Moon, Joonoh; Lee, Chang-Hoon; Lee, Tae-Ho; Kim, Hyoung Chan

    2015-01-01

    The phase transformation and mechanical properties in the weld heat-affected zone (HAZ) of a reduced activation ferritic/martensitic steel were explored. The samples for HAZs were prepared using a Gleeble simulator at different heat inputs. The base steel consisted of tempered martensite and carbides through quenching and tempering treatment, whereas the HAZs consisted of martensite, δ-ferrite, and a small volume of autotempered martensite. The prior austenite grain size, lath width of martensite, and δ-ferrite fraction in the HAZs increased with increase in the heat input. The mechanical properties were evaluated using Vickers hardness and Charpy V-notch impact test. The Vickers hardness in the HAZs was higher than that in the base steel but did not change noticeably with increase in the heat input. The HAZs showed poor impact property due to the formation of martensite and δ-ferrite as compared to the base steel. In addition, the impact property of the HAZs deteriorated more with the increase in the heat input. Post weld heat treatment contributed to improve the impact property of the HAZs through the formation of tempered martensite, but the impact property of the HAZs remained lower than that of base steel.

  17. Comparative small angle neutron scattering (SANS study of Eurofer97 steel neutron irradiated in mixed (HFR and fast spectra (BOR60 reactors

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2016-12-01

    Full Text Available This contribution presents a comparative microstructural investigation, carried out by Small-Angle Neutron Scattering (SANS, of ferritic/martensitic steel Eurofer97 (0.12 C, 9 Cr, 0.2V, 1.08Wwt% neutron irradiated at two different neutron sources, the HFR-Petten (SPICE experiment and the BOR60 reactor (ARBOR experiment. The investigated “SPICE” sample had been irradiated to 16dpa at 250°C, the investigated “ARBOR” one had been irradiated to 32dpa at 330°C. The SANS measurements were carried under a 1 T magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated Eurofer sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. The detected increase in the respective SANS cross-sections of these two samples under irradiation is attributed primarily to the presence of micro-voids, for neutron contrast reasons; it is quite similar in the two samples, despite the higher irradiation dose and temperature of the “ARBOR” sample with respect to the “SPICE” one. This is tentatively correlated with the higher helium content produced under HFR irradiation, playing an important role to stabilize the micro-voids under irradiation. In fact, the size distributions obtained by transformation of the SANS data yield a micro-void volume fraction of 1.3% for the “SPICE” sample and of 0.6% for the “ARBOR” one.

  18. Brittle and ductile rupture of 16MND5 steel. Irradiation effect

    International Nuclear Information System (INIS)

    Al Mundheri, M.; Soulat, P.; Pineau, A.

    1986-06-01

    Toughness tests have been made on 16MND5 steel (A508Cl3 steel) - before and after irradiation at 290 0 C (3.10 19 n/cm 2 , E > 1 MeV). It is shown that toughness is lowered following the irradiation and that it is a decreasing function of the thickness of the test pieces. In parallel, tests on three geometries of entailed specimens, prepared in the non-irradiated material, have been made at different temperatures to apply the methodology of local approach of ductile-brittle rupture [fr

  19. Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

    International Nuclear Information System (INIS)

    Klaar, H.J.

    1975-01-01

    As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    The remote-controlled equipment for measurement of magnetic flux density has been developed in order to evaluate the irradiation damage of austenitic stainless steels. Magnetic flux densities by neutron irradiation in austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), have been measured by the equipment. The results show that irradiation damage affected to magnetic flux density, and indicate the measuring method of magnetic flux density using a small magnetizer with a permanent magnet of 2 mm in diameter is less affected by specimen shape. (author)

  1. Microstructural evolution of reduced-activation martensitic steel under single and sequential ion irradiations

    International Nuclear Information System (INIS)

    Luo, Fengfeng; Guo, Liping; Jin, Shuoxue; Li, Tiecheng; Zheng, Zhongcheng; Yang, Feng; Xiong, Xuesong; Suo, Jinping

    2013-01-01

    Microstructural evolution of super-clean reduced-activation martensitic steels irradiated with single-beam (Fe + ) and sequential-beam (Fe + plus He + ) at 350 °C and 550 °C was studied. Sequential-beam irradiation induced smaller size and larger number density of precipitates compared to single-beam irradiation at 350 °C. The largest size of cavities was observed after sequential-beam irradiation at 550 °C. The segregation of Cr and W and depletion of Fe in carbides were observed, and the maximum depletion of Fe and enrichment of Cr occurred under irradiation at 350 °C

