WorldWideScience

Sample records for high-temperature irradiation behavior

  1. The irradiation behavior of atomized U-Mo alloy fuels at high temperature

    Science.gov (United States)

    Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

    2001-04-01

    Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

  2. High temperature deformation behavior, thermal stability and irradiation performance in Grade 92 steel

    Science.gov (United States)

    Alsagabi, Sultan

    The 9Cr-2W ferritic-martensitic steel (i.e. Grade 92 steel) possesses excellent mechanical and thermophysical properties; therefore, it has been considered to suit more challenging applications where high temperature strength and creep-rupture properties are required. The high temperature deformation mechanism was investigated through a set of tensile testing at elevated temperatures. Hence, the threshold stress concept was applied to elucidate the operating high temperature deformation mechanism. It was identified as the high temperature climb of edge dislocations due to the particle-dislocation interactions and the appropriate constitutive equation was developed. In addition, the microstructural evolution at room and elevated temperatures was investigated. For instance, the microstructural evolution under loading was more pronounced and carbide precipitation showed more coarsening tendency. The growth of these carbide precipitates, by removing W and Mo from matrix, significantly deteriorates the solid solution strengthening. The MX type carbonitrides exhibited better coarsening resistance. To better understand the thermal microstructural stability, long tempering schedules up to 1000 hours was conducted at 560, 660 and 760°C after normalizing the steel. Still, the coarsening rate of M23C 6 carbides was higher than the MX-type particles. Moreover, the Laves phase particles were detected after tempering the steel for long periods before they dissolve back into the matrix at high temperature (i.e. 720°C). The influence of the tempering temperature and time was studied for Grade 92 steel via Hollomon-Jaffe parameter. Finally, the irradiation performance of Grade 92 steel was evaluated to examine the feasibility of its eventual reactor use. To that end, Grade 92 steel was irradiated with iron (Fe2+) ions to 10, 50 and 100 dpa at 30 and 500°C. Overall, the irradiated samples showed some irradiation-induced hardening which was more noticeable at 30°C. Additionally

  3. Release behavior of fission products from irradiated dispersion fuels at high temperatures

    International Nuclear Information System (INIS)

    Iwai, Takashi; Shimizu, Michio; Nakagawa, Tetsuya

    1990-02-01

    As a framework of reduced enrichment fuel program of JMTR Project, the measurements of fission products release rates at high temperatures (600degC - 1100degC) were performed in order to take the data to use for safety evaluation of LEU fuel. Three type miniplates of dispersion silicide and aluminide fuel, 20% enrichment LEU fuel with 4.8 gU/cc (U 3 Si 2 90 %, USi 10 % and U 3 Si 2 50 %, U 3 Si 50 % dispersed in aluminium) and 45 % enrichment MEU fuel with 1.6 gU/cc, were irradiated in JMTR. The burnups attained by one cycle (22 days) irradiation were within 21.6 % - 22.5 % of initial 235 U. The specimens cut down from miniplates were measured on fission products release rates by means of new apparatus specially designed for this experiment. The specimens were heated up within 600degC - 1100degC in dry air. Then fission products such as 85 Kr, 133 Xe, 131 I, 137 Cs, 103 Ru, 129m Te were collected at each temperature and measured on release rates. In the results of measurement, the release rates of 85 Kr, 133 Xe, 131 I, 129m Te from all specimens were slightly less than that of G.W. Parker's data on U-Al alloy fuel. For 137 Cs and 103 Ru from a silicide specimen (U 3 Si 2 90 %, USi 10 % dispersed in aluminium) and 137 Cs from an aluminide specimen, the release rates were slightly higher than that of G.W. Parker's. (author)

  4. Crack initiation behavior of neutron irradiated model and commercial stainless steels in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

    Highlights: • Environmental constant extension rate tensile tests were performed on neutron irradiated steel. • Percentage of intergranular cracking quantified the cracking susceptibility. • Cracking susceptibility varied with test environment, solute addition, and cold work. • No singular microstructural change could explain increases in cracking susceptibility with irradiation dose. • The increment of yield strength due to irradiation correlated well with cracking susceptibility. -- Abstract: The objective of this study was to isolate key factors affecting the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of eleven neutron-irradiated austenitic stainless steel alloys. Four commercial purity and seven high purity stainless steels were fabricated with specific changes in composition and microstructure, and irradiated in a fast reactor spectrum at 320 °C to doses between 4.4 and 47.5 dpa. Constant extension rate tensile (CERT) tests were performed in normal water chemistry (NWC), hydrogen water chemistry (HWC), or primary water (PW) environments to isolate the effects of environment, elemental solute addition, alloy purity, alloy heat, alloy type, cold work, and irradiation dose. The irradiated alloys showed a wide variation in IASCC susceptibility, as measured by the relative changes in mechanical properties and crack morphology. Cracking susceptibility measured by %IG was enhanced in oxidizing environments, although testing in the lowest potential environment caused an increase in surface crack density. Alloys containing solute addition of Ni or Ni + Cr exhibited no IASCC. Susceptibility was reduced in materials cold worked prior to irradiation, and increased with increasing irradiation dose. Irradiation-induced hardening was accounted for by the dislocation loop microstructure, however no relation between crack initiation and radiation hardening was found.

  5. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D. [Laboratoire PROMES-CNRS, Laboratoire Procedes, Materiaux et Energie Solaire, 7 rue du Four Solaire, 66120 Font-Romeu Odeillo (France)

    2007-07-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H{sup +} ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  6. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    International Nuclear Information System (INIS)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D.

    2007-01-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H + ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  7. HTGR fuel behavior at very high temperature

    International Nuclear Information System (INIS)

    Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

    1986-03-01

    Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

  8. Zircaloy behaviour in high temperature irradiated water

    International Nuclear Information System (INIS)

    Urbanic, V.F.

    1982-04-01

    The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

  9. The behavior of lattice defects produced in Al2O3 irradiated by neutrons at high temperatures

    International Nuclear Information System (INIS)

    Atobe, K.; Koizumi, T.; Okada, M.

    2003-01-01

    Single crystals of α-Al 2 O 3 were irradiated by the two reactors, KUR and JMTR, at three different temperatures. Lattice defects produced by irradiation were studied by esr (electron spin resonance). Three kinds of esr spectram, which are denoted as A, B and C spectram, are observed. The spectram A was observed at three different irradiation temperatures and was ascribed to oxygen vacancies. The spectram B showed no angular dependence for the rotation of external magnetic field to the crystal axis, and the defect density of this spectram decreased with an increase of annealing temperature. When the specimen was annealed at 400 degC after irradiation at 200 degC, the spectram C was observed and was presumed to be due to Al-colloids. (Y. Kazumata)

  10. High temperature oxidation behavior of ODS steels

    Science.gov (United States)

    Kaito, T.; Narita, T.; Ukai, S.; Matsuda, Y.

    2004-08-01

    Oxide dispersion strengthened (ODS) steels are being developing for application as advanced fast reactor cladding and fusion blanket materials, in order to allow increased operation temperature. Oxidation testing of ODS steel was conducted under a controlled dry air atmosphere to evaluate the high temperature oxidation behavior. This showed that 9Cr-ODS martensitic steels and 12Cr-ODS ferritic steels have superior high temperature oxidation resistance compared to 11 mass% Cr PNC-FMS and 17 mass% Cr ferritic stainless steel. This high temperature resistance is attributed to earlier formation of the protective α-Cr 2O 3 on the outer surface of ODS steels.

  11. Irradiation test HT-31: high-temperature irradiation behavior of LASL-made extruded fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; Davidson, K.V.; Schell, D.H.

    1977-04-01

    Three LASL-made extruded graphite and coated particle fuel rods have been irradiated in the Oak Ridge National Laboratory High Fluence Isotope Reactor test HT-31. Test conditions were about 9 x 10 21 nvt(E > .18 MeV) at 1250 0 C. The graphite matrix showed little or no effect of the irradiation. LASL-made ZrC containing coated particles with ZrC coats and ZrC-doped pyrolytic carbon coats showed no observable effects of the irradiation

  12. Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles

    International Nuclear Information System (INIS)

    Wagner, P.; Reiswig, R.D.; Hollabaugh, C.M.; White, R.W.; O'Rourke, J.A.; Davidson, K.V.; Schell, D.H.

    1977-10-01

    Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10 21 nvt (E greater than 0.18 MeV) at 1350 0 C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particles with inert kernels showed no irradiation effects. The graded ZrC-C coats on the fissile particles were cracked. It is postulated that the cracking is associated with the low LTI deposition rate and is not related to the ZrC

  13. Development of Environment and Irradiation Effects of High Temperature Materials

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Kim, D. W.; Kim, S. H.

    2009-11-01

    Proposed materials, Mod.9Cr-1Mo steel (32 mm thickness) and 9Cr-1Mo-1W (100 mm thickness), for the reactor vessel were procured, and welded by the qualified welding technologies. Welding soundness was conformed by NDT, and mechanical testings were done along to weld depth. Two new irradiation capsules for use in the OR test hole of HANARO were designed and fabricated. specimens was irradiated in the OR5 test hole of HANARO with a 30MW thermal power at 390±10 .deg. C up to a fast neutron fluence of 4.4x10 19 (n/cm 2 ) (E>1.0 MeV). The dpa was evaluated to be 0.034∼0.07. Base metals and weldments of both Mod.9Cr-1Mo and 9Cr-1Mo-1W steels were tested tensile and impact properties in order to evaluate the irradiation hardening effects due to neutron irradiation. DBTT of base metal and weldment of Mod.9Cr-1Mo steel were -16 .deg. C and 1 .deg. C, respectively. After neutron irradiation, DBTT of weldment of Mod.9Cr-1Mo steel increased to 25 . deg. C. Alloy 617 and several nickel-base superalloys were studied to evaluate high temperature degradation mechanisms. Helium loop was developed to evaluate the oxidation behaviors of materials in the VHTR environments. In addition, creep behaviors in air and He environments were compared, and oxidation layers formed outer surfaces were measured as a function of applied stress and these results were investigated to the creep life

  14. Long duration performance of high temperature irradiation resistant thermocouples

    International Nuclear Information System (INIS)

    Rempe, J.; Knudson, D.; Condie, K.; Cole, J.; Wilkins, S.C.

    2007-01-01

    Many advanced nuclear reactor designs require new fuel, cladding, and structural materials. Data are needed to characterize the performance of these new materials in high temperature, radiation conditions. However, traditional methods for measuring temperature in-pile degrade at temperatures above 1100 C degrees. To address this instrumentation need, the Idaho National Laboratory (INL) developed and evaluated the performance of a high temperature irradiation-resistant thermocouple that contains alloys of molybdenum and niobium. To verify the performance of INL's recommended thermocouple design, a series of high temperature (from 1200 to 1800 C) long duration (up to six months) tests has been initiated. This paper summarizes results from the tests that have been completed. Data are presented from 4000 hour tests conducted at 1200 and 1400 C that demonstrate the stability of this thermocouple (less than 2% drift). In addition, post test metallographic examinations are discussed which confirm the compatibility of thermocouple materials throughout these long duration, high temperature tests. (authors)

  15. High temperature superconductors for fusion magnets -influence of neutron irradiation

    International Nuclear Information System (INIS)

    Chudy, M.; Eisterer, M.; Weber, H. W.

    2010-01-01

    In this work authors present the results of study of influence of neutron irradiation of high temperature superconductors for fusion magnets. High temperature superconductors (type of YBCO (Yttrium-Barium-Copper-Oxygen)) are strong candidates to be applied in the next step of fusion devices. Defects induced by fast neutrons are effective pinning centres, which can significantly improve critical current densities and reduce J c anisotropy. Due to induced lattice disorder, T c is reduced. Requirements for ITER (DEMO) are partially achieved at 64 K.

  16. Irradiation effects of high temperature superconductor of lanthanoid oxides

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Koh-ichi; Kohara, Takao [Himeji Inst. of Tech., Hyogo (Japan)

    1996-04-01

    Neutron irradiation effects on excess oxygen were studied by neutron irradiation on La{sub 2}CuO{sub 4} treated with high pressure oxygen. La{sub 2}CuO{sub 4} was prepared by the usual method and annealed for 10 h under the oxygen pressure of 800-2000 atm. at 600degC. The superconducting transition temperature (Tc) is 27-32K before irradiation (La{sub 2}CuO{sub 4+d}, amount of excess oxygen d=0.03-0.12). Neutron irradiation was carried out by two kinds of experiments. Low irradiation dose test at low temperature (LTL: {approx}20-200K, storage in LN{sub 2}) showed Tc decreased more slowly than that of high temperature range. Experiment at high temperature (Hyd:{approx}80deg{yields}, storage at room temperature) showed -10K/10{sup 18}n/cm{sup 2}, the decrease of Tc was three times larger than that of YBCO type superconductor. (S.Y.)

  17. Is uranium dioxide a glass at high temperature: the reason for its irradiation resistance?

    International Nuclear Information System (INIS)

    Desgranges, Lionel

    2008-01-01

    Electronic intrinsic carriers are shown to have some influence on UO 2 high temperature properties. The physical nature of these carriers, called polarons, is discussed and it is proposed that they could correspond to quasi-broken bonds, in a similar way to intrinsic electronic defects in SiO 2 . It is shown that this hypothesis provides an explanation, at least qualitative, for UO 2 specific behavior at high temperature and under irradiation. (author)

  18. Structural behavior of reinforced concrete structures at high temperatures

    International Nuclear Information System (INIS)

    Yamazaki, N.; Yamazaki, M.; Mochida, T.; Mutoh, A.; Miyashita, T.; Ueda, M.; Hasegawa, T.; Sugiyama, K.; Hirakawa, K.; Kikuchi, R.; Hiramoto, M.; Saito, K.

    1995-01-01

    To establish a method to predict the behavior of reinforced concrete structures subjected simultaneously to high temperatures and external loads, this paper presents the results obtained in several series of tests carried out recently in Japan. This paper reports on the material properties of concrete and steel bars under high temperatures. It also considers the heat transfer properties of thick concrete walls under transient high temperatures, and the structural behavior of reinforced concrete beams subjected to high temperatures. In the tests, data up to 800 C were obtained for use in developing a computational method to estimate the non-linear behavior of reinforced concrete structures exposed to high temperatures. (orig.)

  19. LVDT Development for High Temperature Irradiation Test and Application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Ban, Chae Min; Choo, Kee Nam; Jun, Byung Hyuk [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The LVDT (Linear Variable Differential Transformer) is used to measure the elongation and pressure of a nuclear fuel rod, or the creep and fatigue of the material during a reactor irradiation test. This device must be a radiation-resistant LVDT for use in a research reactor. Norway Halden has LVDTs for an irradiation test by the own development and commercialized. But Halden's LVDTs have limited the temperature of the use until to 350 .deg. C. So, KAERI has been developing a new LVDT for high temperature irradiation test. This paper describes the design of a LVDT, the fabrication process of a LVDT, and the result of the performance test. The designed LVDT uses thermocouple cable for coil wire material and one MI cable as signal cable. This LVDT for a high temperature irradiation test can be used until a maximum of 900 .deg. C. Welding is a very important factor for the fabrication of an LVDT. We are using a 150W fiber laser welding system that consists of a welding head, monitoring vision system and rotary index.

  20. Microstructural Evolution and Mechanical Behavior of High Temperature Solders: Effects of High Temperature Aging

    Science.gov (United States)

    Hasnine, M.; Tolla, B.; Vahora, N.

    2018-04-01

    This paper explores the effects of aging on the mechanical behavior, microstructure evolution and IMC formation on different surface finishes of two high temperature solders, Sn-5 wt.% Ag and Sn-5 wt.% Sb. High temperature aging showed significant degradation of Sn-5 wt.% Ag solder hardness (34%) while aging has little effect on Sn-5 wt.% Sb solder. Sn-5 wt.% Ag experienced rapid grain growth as well as the coarsening of particles during aging. Sn-5 wt.% Sb showed a stable microstructure due to solid solution strengthening and the stable nature of SnSb precipitates. The increase of intermetallic compound (IMC) thickness during aging follows a parabolic relationship with time. Regression analysis (time exponent, n) indicated that IMC growth kinetics is controlled by a diffusion mechanism. The results have important implications in the selection of high temperature solders used in high temperature applications.

  1. The behavior of lattice defects produced in Al{sub 2}O{sub 3} irradiated by neutrons at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Atobe, K.; Koizumi, T. [Naruto Univ. of Education, Tokushima (Japan); Okada, M. [Kyoto Univ., Research Reactor Inst., Kumatori, Osaka (Japan)

    2003-01-01

    Single crystals of {alpha}-Al{sub 2}O{sub 3} were irradiated by the two reactors, KUR and JMTR, at three different temperatures. Lattice defects produced by irradiation were studied by esr (electron spin resonance). Three kinds of esr spectram, which are denoted as A, B and C spectram, are observed. The spectram A was observed at three different irradiation temperatures and was ascribed to oxygen vacancies. The spectram B showed no angular dependence for the rotation of external magnetic field to the crystal axis, and the defect density of this spectram decreased with an increase of annealing temperature. When the specimen was annealed at 400 degC after irradiation at 200 degC, the spectram C was observed and was presumed to be due to Al-colloids. (Y. Kazumata)

  2. High temperature oxidation behavior of TiAl-based intermetallics

    International Nuclear Information System (INIS)

    Stroosnijder, M.F.; Sunderkoetter, J.D.; Haanappel, V.A.C.

    1996-01-01

    TiAl-based intermetallic compounds have attracted considerable interest as structural materials for high-temperature applications due to their low density and substantial mechanical strength at high temperatures. However, one major drawback hindering industrial application arises from the insufficient oxidation resistance at temperatures beyond 700 C. In the present contribution some general aspects of high temperature oxidation of TiAl-based intermetallics will be presented. This will be followed by a discussion of the influence of alloying elements, in particular niobium, and of the effect of nitrogen in the oxidizing environment on the high temperature oxidation behavior of such materials

  3. Microstructure and hardness evolution of nanochannel W films irradiated by helium at high temperature

    Science.gov (United States)

    Qin, Wenjing; Wang, Yongqiang; Tang, Ming; Ren, Feng; Fu, Qiang; Cai, Guangxu; Dong, Lan; Hu, Lulu; Wei, Guo; Jiang, Changzhong

    2018-04-01

    Plasma facing materials (PFMs) face one of the most serious challenges in fusion reactors, including unprecedented harsh environment such as 14.1 MeV neutron and transmutation gas irradiation at high temperature. Tungsten (W) is considered to be one of the most promising PFM, however, virtually insolubility of helium (He) in W causes new material issues such as He bubbles and W "fuzz" microstructure. In our previous studies, we presented a new strategy using nanochannel structure designed in the W film to increase the releasing of He atoms and thus to minimize the He nucleation and "fuzz" formation behavior. In this work, we report the further study on the diffusion of He atoms in the nanochannel W films irradiated at a high temperature of 600 °C. More specifically, the temperature influences on the formation and growth of He bubbles, the lattice swelling, and the mechanical properties of the nanochannel W films were investigated. Compared with the bulk W, the nanochannel W films possessed smaller bubble size and lower bubble areal density, indicating that noticeable amounts of He atoms have been released out along the nanochannels during the high temperature irradiations. Thus, with lower He concentration in the nanochannel W films, the formation of the bubble superlattice is delayed, which suppresses the lattice swelling and reduces hardening. These aspects indicate the nanochannel W films have better radiation resistance even at high temperature irradiations.

  4. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  5. Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Wang, Lumin; Was, Gary

    2010-01-01

    Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.

  6. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  7. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  8. 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.

  9. Modeling high temperature materials behavior for structural analysis

    CERN Document Server

    Naumenko, Konstantin

    2016-01-01

    This monograph presents approaches to characterize inelastic behavior of materials and structures at high temperature. Starting from experimental observations, it discusses basic features of inelastic phenomena including creep, plasticity, relaxation, low cycle and thermal fatigue. The authors formulate constitutive equations to describe the inelastic response for the given states of stress and microstructure. They introduce evolution equations to capture hardening, recovery, softening, ageing and damage processes. Principles of continuum mechanics and thermodynamics are presented to provide a framework for the modeling materials behavior with the aim of structural analysis of high-temperature engineering components.

  10. Corrosion behavior of low energy, high temperature nitrogen ion ...

    Indian Academy of Sciences (India)

    Corrosion behavior of low energy, high temperature nitrogen ion-implanted AISI 304 stainless steel. M GHORANNEVISS1, A SHOKOUHY1,∗, M M LARIJANI1,2,. S H HAJI HOSSEINI 1, M YARI1, A ANVARI4, M GHOLIPUR SHAHRAKI1,3,. A H SARI1 and M R HANTEHZADEH1. 1Plasma Physics Research Center, Science ...

  11. Experimental study of pyrolytic boron nitride at high temperature with and without proton and VUV irradiations

    International Nuclear Information System (INIS)

    Balat-Pichelin, M.; Eck, J.; Heurtault, S.; Glénat, H.

    2014-01-01

    Highlights: • New results for the high temperature study of pBN in high vacuum for the heat shield of solar probes. • Physico-chemical behavior of pBN studied up to 1700 K with proton and VUV irradiations. • Rather low effect of synergistic aggressions on the microstructure of pBN material. • The α/ε ratio of pBN coating on C/C measured up to 2200 K is 20% lower than for the C/C itself. - Abstract: In the frame of future exploration missions such as Solar Probe Plus (NASA) and PHOIBOS (ESA), research was carried out to study pyrolytic BN material envisaged as coating for their heat shields. The physico-chemical behavior of CVD pBN at very high temperature with or without hydrogen ions and VUV (Vacuum Ultra-Violet) irradiations was studied in high vacuum together with the in situ measurement of the thermal radiative properties conditioning the thermal equilibrium of the heat shield. Experimental results obtained on massive pBN samples are presented through in situ mass spectrometry and mass loss rate, and post-test microstructural characterization by XRD, SEM, AFM and nano-indentation techniques, some of them leading to mechanical properties. It could be concluded that synergistic effect of high temperature, protons and VUV radiation has an impact on the emission of gaseous species, the mass loss rate and the mechanical properties of the material

  12. Behaviors of SiC fibers at high temperature

    International Nuclear Information System (INIS)

    Colin, C.; Falanga, V.; Gelebart, L.