  2. Ferrite re-crystallization kinetics on a C-Mn steel and on two micro alloyed steels after dual-phase strain; Cinetica de recristalizacao da ferrita em um aco C-Mn e dois acos microligados apos deformacao na regiao bifasica

    Energy Technology Data Exchange (ETDEWEB)

    Simieli, Eider A. [Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil)

    1991-12-31

    Ferrite recrystallization was investigated in two micro alloyed steels deformed in the inter critical range. A reference steel was also used, which had a composition of 0,06% C and 1,31% Mn. (author). 15 refs., 7 figs., 3 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  4. Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [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; Li, Tiecheng; Zheng, Zhongcheng [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); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-12-15

    Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100 keV helium ions to 0.8 dpa between 300 K and 723 K. At irradiation temperatures T{sub irr} ⩽ 573 K, small defects with both Burger vectors b = 1/2〈1 1 1〉 and b = 〈1 0 0〉 were observed, while at T{sub irr} ⩾ 623 K, the microstructure was dominated by large convoluted interstitial dislocation loops with b = 〈1 0 0〉. Only small cavities were found in the steels irradiated at 723 K.

  5. Processing of a novel nano-structured ferritic steel via spark plasma sintering and investigation of its mechanical and microstructural characteristics

    International Nuclear Information System (INIS)

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

    2015-01-01

    Nano-structured ferritic steels (NFSs) with 12-14 wt% Cr have attracted widespread interest for potential high temperature structural and fuel cladding applications in advanced nuclear reactors. They have excellent high temperature mechanical properties and high resistance to radiation-induced damage. The properties of the NFSs depend on the composition that mainly consists of Cr, Ti, W or Mo, and Y 2 O 3 as alloying constituents. In this study, a novel nano-structured ferritic steel (Fe-14Cr-1Ti-0.3Mo-0.5La 2 O 3 , wt%) termed as 14LMT was developed via high energy ball milling and spark plasma sintering. Vickers microhardness values were measured. Microstructural studies of the developed NFSs were performed by EBSD and TEM, which revealed a bimodal grain size distribution. A significant number density of nano-precipitates was observed in the microstructure. The diameter of the precipitates varied between 2-70 nm and the morphology from the spherical to faceted shape. The Cr-La-Ti-O-enriched nano-clusters were identified by APT studies. (authors)

  6. High-temperature creep rupture of low alloy ferritic steel butt-welded pipes subjected to combined internal pressure and end loadings.

    Science.gov (United States)

    Vakili-Tahami, F; Hayhurst, D R; Wong, M T

    2005-11-15

    Constitutive equations are reviewed and presented for low alloy ferritic steels which undergo creep deformation and damage at high temperatures; and, a thermodynamic framework is provided for the deformation rate potentials used in the equations. Finite element continuum damage mechanics studies have been carried out using these constitutive equations on butt-welded low alloy ferritic steel pipes subjected to combined internal pressure and axial loads at 590 and 620 degrees C. Two dominant modes of failure have been identified: firstly, fusion boundary failure at high stresses; and, secondly, Type IV failure at low stresses. The stress level at which the switch in failure mechanism takes place has been found to be associated with the relative creep resistance and lifetimes, over a wide range of uniaxial stresses, for parent, heat affected zone, Type IV and weld materials. The equi-biaxial stress loading condition (mean diameter stress equal to the axial stress) has been confirmed to be the worst loading condition. For this condition, simple design formulae are proposed for both 590 and 620 degrees C.

  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

    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.

  8. 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.

  9. Comparison of swelling for structural materials on neutron and ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.

    1986-03-01

    The swelling of V-base alloys, Type 316 stainless steel, Fe-25Ni-15Cr alloys, ferritic steels, Cu, Ni, Nb-1% Zr, and Mo on neutron irradiation is compared with the swelling for these materials on ion irradiation. The results of this comparison show that utilization of the ion-irradiation technique provides for a discriminative assessment of the potential for swelling of candidate materials for fusion reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

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

  11. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

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

    International Nuclear Information System (INIS)

    1981-04-01

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

  13. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improve