    2010-01-01

    On the one hand, considering the improvements of mechanical and thermal behaviours of the last generation of SiC fibers (Hi-Nicalon S, Tyranno SA3); on the other hand, regarding physical and chemical properties and stability under irradiation, SiC/SiC composites are potential candidates for nuclear applications in advanced fission and fusion reactors. CEA must characterize and optimize these composites before their uses in reactors. In order to study this material, CEA is developing a multi-scale approach by modelling from fibers to bulk composite specimen: fibres behaviours must be well known in first. Thus, CEA developed a specific tensile test device on single fibers at high temperature, named MecaSiC. Using this device, we have already characterized the thermoelastic and thermoelectric behaviours of SiC fibers. Additional results about the plastic properties at high temperatures were also obtained. Indeed, we performed tensile tests between 1200 degrees C up to 1700 degrees C to characterize this plastic behaviour. Some thermal annealing, up to 3 hours at 1700 degrees C, had been also performed. Furthermore, we compare the mechanical behaviours with the thermal evolution of the electric resistivity of these SiC fibers. Soon, MecaSiC will be coupled to a new charged particle accelerator. Thus, in this configuration, we will be able to study in-situ irradiation effects on fibre behaviours, as swelling or creep for example

  13. Microstructural evolution under high temperature irradiation: fundamental aspects

    International Nuclear Information System (INIS)

    Martin, G.; Valentin, P.

    1984-01-01

    In view of the impossibility to propose theoretically established scaling laws for extrapolating microstructural evolutions to unknown irradiation conditions, a full modelization of microstructural evolution at the atomistic level cannot be avoided. We briefly review the main models available for describing: defect balance under irradiation, the nucleation of clusters of various types, the development of each of the components of the microstructure, synergistic effects among the latter. Attention is called on the problems which remain to be solved at each step. In particular, the swelling incubation phenomenon is just being studied from the fundamental viewpoint. A table of available relevant observations thereof is given. The existence of dose-rate thresholds accross which microstructural evolution undergoes a qualitative change is stressed. Such thresholds call for a detailed modelization of microstructural evolution in order to propose safe extrapolation techniques [fr

  14. An investigation of methods for neutron dose measurement in high temperature irradiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Kosako, Toshisou; Sugiura, Nobuyuki [Tokyo Univ. (Japan); Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan)] [and others

    2000-10-01

    The Japan Atomic Energy Research Institute (JAERI) has been conducting the innovative basic research on high temperature since 1994, which is a series of high temperature irradiation studies using the High Temperature Engineering Test Reactor (HTTR). 'The Task Group for Evaluation of Irradiation Dose under High Temperature Radiation' was founded in the HTTR Utilization Research Committee, which is the promoting body of the innovative basic research. The present report is a summary of investigation which has been made by the Task Group on the present status and subjects of research and development of neutron detectors in high temperature irradiation fields, in view of contributing to high temperature irradiation research using the HTTR. Detectors investigated here in the domestic survey are the following five kinds of in-core detectors: 1) small fission counter, 2) small fission chamber, 3) self-powered detector, 4) activation detector, and 5) optical fiber. In addition, the research and development status in Russia has been investigated. The present report will also be useful as nuclear instrumentation of high temperature gas-cooled reactors. (author)

  15. Effects of high temperature neutron irradiation on the physical, chemical and mechanical properties of fine-grained isotropic graphite

    International Nuclear Information System (INIS)

    Matsuo, H.; Nomura, S.; Imai, H.; Oku, T.; Eto, M.

    1987-01-01

    Effects of neutron irradiation on the dimensional change, coefficient of thermal expansion(CTE), thermal conductivity, corrosion rate, Young's modulus and strengths were studied for the candidate graphite material IG-110 of the experimental very high temperature gas-cooled reactor(VHTR) after irradiation at 585 - 1273 deg C to neutron fluences of up to about 3 x 10 25 n/m 2 (E > 29 fJ) in the JMTR and JRR-2, and to about 7 x 10 25 n/m 2 (E > 29 fJ) in the HFR. The results were compared with the irradiation behaviors of other graphites. Dimensional shrinkage was observed in the whole irradiation temperature range, showing lower value than 2 %. The shrinkage rate showed the minimum in the irradiation temperature of around 850 deg C, followed by the increase for the samples irradiated at higher temperatures. The dimensional stability of the material was clarified to be almost the same with that of H451 graphite. The CTE, thermal resistivity and Young's modulus increased in the early stage of irradiation and then only the CTE decreased while the thermal resistivity and Young's modulus levelled off with further irradiation. The neutron fluence showing the maximum CTE shifted to the lower fluence with increasing irradiation temperature. The increases of both thermal resistivity and Young's modulus were remarkable for the samples irradiated at lower temperatures. Compressive and bending strengths measured at room temperature increased after irradiation as well. The corrosion rate with water-vapor of 0.65 % in helium at high temperatures decreased owing to irradiation and the reduction was independent of irradiation temperature and neutron fluence. The activation energy for the reaction was estimated to be the same before and after irradiation. (author)

  16. High-Temperature Corrosion Behavior of Alloy 617 in Helium Environment of Very High Temperature Gas Reactor

    International Nuclear Information System (INIS)

    Lee, Gyeong-Geun; Jung, Sujin; Kim, Daejong; Jeong, Yong-Whan; Kim, Dong-Jin

    2012-01-01

    Alloy 617 is a Ni-base superalloy and a candidate material for the intermediate heat exchanger (IHX) of a very high temperature gas reactor (VHTR) which is one of the next generation nuclear reactors under development. The high operating temperature of VHTR enables various applications such as mass production of hydrogen with high energy efficiency. Alloy 617 has good creep resistance and phase stability at high temperatures in an air environment. However, it was reported that the mechanical properties decreased at a high temperature in an impure helium environment. In this study, high-temperature corrosion tests were carried out at 850°C-950°C in a helium environment containing the impurity gases H_2, CO, and CH_4, in order to examine the corrosion behavior of Alloy 617. Until 250 h, Alloy 617 specimens showed a parabolic oxidation behavior at all temperatures. The activation energy for oxidation in helium environment was 154 kJ/mol. The SEM and EDS results elucidated a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbides. The thickness and depths of degraded layers also showed a parabolic relationship with time. A normal grain growth was observed in the Cr-rich surface oxide layer. When corrosion tests were conducted in a pure helium environment, the oxidation was suppressed drastically. It was elucidated that minor impurity gases in the helium would have detrimental effects on the high temperature corrosion behavior of Alloy 617 for the VHTR application.

  17. An investigation of high-temperature irradiation test program of new ceramic materials

    International Nuclear Information System (INIS)

    Ishino, Shiori; Terai, Takayuki; Oku, Tatsuo

    1999-08-01

    The Japan Atomic Energy Research Institute entrusted the Atomic Energy Society of Japan with an investigation into the trend of irradiation processing/damage research on new ceramic materials. The present report describes the result of the investigation, which was aimed at effective execution of irradiation programs using the High Temperature Engineering Test Reactor (HTTR) by examining preferential research subjects and their concrete research methods. Objects of the investigation were currently on-going preliminary tests of functional materials (high-temperature oxide superconductor and high-temperature semiconductor) and structural materials (carbon/carbon and SiC/SiC composite materials), together with newly proposed subjects of, e.g., radiation effects on ceramics-coated materials and super-plastic ceramic materials as well as microscopic computer simulation of deformation and fracture of ceramics. These works have revealed 1) the background of each research subject, 2) its objective and significance from viewpoints of science and engineering, 3) research methodology in stages from preliminary tests to real HTTR irradiation, and 4) concrete HTTR-irradiation methods which include main specifications of test specimens, irradiation facilities and post-irradiation examination facilities and apparatuses. The present efforts have constructed the important fundamentals in the new ceramic materials field for further planning and execution of the innovative basic research on high-temperature engineering. (author)

  18. High-Temperature Creep-Fatigue Behavior of Alloy 617

    Directory of Open Access Journals (Sweden)

    Rando Tungga Dewa

    2018-02-01

    Full Text Available This paper presents the high-temperature creep-fatigue testing of a Ni-based superalloy of Alloy 617 base metal and weldments at 900 °C. Creep-fatigue tests were conducted with fully reversed axial strain control at a total strain range of 0.6%, 1.2%, and 1.5%, and peak tensile hold time of 60, 180, and 300 s. The effects of different constituents on the combined creep-fatigue endurance such as hold time, strain range, and stress relaxation behavior are discussed. Under all creep-fatigue tests, weldments’ creep-fatigue life was less than base metal. In comparison with the low-cycle fatigue condition, the introduction of hold time decreased the cycle number of both base metal and weldments. Creep-fatigue lifetime in the base metal was continually decreased by increasing the tension hold time, except for weldments under longer hold time (>180 s. In all creep-fatigue tests, intergranular brittle cracks near the crack tip and thick oxide scales at the surface were formed, which were linked to the mixed-mode creep and fatigue cracks. Creep-fatigue interaction in the damage-diagram (D-Diagram (i.e., linear damage summation was evaluated from the experimental results. The linear damage summation was found to be suitable for the current limited test conditions, and one can enclose all the data points within the proposed scatter band.

  19. Mechanical behavior of high strength ceramic fibers at high temperatures

    Science.gov (United States)

    Tressler, R. E.; Pysher, D. J.

    1991-01-01

    The mechanical behavior of commercially available and developmental ceramic fibers, both oxide and nonoxide, has been experimentally studied at expected use temperatures. In addition, these properties have been compared to results from the literature. Tensile strengths were measured for three SiC-based and three oxide ceramic fibers for temperatures from 25 C to 1400 C. The SiC-based fibers were stronger but less stiff than the oxide fibers at room temperature and retained more of both strength and stiffness to high temperatures. Extensive creep and creep-rupture experiments have been performed on those fibers from this group which had the best strengths above 1200 C in both single filament tests and tests of fiber bundles. The creep rates for the oxides are on the order of two orders of magnitude faster than the polymer derived nonoxide fibers. The most creep resistant filaments available are single crystal c-axis sapphire filaments. Large diameter CVD fabricated SiC fibers are the most creep and rupture resistant nonoxide polycrystalline fibers tested to date.

  20. In-situ high temperature irradiation setup for temperature dependent structural studies of materials under swift heavy ion irradiation

    International Nuclear Information System (INIS)

    Kulriya, P.K.; Kumari, Renu; Kumar, Rajesh; Grover, V.; Shukla, R.; Tyagi, A.K.; Avasthi, D.K.

    2015-01-01

    An in-situ high temperature (1000 K) setup is designed and installed in the materials science beam line of superconducting linear accelerator at the Inter-University Accelerator Centre (IUAC) for temperature dependent ion irradiation studies on the materials exposed with swift heavy ion (SHI) irradiation. The Gd 2 Ti 2 O 7 pyrochlore is irradiated using 120 MeV Au ion at 1000 K using the high temperature irradiation facility and characterized by ex-situ X-ray diffraction (XRD). Another set of Gd 2 Ti 2 O 7 samples are irradiated with the same ion beam parameter at 300 K and simultaneously characterized using in-situ XRD available in same beam line. The XRD studies along with the Raman spectroscopic investigations reveal that the structural modification induced by the ion irradiation is strongly dependent on the temperature of the sample. The Gd 2 Ti 2 O 7 is readily amorphized at an ion fluence 6 × 10 12 ions/cm 2 on irradiation at 300 K, whereas it is transformed to a radiation-resistant anion-deficient fluorite structure on high temperature irradiation, that amorphized at ion fluence higher than 1 × 10 13 ions/cm 2 . The temperature dependent ion irradiation studies showed that the ion fluence required to cause amorphization at 1000 K irradiation is significantly higher than that required at room temperature irradiation. In addition to testing the efficiency of the in-situ high temperature irradiation facility, the present study establishes that the radiation stability of the pyrochlore is enhanced at higher temperatures

  1. Status for development of a capsule and instruments for high-temperature irradiation in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Man Soon; Choo, Kee Nam; Lee, Chul Yong; Yang, Seong Woo; Shim, Kyue Taek; Chung, Hwan-Sung [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

    2012-03-15

    As the reactors planned in the Gen-IV program will be operated at high temperature and under high neutron flux, the requirements for irradiation of materials at high temperature are recently being gradually increased. The irradiation tests of materials in HANARO up to the present have been performed usually at temperatures below 300degC at which the RPV materials of the commercial reactors are being operated. To overcome the restriction for high-temperature use of Al thermal media of the existing standard capsule, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated as a more advanced capsule than the single thermal media capsule. (author)

  2. The high temperature behavior of In2O3

    NARCIS (Netherlands)

    Wit, J.H.W. de

    The electrical conductivity of In2O3 has been measured up to 1400°C in air. The temperature dependence of the conductivity at high temperatures yields an activation energy of 1.5 ± 0.1 eV. This activation energy is interpreted in terms of a nonstoichiometric decomposition of the compound. This

  3. Preparation of the Crosslinked Polyethersulfone Films by High Temperature Electron-Beam Irradiation

    International Nuclear Information System (INIS)

    Li, J.

    2006-01-01

    The aromatic polymers, mainly so called engineering plastics, were famed for the good stability under irradiation. However, high temperature irradiation of the aromatic polymers can result the crosslinked structure, due to the improved molecular mobility. Polyethersulfone (PES) is a wide used engineering plastic because of the high performance and high thermal stability. PES films were irradiated by electron-beam under nitrogen atmosphere above the glass transition temperature and then the covalently crosslinked PES (RX-PES) films were obtained. The irradiations were also performed at ambient temperature for comparison. The network structure formation of the RX-PES films was confirmed by the appearance of the gel, which were measured by soaking the irradiated PES films in the N,N-dimethylformamide (DMF) at room temperature. When the PES films were irradiated to 300 kGy, there was gel appeared. The gel percent increased with the increasing in the absorbed dose, and saturated when the absorbed dose exceeded 1200 kGy. However, there was no gel formed for the PES films irradiated at ambient temperature even to 2250 kGy. The G(S) and G(X) were calculated according to the Y-crosslinking mechanism. The results values are consistent in error range. G(S) of 0.10 and G(X) of 0.23 were obtained. As calculated, almost all the macromolecular radicals produced by chain scission were used for crosslinking. Also, the glass transition temperature of the RX-PES films increased with the increasing in the absorbed doses, while the glass transition temperature of the PES films irradiated at ambient temperature decreased with the increasing in the absorbed doses. The blending films of the PES with FEP or ETFE were prepared and the high temperature irradiation effects were also studies

  4. Instrumented indentation for characterization of irradiated metals at room and high temperatures

    International Nuclear Information System (INIS)

    Sacksteder, Irene

    2011-01-01

    The reliability and sustainability of future fusion power plants will highly depend on the aptitude of materials to withstand severe irradiation conditions induced by the burning plasma in reactors. The so-called reduced-activation ferritic-martensitic (RAFM) steels are the current promising candidates for the structural applications considering the reactor's first wall. These steels exhibit irradiation embrittlement and hardening for defined irradiation conditions that are mainly characterized by the irradiation temperature and the irradiation dose. A proper characterization of such irradiated steels implies the use of adapted mechanical testing tools. In the present study, the instrumented indentation technique makes use of a post-processing tool based on neural networks. This technique has been selected for its ability to examine tensile properties by multistage indents on miniaturized irradiated metallic samples. The steel specimens studied in this project have been neutron-irradiated up to a dose of 15 dpa. They have been subsequently tested at room temperature in a Hot Cell by means of an adapted commercial indentation device. The significant irradiation-induced hardening effect present in the range of 250-350 deg C could be observed in the hardness and material's strength parameters. These two material parameters show a similar evolution with increasing irradiation temperatures. Post-irradiation annealing treatments of Eurofer97 have been realized and leads to a partial recovery of the irradiation damage. Considering the demands for characterization in irradiated steels at high temperature and for post-irradiation annealing experiments, the existing instrumented indentation device has been further developed during this work. A conceptual design has been proposed for an indentation testing machine, operating at up to 650 deg C, while remaining the critical temperature limit for tensile strength of the newly developed oxide dispersion strengthening ferritic

  5. New temperature monitoring devices for high-temperature irradiation experiments in the high flux reactor Petten

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M.; Futterer, M. A.; Lapetite, J. M. [European Commission Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, 61438 Flers Cedex (France); Morice, R. [Laboratoire National de Metrologie et d' Essais, 1 rue Gaston Boissier, 75724 Paris (France)

    2009-07-01

    Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some research and development work for further improvement of thermocouples and other on-line instrumentation will be outlined. (authors)

  6. High-temperature annealing of proton irradiated beryllium – A dilatometry-based study

    Energy Technology Data Exchange (ETDEWEB)

    Simos, Nikolaos, E-mail: simos@bnl.gov [Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed; Zhong, Zhong; Ghose, Sanjit [Brookhaven National Laboratory, Upton, NY, 11973 (United States); Savkliyildiz, Ilyas [Rutgers University (United States)

    2016-08-15

    S−200 F grade beryllium has been irradiated with 160 MeV protons up to 1.2 10{sup 20} cm{sup −2} peak fluence and irradiation temperatures in the range of 100–200 °C. To address the effect of proton irradiation on dimensional stability, an important parameter in its consideration in fusion reactor applications, and to simulate high temperature irradiation conditions, multi-stage annealing using high precision dilatometry to temperatures up to 740 °C were conducted in air. X-ray diffraction studies were also performed to compliment the macroscopic thermal study and offer a microscopic view of the irradiation effects on the crystal lattice. The primary objective was to qualify the competing dimensional change processes occurring at elevated temperatures namely manufacturing defect annealing, lattice parameter recovery, transmutation {sup 4}He and {sup 3}H diffusion and swelling and oxidation kinetics. Further, quantification of the effect of irradiation dose and annealing temperature and duration on dimensional changes is sought. The study revealed the presence of manufacturing porosity in the beryllium grade, the oxidation acceleration effect of irradiation including the discontinuous character of oxidation advancement, the effect of annealing duration on the recovery of lattice parameters recovery and the triggering temperature for transmutation gas diffusion leading to swelling.

  7. A Study on the High Temperature Irradiation Test Possibility for the HANARO Outer Core Region

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Hwan; Cho, M. S.; Choo, K. N.; Shin, Y. T.; Sohn, J. M.; Park, S. J.; Kim, B. G

    2008-01-15

    1. Information on the neutron flux levels and the gamma heat of the concerned test holes, which have been produced from a series of nuclear analysis and tests performed at KAERI since 1993, were collected and analyzed to develop the nuclear data for the concerned test holes of HANARO and to develop the new design concepts of a capsule for the high temperature irradiation devices. 2. From the literature survey and analysis about the system design characteristics of the new concepts of irradiation devices in the ATR and MIT reactor, U.S. and the JHR reactor, France, which are helpful in understanding the key issues for the on-going R and D programmes related to a SFR and a VHTR, the most important parameters for the design of high temperature irradiation devices are identified as the neutron spectrum, the heat generation density, the fuel and cladding temperature, and the coolant chemistry. 3. From the thermal analysis of a capsule by using a finite element program ANSYS, high temperature test possibility at the OR and IP holes of HANARO was investigated based on the data collected from a literature survey. The OR holes are recommended for the tests of the SFR and VHTR nuclear materials. The IP holes could be applicable for an intermediate temperature irradiation of the SWR and LMR materials. 4. A thermal analysis for the development of a capsule with a new configuration was also performed. The size of the center hole, which is located at the thermal media of a capsule, did not cause specimen temperature changes. The temperature differences are found to be less than 2%. The introduction of an additional gap in the thermal media was able to contribute to an increase in the specimen temperature by up to 27-90 %.

  8. Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

    International Nuclear Information System (INIS)

    Kano, F.; Tsuchiya, Y.; Oka, K.

    2009-01-01

    The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

  9. Tribological behavior of zirconium coatings in high temperature helium

    International Nuclear Information System (INIS)

    Cachon, Lionel; Albaladejo, Serge; Taraud, Pascal

    2005-01-01

    In France, a comprehensive research and development program is leaded by the CEA, since 2001, for the Gas Cooled Reactor (GCR) project using helium as cooling fluid, in order to establish the feasibility of the technology of an early VHTR prototype to be started by 2015, and then to qualify the generic VHTR technology, so as to meet similar objectives for the GFR. In this frame a tribology program has been launched. The purpose of the work presented in this paper is to describe the CEA Helium tribology study: high temperature gas cooled reactors require wear protection (thermal barriers, control rod drive mechanisms, reactor internals, ...). Tests in helium atmosphere are necessary to be fully representative of tribological environments and finally to check the possible materials or coatings which can provide a reliable answer to these situations. The main characteristics and first experimental results are thus described. This paper focus on tribology tests leaded in the temperature range 800-1000degC, on ceramic (ZrO 2 -Y 2 O 3 ) with and without solid lubricant like CaF2). (author)

  10. Survey report on high temperature irradiation experiment programs for new ceramic materials in the HTTR (High Temperature Engineering Test Reactor). 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-02-01

    A survey research on status of research activities on new ceramic materials in Japan was carried out under contract between Japan Atomic Energy Research Institute and Atomic Energy Society of Japan. The purpose of the survey is to provide information to prioritize prospective experiments and tests in the HTTR. The HTTR as a high temperature gas cooled reactor has a unique and superior capability to irradiate large-volumed specimen at high temperature up to approximately 800degC. The survey was focused on mainly the activities of functional ceramics and heat resisting ceramics as a kind of structural ceramics. As the result, the report recommends that the irradiation experiment of functional ceramics is feasible to date. (K. Itami)

  11. High-temperature behavior of advanced spacecraft TPS

    Science.gov (United States)

    Pallix, Joan

    1994-05-01

    The objective of this work has been to develop more efficient, lighter weight, and higher temperature thermal protection systems (TPS) for future reentry space vehicles. The research carried out during this funding period involved the design, analysis, testing, fabrication, and characterization of thermal protection materials to be used on future hypersonic vehicles. This work is important for the prediction of material performance at high temperature and aids in the design of thermal protection systems for a number of programs including programs such as the National Aerospace Plane (NASP), Pegasus and Pegasus/SWERVE, the Comet Rendezvous and Flyby Vehicle (CRAF), and the Mars mission entry vehicles. Research has been performed in two main areas including development and testing of thermal protection systems (TPS) and computational research. A variety of TPS materials and coatings have been developed during this funding period. Ceramic coatings were developed for flexible insulations as well as for low density ceramic insulators. Chemical vapor deposition processes were established for the fabrication of ceramic matrix composites. Experimental testing and characterization of these materials has been carried out in the NASA Ames Research Center Thermophysics Facilities and in the Ames time-of-flight mass spectrometer facility. By means of computation, we have been better able to understand the flow structure and properties of the TPS components and to estimate the aerothermal heating, stress, ablation rate, thermal response, and shape change on the surfaces of TPS. In addition, work for the computational surface thermochemistry project has included modification of existing computer codes and creating new codes to model material response and shape change on atmospheric entry vehicles in a variety of environments (e.g., earth and Mars atmospheres).

  12. The corrosion behavior of hafnium in high-temperature-water environments

    Energy Technology Data Exchange (ETDEWEB)

    Rishel, D.M.; Smee, J.D.; Kammenzind, B.F.

    1999-10-01

    The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.

  13. Unusual high-temperature behavior of neptunium and plutonium systems

    International Nuclear Information System (INIS)

    Brewer, L.

    1991-01-01

    The initial Manhattan Project effort to predict the properties of transuranium elements was based on the assumption that the lanthanides would serve as models. However, since then, many unexpected differences have been found. With the extensive work on the electronic configurations of the gaseous atoms, it is now possible to understand the reasons for the different behavior of the 5f elements compared to the 4f elements. In this paper a review of these differences are presented along with examples of other unexpected behavior such as the very strong generalized Lewis acid-base interactions of the 5f elements with the platinum-group metals that provide techniques for preparation of pure metal samples. The thermodynamic properties of the transuranium metals and some of the compounds will be reviewed

  14. Evaluation of Candidate Linear Variable Displacement Transducers for High Temperature Irradiations in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Knudson, D.L.; Rempe, J.L.; Daw, J.E.

    2009-01-01

    The United States (U.S.) Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to promote nuclear science and technology in the U.S. Given this designation, the ATR is supporting new users from universities, laboratories, and industry as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A fundamental component of the ATR NSUF program is to develop in-pile instrumentation capable of providing real-time measurements of key parameters during irradiation experiments. Dimensional change is a key parameter that must be monitored during irradiation of new materials being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. Such materials can experience significant changes during high temperature irradiation. Currently, dimensional changes are determined by repeatedly irradiating a specimen for a defined period of time in the ATR and then removing it from the reactor for evaluation. The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data (i.e., only characterizing the end state when samples are removed from the reactor) and may disturb the phenomena of interest. To address these issues, the Idaho National Laboratory (INL) recently initiated efforts to evaluate candidate linear variable displacement transducers (LVDTs) for use during high temperature irradiation experiments in typical ATR test locations. Two nuclear grade LVDT vendor designs were identified for consideration - a smaller diameter design qualified for temperatures up to 350 C and a larger design with capabilities to 500 C. Initial evaluation efforts include collecting calibration data as a function of temperature, long duration testing of LVDT response while held at high temperature, and the assessment of changes

  15. Hydrogen permeation behavior through F82H at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Matsuda, S.; Katayama, K.; Shimozori, M.; Fukada, S. [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Kyushu (Japan); Ushida, H. [Energy Science and Engineering, Faculty of Engineering, Kyushu University, Kyushu (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur (Malaysia)

    2015-03-15

    F82H is a primary candidate of structural material and coolant pipe material in a blanket of a fusion reactor. Understanding tritium permeation behavior through F82H is important. In a normal operation of a fusion reactor, the temperature of F82H will be controlled below 550 C. degrees because it is considered that F82H can be used up to 30,000 hours at 550 C. degrees. However, it is necessary to assume the situation where F82H is heated over 550 C. degrees in a severe accident. In this study, hydrogen permeation behavior through F82H was investigated in the temperature range from 500 to 800 C. degrees. In some cases, water vapor was added in a sample gas to investigate an effect of water vapor on hydrogen permeation. The permeability of hydrogen in the temperature range from 500 to 700 C. degrees agreed well with the permeability reported by E. Serra et al. The degradation of the permeability by water vapor was not observed. After the hydrogen permeation reached in a steady state at 700 C. degrees, the F82H sample was heated to 800 C. degrees. The permeability of hydrogen through F82H sample which was once heated up to 800 C. degrees was lower than that of the original one. (authors)

  16. High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Suzuki, Motoye

    1982-03-01

    In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

  17. Irradiation effects on C/C composite materials for high temperature nuclear applications

    International Nuclear Information System (INIS)

    Eto, M.; Ugachi, H.; Baba, S.I.; Ishiyama, S.; Ishihara, M.; Hayashi, K.

    2000-01-01

    Excellent characteristics such as high strength and high thermal shock resistance of C/C composite materials have led us to try to apply them to the high temperature components in nuclear facilities. Such components include the armour tile of the first wall and divertor of fusion reactor and the elements of control rod for the use in HTGR. One of the most important aspects to be clarified about C/C composites for nuclear applications is the effect of neutron irradiation on their properties. At the Japan Atomic Energy Research Institute (JAERI), research on the irradiation effects on various properties of C/C composite materials has been carried out using fission reactors (JRR-3, JMTR), accelerators (TANDEM, TIARA) and the Fusion Neutronics Source (FNS). Additionally, strength tests of some neutron-irradiated elements for the control rod were carried out to investigate the feasibility of C/C composites. The paper summarises the R and D activities on the irradiation effects on C/C composites. (authors)

  18. Behavior of mercury in high-temperature vitrification processes

    International Nuclear Information System (INIS)

    Goles, R.W.; Holton, K.K.; Sevigny, G.J.

    1992-01-01

    This paper reports that the Pacific Northwest Laboratory (PNL) has evaluated the waste processing behavior of mercury in simulated defense waste. A series of tests were performed under various operating conditions using an experimental-scale liquid-fed ceramic melter (LFCM). This solidification technology had no detectable capacity for incorporating mercury into its product, borosilicate glass. Chemically, the condensed mercury effluent was composed almost entirely of chlorides, and except in a low-temperature test, Hg 2 Cl 2 was the primary chloride formed. As a result, combined mercury accounted for most of the insoluble mass collected by the process quench scrubber. Although macroscopic quantities of elemental mercury were never observed in process secondary waste streams, finely divided and dispersed mercury that blackened all condensed Hg 2 Cl 2 residues was capable of saturating the quenched process exhaust with mercury vapor. The vapor pressure of mercury, however, in the quenched melter exhaust was easily and predictably controlled with the off-gas stream chiller

  19. High-temperature vaporization behavior of oxygen-deficient thoria

    International Nuclear Information System (INIS)

    Ackermann, R.J.; Tetenbaum, M.

    1979-01-01

    The experimental results of the present study on the vaporization behavior of oxygen-deficient thoria are directed toward a more precise and detailed study of the lower phase boundary (l.p.b.) and congruently vaporizing composition (c.v.c), and intermediate compositions, and the corresponding oxygen potentials and total pressure at temperatures above 2000K. The l.p.b. and c.v.c. values were found to fit an equation of the form log x = A + (B/T), where x is the stoichiometric defect in ThO 2 -x. Oxygen potentials corresponding to the l.p.b. and c.v.c. have been estimated from vapor pressures and thermodynamic data. A very sharp decrease in oxygen potential occurs when thoria isreduced only slightly from the stoichiometric composition. In the temperature range from 2400 to 2655 K, the oxygen partial pressure dependency of x in ThO 2 -x was found to be approximately proportional to PO 2 - 1 /4to PO 2 - 1 /. The small extent of reduction over a wide range of oxygen potentials at these temperatures is a clear illustration of the higher stability of the ThO 2 -x phase compared with that of UO 2 -x. Values of ΔHO 2 and ΔSO 2 have been estimated for selected compositions from the dependence of the measured oxygen potential on temperature. Estimates of the standard free energy of formation of bivariant ThO 2 -x compositions have been made. A substantial increase in the total pressure of thorium-bearing species occurs when stoichiometric thoria is reduced toward the lower phase boundary. (orig.) [de

  20. Use of TRIGA-pulsed irradiations for high-temperature Doppler measurements

    Energy Technology Data Exchange (ETDEWEB)

    Foell, W K; Cashwell, R J; Bhattacharyya, S K [Argonne National Laboratory, Argonne, IL (United States); Russell, G J [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

    1974-07-01

    Conventional activation and reactivity measurements of the nuclear Doppler Effect have been limited to temperatures of about 2000{sup o}K because of problems with furnace equipment. There is a need for Doppler data at higher temperatures for design of reactors and analysis of reactor accidents. To fill this need, a novel technique using pulsed-mode operation of a TRIGA reactor has been developed at the University of Wisconsin. This new method, the Pulsed Activation Doppler (PAD) technique, has been used successfully for high temperature Doppler measurements of UO{sub 2} fuel pellets. In the PAD technique, UO{sub 2} test pellets were doped with varying amounts of U-235, with fissile enrichments varying from 0.22% to 12% by weight. The pellets were encapsulated in individual irradiation cells and electrically preheated to predetermined temperatures. Pyrofoam-graphite heaters were used to give preheat temperatures of up to 1720 deg. K. The cells were then positioned in the University of Wisconsin TRIGA reactor core and pulse-irradiated. During the rapid irradiation, adiabatic fission energy deposition occurred in the pellets and very high temperatures (over 3115 deg, K) were attained. Corresponding resonance neutron captures occurred at the elevated temperatures. The Doppler Ratio was deduced from the gamma activities of the Np-239 in the heated and unheated reference pellets. UO{sub 2} pellets of two nominal diameters, 210 mils (a surface-to-mass ratio, s/m = 1.1 cm{sup 2} /gm) and 360 mils (s/m = 0.63 cm{sup 2}/gm), were used for the experiments. For the 210 mil diameter pellets there was very good agreement between experimental results and Doppler ratios predicted both from extrapolations of the Hellstrand low-temperature resonance integral correlations and from GAROL calculations. Significantly, the agreement was good even for those pellets which experienced extensive melting. For the 360 mil diameter pellets the theoretical predictions were 10-15% lower than

  1. High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Manzel, R.; Everett, M. R.; Graham, L. W.

    1971-05-15

    The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 1021 n cm-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.

  2. Electrochemical behavior of TIO{sub 2} deposited stainless steel in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Okamura, M.; Yamamoto, S. [Toshiba Corp., Kawasaki-city, Kanagawa (Japan); Urata, H.; Takagi, J. [Toshiba Corp., Yokohama-city, Kanagawa (Japan)

    2010-07-01

    It has previously been confirmed that the electrochemical corrosion potential (ECP) of stainless steel (SS) shifts in the negative direction by deposition of TiO{sub 2}. Recently we showed that TiO{sub 2} could decrease the ECP of SS in the absence of UV irradiation. In this study we measured the anodic polarization curve in high temperature water under UV irradiation and none irradiation condition and considered the mechanism of the ECP shift by TiO{sub 2} deposition. The anodic current density of the specimen increased with increasing the UV irradiation intensity and with increasing the amount of TiO{sub 2} deposition under none UV irradiation. Furthermore the oxide film of the specimen affects on the anodic current density was clarified. It was verified the ECP shift is caused by the anodic current density increasing with TiO{sub 2} deposition under both conditions of UV and none UV irradiation. (author)

  3. Comparative Analysis of Single and Dual Irradiation Pass of Deep Burn High Temperature Reactor Scenario

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Jo, Chang Keun; Noh, Jae Man

    2012-01-01

    A concept of a deep-burn (DB) of trans uranic (TRU) elements in a high temperature reactor (HTR) has been proposed and studied with a single irradiation pass. However, there is still a significant amount of TRU after burn in an HTR. Therefore, it is necessary to burn more TRU in a fast reactor (FR) with repeated reprocessing such as a pyro-process. In this study, the fuel cycle calculations are performed and the results are compared for a singlepass DB-HHR scenario and a dual-pass sodium-cooled fast reactor (SFR) scenario. For the analysis, front-end and back-end parameters are compared. The calculations were performed by the DANESS (Dynamic Analysis of Nuclear Energy System Strategies), which is an integrated system dynamic fuel cycle analysis code

  4. Erosion and mass transfer of Mo, W and Nb under neutron irradiation of high temperature materials

    International Nuclear Information System (INIS)

    Berzhatyj, V.I.; Luk'yanov, A.N.; Zavalishin, A.A.; Tkach, V.N.; Fedorenko, A.I.

    1980-01-01

    Studies have been made of the medium composition in thermionic fuel elements of two types during reactor tests; erosion and mass transfer of electrode materials have been investigated in the after-reactor analysis of the tested fuel elements. The studies of electrode material evaporation at the conditions approaching (in environment temperature and composition) those of reactor tests of thermionic fuel elements have shown that the process proceeds in the form of metal oxides. Evaporation rates are determined, the mechanism of evaporation is discussed, and the analytical dependences are obtained for calculating the evaporation rates of Mo and W at certain temperature and gaseous medium composition. It is found that the main contribution to the material transfer off the Mo and Nb surfaces under a high-temperature reactor irradiation comes through the thermal evaporation; in the case of tungsten at the same experimental conditions the rates of mass transfer due to thermal evaporation and neutron sputtering are nearly the same [ru

  5. Superconductivity degradation in Gd-containing high temperature superconductors (HTSC) under thermal neutron irradiation

    International Nuclear Information System (INIS)

    Petrov, A.; Kudrenitskis, I.; Makletsov, A.; Arhipov, A.; Karklin, N.

    1999-01-01

    The physical properties of ordered crystals are extremely sensitive to the degree of order in the distribution of the various kinds of atoms over the corresponding sites in the crystal lattice. An increasingly popular means of creating disordered states is to use nuclear radiation. The type of radiation defects which appear and the nature and degree of the structural changes in ordered crystals depend on the kind of radiation and the fluence level, the irradiation temperature, the type of crystal structure, the composition and initial disorder of the material, the character of the interatomic forces, etc. There are many such scientific publications where the effects of fast neutron irradiation on high temperature superconductors (HTSC) have been studied in both polycrystalline and single crystalline superconductors. It is known also that the role of thermal neutrons in structural defects forming is negligible in comparison with fast neutrons because of their small (∼0.025 eV) energy. But it is evident enough that in superconductors containing isotopes with large thermal neutron cross sections the important results concerning the role of point defects could be obtained. Such point defects are creating due to soft displacements of isotopes having interacted with thermal neutrons. Such the possibility of creating point defects in solids including HTSC is investigating by several groups (Austria, USA, China, Latvia) and these investigations have found the support in the person of IAEA. In this review the authors consider the changes brought about by thermal-neutron irradiation (E∼0.025 eV) in the structure, superconducting and magnetic properties of gadolinium containing ordered HTSC with the structure 123, whose extreme electric and magnetic properties continue to attract both research and practical interest. All of the studies reviewed have been done on bulk polycrystalline samples RBa 2 Cu 3 O 7-δ (where R - natural mixture of Gd isotopes, 155 Gd, 157 Gd, 160

  6. Array-type sensor to determine corrosive conditions in high temperature water under gamma rays irradiation

    International Nuclear Information System (INIS)

    Satoh, T.; Tsukada, T.; Uchida, S.; Katoh, C.

    2010-01-01

    One of the problems to determine electrochemical corrosion potential (ECP) in high temperature water under irradiation is to apply long-lived and reliable reference electrodes. In order to avoid troubles due to the reference electrode, a new concept to determine ECP without the reference electrode has been proposed. Several metal plates are applied as working electrodes and at the same time as the reference electrodes. Potential of the metal plates with stable oxide films on their surfaces show stable values in high temperature water. As a result of the combination of their potential values, ECP of each metal can be determined without any specific reference electrode. Array-type sensors consisting of several metal plates, e.g., Fe, Ni, Cr, Zr, Pt, Pd, Re, Ir, with well developed oxide films on their surface were prepared for ECP measurement in high temperature water under neutron/gamma ray irradiations. In order to confirm the feasibility of this concept, responses of the redox potentials of the pure metals to changes in the simulated BWR reactor water conditions were measured and the ECP was determined by the differences in potentials between a couple of metal plates. Major conclusions of the study are as follows: 1) The redox potentials of the Fe, Pt, Zr, Ir, Pd, and Re electrodes showed the different dependences on the changes in O 2 and H 2 O 2 concentrations. The redox potentials of the electrodes increased as the oxidant concentrations increased except for Zr electrode. The potential of the Zr electrode was kept the very low potential at the wide range of O 2 and H 2 O 2 concentrations differed form the other electrodes. 2) It was estimated that the redox potential of highly soluble metal may be increased, while that of low soluble metal may be decreased by an oxide film. The stable oxide film would cause the stable potential response of the electrode with oxide film. 3) The relationship between the oxidant concentrations and the redox potentials of the

  7. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  8. Artificial neural networks in prediction of mechanical behavior of concrete at high temperature

    International Nuclear Information System (INIS)

    Mukherjee, A.; Nag Biswas, S.

    1997-01-01

    The behavior of concrete structures that are exposed to extreme thermo-mechanical loading is an issue of great importance in nuclear engineering. The mechanical behavior of concrete at high temperature is non-linear. The properties that regulate its response are highly temperature dependent and extremely complex. In addition, the constituent materials, e.g. aggregates, influence the response significantly. Attempts have been made to trace the stress-strain curve through mathematical models and rheological models. However, it has been difficult to include all the contributing factors in the mathematical model. This paper examines a new programming paradigm, artificial neural networks, for the problem. Implementing a feedforward network and backpropagation algorithm the stress-strain relationship of the material is captured. The neural networks for the prediction of uniaxial behavior of concrete at high temperature has been presented here. The results of the present investigation are very encouraging. (orig.)

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

    International Nuclear Information System (INIS)

    Lambard, V.

    2000-01-01

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

  10. Influence of High Temperature Treatment on Mechanical Behavior of a Coarse-grained Marble

    Science.gov (United States)

    Rong, G.; Peng, J.; Jiang, M.

    2017-12-01

    High temperature has a significant influence on the physical and mechanical behavior of rocks. With increasing geotechnical engineering structures concerning with high temperature problems such as boreholes for oil or gas production, underground caverns for storage of radioactive waste, and deep wells for injection of carbon dioxides, etc., it is important to study the influence of temperature on the physical and mechanical properties of rocks. This paper experimentally investigates the triaxial compressive properties of a coarse-grained marble after exposure to different high temperatures. The rock specimens were first heated to a predetermined temperature (200, 400, and 600 oC) and then cooled down to room temperature. Triaxial compression tests on these heat-treated specimens subjected to different confining pressures (i.e., 0, 5, 10, 15, 20, 25, 30, 35, and 40 MPa) were then conducted. Triaxial compression tests on rock specimens with no heat treatment were also conducted for comparison. The results show that the high temperature treatment has a significant influence on the microstructure, porosity, P-wave velocity, stress-strain relation, strength and deformation parameters, and failure mode of the tested rock. As the treatment temperature gradually increases, the porosity slightly increases and the P-wave velocity dramatically decreases. Microscopic observation on thin sections reveals that many micro-cracks will be generated inside the rock specimen after high temperature treatment. The rock strength and Young's modulus show a decreasing trend with increase of the treatment temperature. The ductility of the rock is generally enhanced as the treatment temperature increases. In general, the high temperature treatment weakens the performance of the tested rock. Finally, a degradation parameter is defined and a strength degradation model is proposed to characterize the strength behavior of heat-treated rocks. The results in this study provide useful data for

  11. Comparison of the thermodynamic properties and high temperature chemical behavior of lanthanide and actinide oxides

    International Nuclear Information System (INIS)

    Ackermann, R.J.; Rauh, E.G.

    1977-01-01

    The thermodynamic properties of the lanthanide and actinide oxides are examined, compared, and associated with a variety of high temperature chemical behavior. Trends are cited resulting from a number of thermodynamic and spectroscopic correlations involving solid phases, species in aqueous solution, and molecules and ions in the vapor phase. Inadequacies in the data and alternative approaches are discussed. The characterization of nonstoichiometric phases stable only at high temperatures is related to a network of heterogeneous and homogeneous equilibria. A broad perspective of similarity and dissimilarity between the lanthanides and actinides emerges and forms the basis of the projected needs for further study

  12. Decarburization behavior and mechanical properties of Inconel 617 during high temperature oxidation in He environment

    International Nuclear Information System (INIS)

    Kim, Young Do; Kim, Dae Gun; Jo, Tae Sun; Kim, Hoon Sup; Lim, Jeong Hun

    2010-04-01

    Among Generation IV reactor concepts, high temperature gas-cooled reactors (HTGRs) are high-efficiency systems designed for the economical production of hydrogen and electricity. Inconel 617 is a solid-solution strengthening Ni-based superalloy that shows excellent strength, creep-rupture strength, and oxidation resistance at high temperatures. Thus, it is a desirable candidate for tube material of IHX and HGD in HTGRs. In spite of these excellent properties, aging degradation by long time exposure at high temperature induced to deterioration of mechanical properties and furthermore alloys' lifetime because of Cr-depleted zone and carbide free zone below external scale. Also, machinability of Inconel 617 is a important property for system design. In this study, oxidation and decarbrization behavior were evaluated at various aging temperature and environment. Also, cold rolling was carried out for the machinability evaluation of Inconel 617 and then microstructure change was evaluated

  13. High temperature oxidation behavior of SiC coating in TRISO coated particles

    International Nuclear Information System (INIS)

    Liu, Rongzheng; Liu, Bing; Zhang, Kaihong; Liu, Malin; Shao, Youlin; Tang, Chunhe

    2014-01-01

    Highlights: • High temperature oxidation tests of SiC coating in TRISO particles were carried out. • The dynamic oxidation process was established. • Oxidation mechanisms were proposed. • The existence of silicon oxycarbides at the SiO 2 /SiC interface was demonstrated. • Carbon was detected at the interface at high temperatures and long oxidation time. - Abstract: High temperature oxidation behavior of SiC coatings in tristructural-isotropic (TRISO) coated particles is crucial to the in-pile safety of fuel particles for a high temperature gas cooled reactor (HTGR). The postulated accident condition of air ingress was taken into account in evaluating the reliability of the SiC layer. Oxidation tests of SiC coatings were carried out in the ranges of temperature between 800 and 1600 °C and time between 1 and 48 h in air atmosphere. Based on the microstructure evolution of the oxide layer, the mechanisms and kinetics of the oxidation process were proposed. The existence of silicon oxycarbides (SiO x C y ) at the SiO 2 /SiC interface was demonstrated by X-ray photospectroscopy (XPS) analysis. Carbon was detected by Raman spectroscopy at the interface under conditions of very high temperatures and long oxidation time. From oxidation kinetics calculation, activation energies were 145 kJ/mol and 352 kJ/mol for the temperature ranges of 1200–1500 °C and 1550–1600 °C, respectively

  14. Studies of the corrosion and cracking behavior of steels in high temperature water by electrochemical techniques

    International Nuclear Information System (INIS)

    Cheng, Y.F.; Bullerwell, J.; Steward, F.R.

    2003-01-01

    Electrochemical methods were used to study the corrosion and cracking behavior of five Fe-Cr alloy steels and 304L stainless steel in high temperature water. A layer of magnetite film forms on the metal surface, which decreases the corrosion rate in high temperature water. Passivity can be achieved on A-106 B carbon steel with a small content of chromium, which cannot be passivated at room temperature. The formation rate and the stability of the passive film (magnetite film) increased with increasing Cr-content in the steels. A mechanistic model was developed to simulate the corrosion and cracking processes of steels in high temperature water. The crack growth rate on steels was calculated from the maximum current of the repassivation current curves according to the slip-oxidation model. The highest crack growth rate was found for 304L stainless steel in high temperature water. Of the four Fe-Cr alloys, the crack growth rate was lower on 0.236% Cr- and 0.33% Cr-steels than on 0.406% Cr-steel and 2.5% Cr-1% Mo steel. The crack growth rate on 0.33% Cr-steel was the smallest over the tested potential range. A higher temperature of the electrolyte led to a higher rate of electrochemical dissolution of steel and a higher susceptibility of steel to cracking, as shown by the positive increase of the electrochemical potential. An increase in Cr-content in the steel is predicted to reduce the corrosion rate of steel at high temperatures. However, this increase in Cr-content is predicted not to reduce the susceptibility of steel to cracking at high temperatures. (author)

  15. Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Jinling, E-mail: Jinling_Gao@yeah.net; Yao, Wenjuan, E-mail: wj_yao@yeah.net; Ma, Yudong

    2016-08-15

    Graphical abstract: An analytical model and a numerical procedure are developed to study the mechanical response of IG-110 graphite bricks in HTGR subjected to high temperature and irradiation. The calculation results show great accordance with each other. Rational suggestions on the calculation and design of the IG-110 graphite structure are proposed based on the sensitivity analyses including temperature, irradiation dimensional change, creep and Poisson’s ratio. - Highlights: • Analytical solution of stress and displacement of IG-110 graphite components in HTGR. • Finite element procedure developed for stress analysis of HTGR graphite component. • Parameters analysis of mechanical response of graphite components during the whole life of the reflector. - Abstract: Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson

  16. An explanation of the irreversibility behavior in the highly- anisotropic high-temperature superconductors

    International Nuclear Information System (INIS)

    Gray, K.E.; Kim, D.H.

    1991-01-01

    The wide temperature range of the reversible, lossy state of the new high-temperature superconductors in a magnetic field was recognized soon after their discovery. This behavior, which had gone virtually undetected in conventional superconductors, has generated considerable interest, both for a fundamental understanding of the HTS and because it degrades the performance of HTS for finite-field applications. We show that recently proposed explanation of this behavior for the highly-anisotropic high-temperature superconductors, as a dimensional crossover of the magnetic vortices, is strongly supported by recent experiments on a Bi 2 Sr 2 CaCu 2 O x single crystal using the high-Q mechanical oscillator techniques

  17. Influence of variations in creep curve on creep behavior of a high-temperature structure

    International Nuclear Information System (INIS)

    Hada, Kazuhiko

    1986-01-01

    It is one of the key issues for a high-temperature structural design guideline to evaluate the influence of variations in creep curve on the creep behavior of a high-temperature structure. In the present paper, a comparative evaluation was made to clarify such influence. Additional consideration was given to the influence of the relationship between creep rupture life and minimum creep rate, i.e., the Monkman-Grant's relationship, on the creep damage evaluation. The consideration suggested that the Monkman-Grant's relationship be taken into account in evaluating the creep damage behavior, especially the creep damage variations. However, it was clarified that the application of the creep damage evaluation rule of ASME B and P.V. Code Case N-47 to the ''standard case'' which was predicted from the average creep property would predict the creep damage on the safe side. (orig./GL)

  18. High temperature strength and aging behavior of 12%Cr-15%Mn austenitic steels

    International Nuclear Information System (INIS)

    Miyahara, Kazuya; Bae, Dong-Su; Sakai, Hidenori; Hosoi, Yuzo

    1993-01-01

    High Mn-Cr austenitic steels are still considered to be an important high temperature structural material from the point of view of reduced radio-activation. The objective of the present study is to make a fundamental research of mechanical properties and microstructure of 12%Cr-15%Mn austenitic steels. Especially the effects of alloying elements of V and Ti on the mechanical properties and microstructure evolution of high Mn-Cr steels were studied. Precipitation behaviors of carbides, nitrides and σ phase are investigated and their remarkable effects on the high temperature strength are found. The addition of V was very effective for strengthening the materials with the precipitation of fine VN. Ti was also found to be beneficial for the improvement of high temperature strength properties. The results of high temperature strengths of the 12Cr-15Mn austenitic steels were compared with those of the other candidate and/or reference materials, for example, JFMS (modified 9Cr-2Mo ferritic stainless steel) and JPCAs (modified 316 austenitic stainless steels). (author)

  19. High temperature mechanical properties and surface fatigue behavior improving of steel alloy via laser shock peening

    International Nuclear Information System (INIS)

    Ren, N.F.; Yang, H.M.; Yuan, S.Q.; Wang, Y.; Tang, S.X.; Zheng, L.M.; Ren, X.D.; Dai, F.Z.

    2014-01-01

    Highlights: • The properties of 00C r 12 were improved by laser shock processing. • A deep layer of residual compressive stresses was introduced. • Fatigue life was enhanced about 58% at elevated temperature up to 600 °C. • The pinning effect is the reason of prolonging fatigue life at high temperature. - Abstract: Laser shock peening was carried out to reveal the effects on ASTM: 410L 00C r 12 microstructures and fatigue resistance in the temperature range 25–600 °C. The new conception of pinning effect was proposed to explain the improvements at the high temperature. Residual stress was measured by X-ray diffraction with sin 2 ψ method, a high temperature extensometer was utilized to measure the strain and control the strain signal. The grain and precipitated phase evolutionary process were observed by scanning electron microscopy. These results show that a deep layer of compressive residual stress is developed by laser shock peening, and ultimately the isothermal stress-controlled fatigue behavior is enhanced significantly. The formation of high density dislocation structure and the pinning effect at the high temperature, which induces a stronger surface, lower residual stress relaxation and more stable dislocation arrangement. The results have profound guiding significance for fatigue strengthening mechanism of components at the elevated temperature

  20. Electrical treeing behaviors in silicone rubber under an impulse voltage considering high temperature

    Science.gov (United States)

    Yunxiao, ZHANG; Yuanxiang, ZHOU; Ling, ZHANG; Zhen, LIN; Jie, LIU; Zhongliu, ZHOU

    2018-05-01

    In this paper, work was conducted to reveal electrical tree behaviors (initiation and propagation) of silicone rubber (SIR) under an impulse voltage with high temperature. Impulse frequencies ranging from 10 Hz to 1 kHz were applied and the temperature was controlled between 30 °C and 90 °C. Experimental results show that tree initiation voltage decreases with increasing pulse frequency, and the descending amplitude is different in different frequency bands. As the pulse frequency increases, more frequent partial discharges occur in the channel, increasing the tree growth rate and the final shape intensity. As for temperature, the initiation voltage decreases and the tree shape becomes denser as the temperature gets higher. Based on differential scanning calorimetry results, we believe that partial segment relaxation of SIR at high temperature leads to a decrease in the initiation voltage. However, the tree growth rate decreases with increasing temperature. Carbonization deposition in the channel under high temperature was observed under microscope and proven by Raman analysis. Different tree growth models considering tree channel characteristics are proposed. It is believed that increasing the conductivity in the tree channel restrains the partial discharge, holding back the tree growth at high temperature.

  1. New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M.; Vlahovic, L.; Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe, (Germany); Sadli, M.; Failleau, G. [Laboratoire Commun de Metrologie, LNE-Cnam, Saint-Denis, (France); Fuetterer, M.; Lapetite, J.M. [European Commission, Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, (Netherlands); Fourrez, S. [Thermocoax, 8 rue du pre neuf, F-61100 St Georges des Groseillers, (France)

    2015-07-01

    Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand high temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)

  2. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

  3. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    International Nuclear Information System (INIS)

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

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

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

    Science.gov (United States)

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

    1988-07-01

    Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

  5. A Status of Art-Report on the Fission Products Behavior Released from Spent Fuel at High Temperature Conditions

    International Nuclear Information System (INIS)

    Park, Geun Il; Kim, J. H.; Lee, J. W.

    2003-04-01

    The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests

  6. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  7. High temperature oxidation and corrosion behavior of Ni-base superalloy in He environment

    International Nuclear Information System (INIS)

    Lee, Gyoeng Geun; Park, Ji Yeon; Jung, Su jin

    2010-11-01

    Ni-base superalloy is considered as a IHX (Intermediate Heat Exchanger) material for VHTR (Very High Temperature Gas-Cooled Reactor). The helium environment in VHTR contains small amounts of impure gases, which cause oxidation, carburization, and decarburization. In this report, we conducted the literature survey about the high temperature behavior of Ni-base superalloys in air and He environments. The basic information of Ni-base superalloy and the basic metal-oxidation theory were briefly stated. The He effect on the corrosion of Ni-base superalloy was also summarized. This works would provide a brief suggestion for the next research topic for the application of Ni-base superalloy to VHTR

  8. Strain-rate dependent fatigue behavior of 316LN stainless steel in high-temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Jibo [CAS Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Wu, Xinqiang, E-mail: xqwu@imr.ac.cn [CAS Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Han, En-Hou; Ke, Wei; Wang, Xiang [CAS Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Sun, Haitao [Nuclear and Radiation Safety Center, SEPA, Beijing 100082 (China)

    2017-06-15

    Low cycle fatigue behavior of forged 316LN stainless steel was investigated in high-temperature water. It was found that the fatigue life of 316LN stainless steel decreased with decreasing strain rate from 0.4 to 0.004 %s{sup −1} in 300 °C water. The stress amplitude increased with decreasing strain rate during fatigue tests, which was a typical characteristic of dynamic strain aging. The fatigue cracks mainly initiated at pits and slip bands. The interactive effect between dynamic strain aging and electrochemical factors on fatigue crack initiation is discussed. - Highlights: •The fatigue lives of 316LN stainless steel decrease with decreasing strain rate. •Fatigue cracks mainly initiated at pits and persistent slip bands. •Dynamic strain aging promoted fatigue cracks initiation in high-temperature water.

  9. Progress report on irradiation experiment on small size specimens in high temperature flux module

    Energy Technology Data Exchange (ETDEWEB)

    Ramesh, M.; Jacquet, P.; Chaouadi, R.

    2011-02-15

    This report describes the progress made in IFREC/DEMO Research and Development Program during the year 2010 at SCK/CEN. This task is part of demonstrating the possibility to irradiate small specimens in the HFTM modules that will be used in DEMO. Different small specimens of three candidate materials of DEMO fusion reactor will be irradiated with the objective of validating the specimen geometry and size to reliably characterize the mechanical properties of unirradiated and in future of irradiated materials.

  10. Mechanical properties of Mo and TZM alloy neutron-irradiated at high temperatures

    International Nuclear Information System (INIS)

    Ueda, Kazukiyo; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori

    1997-01-01

    This work reports the mechanical properties of irradiated molybdenum (Mo) and its alloy, TZM. Recrystallized and stress-relieved specimens were irradiated at five temperatures between 373 and 800degC in FFTF/MOTA to fluence levels of 6.8 to 34 dpa. Irradiation embrittlement and hardening were evaluated by three-point bend test and Vickers hardness test, respectively. Stress-relieved materials showed the enough ductility even after high fluence irradiation. The role of layered structure of stress-relieved specimen was discussed. (author)

  11. Transmission electron-microscopic studies of structural changes in polycrystalline graphite after high temperature irradiation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Gurovich, B.A.; Shtrombakh, Ya.I.; Karpukhin, V.I.

    1985-01-01

    Transmission electron-microscopic investigation of polycrystalline graphite before and after irradiation is carried out. The direct use of graphite samples after ion thinning, as an inquiry subject is the basic peculiarity of the work. Main structural components of MPG-6 graphite before and after irradiation are revealed, the structural mechanism of the reactor graphite destruction under irradiation is demonstrated. The mean values of L αm and L cm crystallite dimensions are determined. Radiation defects, occuring in some crystallites after irradiation are revealed by the dark-field electron microscopy method

  12. Isothermal oxidation behavior of ternary Zr-Nb-Y alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Prajitno, Djoko Hadi, E-mail: djokohp@batan.go.id [Research Center for Nuclear Materials and Radiometry, Jl. Tamansari 71, Bandung 40132 (Indonesia); Soepriyanto, Syoni; Basuki, Eddy Agus [Metallurgy Engineering, Institute Technology Bandung, Jl. Ganesha 10, Bandung 40132 (Indonesia); Wiryolukito, Slameto [Materials Engineering, Institute Technology Bandung, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2014-03-24

    The effect of yttrium content on isothermal oxidation behavior of Zr-2,5%Nb-0,5%Y, Zr-2,5%Nb-1%Y Zr-2,5%Nb-1,5%Y alloy at high temperature has been studied. High temperature oxidation carried out at tube furnace in air at 600,700 and 800°C for 1 hour. Optical microscope is used for microstructure characterization of the alloy. Oxidized and un oxidized specimen was characterized by x-ray diffraction. In this study, kinetic oxidation of Zr-2,5%Nb with different Y content at high temperature has also been studied. Characterization by optical microscope showed that microstructure of Zr-Nb-Y alloys relatively unchanged and showed equiaxed microstructure. X-ray diffraction of the alloys depicted that the oxide scale formed during oxidation of zirconium alloys is monoclinic ZrO2 while unoxidised alloy showed two phase α and β phase. SEM-EDS examination shows that depletion of Zr composition took place under the oxide layer. Kinetic rate of oxidation of zirconium alloy showed that increasing oxidation temperature will increase oxidation rate but increasing yttrium content in the alloys will decrease oxidation rate.

  13. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Windl, Wolfgang [The Ohio State Univ., Columbus, OH (United States); Dickerson, Bryan [Luna Innovations, Inc. (United States)

    2013-01-03

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica's optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

  14. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to High Temperature

    International Nuclear Information System (INIS)

    Blue, Thomas; Windl, Wolfgang; Dickerson, Bryan

    2013-01-01

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica'@@s optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and

  15. High-temperature method of rapid separation of In-111 from irradiated silver targets

    International Nuclear Information System (INIS)

    Mazgaj, Z.; Kolaczkowski, A.; Mikulski, J.; Novgorodov, A.F.; Zielinski, A.; Joint Inst. for Nuclear Research, Dubna

    1990-01-01

    A high-temperature method of separation of In-111 from α-particle activated silver targets was developed. The separation is carried out under reduced pressure, in the atmosphere of HCl and H 2 O vapours. Indium-111, adsorbed on a quartz collector, is washed out quantitatively with 0.1 N HCl. The contaminant, Cd-109 (product of decay of In-109), is removed from the preparation by means of ion-exchange chromatography. 4 tabs., 6 refs. (author)

  16. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  17. The tensile behavior of Ti36Ni49Hf15 high temperature shape memory alloy

    International Nuclear Information System (INIS)

    Wang, Y.Q.; Zheng, Y.F.; Cai, W.; Zhao, L.C.

    1999-01-01

    Recently, ternary Ti-Ni-Hf alloys have attracted great interest in the field of high temperature shape memory materials research and development. Extensive studies have been made on its manufacture process, constitutional phases, phase transformation behavior, the structure, substructure and interface structure of martensite and the precipitation behavior during ageing. Yet up to date there is no report about the fundamental mechanical properties of Ti-Ni-Hf alloys, such as the stress-strain data, the variation laws of the yield strength and elongation with the temperature. In the present study, tensile tests at various temperatures are employed to investigate the mechanical behavior of Ti-Ni-Hf alloy with different matrix structures, from full martensite to full parent phase structure, with the corresponding deformation mechanism discussed

  18. Microscale interfacial behavior at vapor film collapse on high-temperature particle surface

    International Nuclear Information System (INIS)

    Abe, Yutaka; Tochio, Daisuke

    2009-01-01

    It has been pointed out that vapor film on a premixed high-temperature droplet surface should be collapsed to trigger vapor explosion. Thus, it is important to clarify the micromechanism of vapor film collapse behavior for the occurrence of vapor explosion. In the present study, microscale vapor-liquid interface behavior upon vapor film collapse caused by an external pressure pulse is experimentally observed and qualitatively analyzed. In the analytical investigation, interfacial temperature and interface movement were estimated with heat conduction analysis and visual data processing technique. Results show that condensation can possibly occur at the vapor-liquid interface when the pressure pulse arrived. That is, this result indicates that the vapor film collapse behavior is dominated not by fluid motion but by phase change. (author)

  19. Incoloy 800 anodic behavior in sulfate and chloride solutions at high temperature

    International Nuclear Information System (INIS)

    Lafont, C.; Alvarez, M.G.

    1992-01-01

    The anodic behavior and pitting corrosion resistance of Incoloy 800 in concentrated aqueous chloride and sulphate solutions has been studied by means of electrochemical techniques. The effect of different environmental variables, such as temperature (in the 100 0 C to 280 0 C range) and sulphate ion concentration (0.02 M to 2 M), was evaluated. In another set of experiments, the influence of sulphate ions additions on the pitting resistance and pitting morphology of Incoloy 800 in chloride solutions at high temperature was also examined. (author)

  20. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Vasudevan, Vijay; Carroll, Laura; Sham, Sam

    2015-01-01

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  1. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, Vijay [Univ. of Cincinnati, OH (United States); Carroll, Laura [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-06

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  2. Experimental study under uniaxial cyclic behavior at room and high temperature of 316L stainless steel

    International Nuclear Information System (INIS)

    Kang Guozheng; Gao Qing; Yang Xianjie; Sun Yafang

    2001-01-01

    An experimental study was carried out of the cyclic properties of 316L stainless steel subjected to uniaxial strain and stress at room and high temperature. The effects of cyclic strain amplitude, temperature and their histories on the cyclic deformation behavior of 316L stainless steel are investigated. And, the influences of stress amplitude, mean stress, temperature and their histories on ratcheting are also analyzed. It is shown that either uniaxial cyclic property under cyclic strain or ratcheting under asymmetric uniaxial cyclic stress depends not only on the current temperature and loading state, but also on the previous temperature and loading history. Some significant results are obtained

  3. High Temperature Tensile Properties of Unirradiated and Neutron Irradiated 20 Cr-35 Ni Austenitic Steel

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R B; Solly, B

    1966-12-15

    The tensile properties of an unirradiated and neutron irradiated (at 40 deg C) 20 % Cr, 35 % Ni austenitic steel have been studied at 650 deg C, 750 deg C and 820 deg C. The tensile elongation and mode of fracture (transgranular) of unirradiated specimens tested at room temperature and 650 deg C are almost identical. At 750 deg C and 820 deg C the elongation decreases considerably and a large part of the total elongation is non-uniform. Furthermore, the mode of fracture at these temperatures is intergranular and microscopic evidence suggests that fracture is caused by formation and linkup of grain boundary cavities. YS and UTS decrease monotonically with temperature. Irradiated specimens show a further decrease in ductility and an increase in the tendency to grain boundary cracking. Irradiation has no significant effect on the YS, but the UTS are reduced. The embrittlement of the irradiated specimens is attributed to the presence of He and Li atoms produced during irradiation and the possible mechanisms are discussed. Prolonged annealing of irradiated and unirradiated specimens at 650 deg C appears to have no significant effect on tensile properties.

  4. High-temperature irradiation effects on mechnical properties of HTGR graphites

    International Nuclear Information System (INIS)

    Oku, Tatsuo; Eto, Motokuni; Fujisaki, Katsuo

    1978-04-01

    The irradiation effects on stress-strain relation, Young's modulus, tensile strength, bending strength and compressive strength of HTGR graphites were studied in irradiation temperature ranges of 200 - 300 0 C and 800 - 1400 0 C and in neutron fluences up to 7.4 x 10 20 n/cm 2 and 3 x 10 21 n/cm 2 (> 0.18 MeV). Fracture criteria and strain energy to fracture of the unirradiated and the irradiated graphites were also examined. (1) Neutron fluence dependences are similar in Young's modulus, tensile strength and bending strength. (2) The change of compressive strength and of tensile and bending strengths with neutron fluence differ; the former varies with graphite kind. (3) At lower irradiation temperatures the bending fracture strain energy decreases with increasing neutron fluence and at higher irradiation temperatures it increases. (4) The fracture criteria of graphites deviates from the constant strain energy theory (α = 0.5) and the constant strain theory (α = 1), shifting from α asymptotically equals 0.5 to α asymptotically equals 1 with increasing irradiation temperature. (auth.)

  5. Effect of High Temperature on the Tensile Behavior of CFRP and Cementitious Composites

    Science.gov (United States)

    Toutanji, Houssam A.

    1999-01-01

    Concrete and other composite manufacturing processes are continuing to evolve and become more and more suited for use in non-Earth settings such as the Moon and Mars. The fact that structures built in lunar environments would experience a range of effects from temperature extremes to bombardment by micrometeorites and that all the materials for concrete production exist on the Moon means that concrete appears to be the most feasible building material. it can provide adequate shelter from the harshness of the lunar environment and at the same time be a cost effective building material. With a return to the Moon planned by NASA to occur after the turn of the century, it will be necessary to include concrete manufacturing as one of the experiments to be conducted in one of the coming missions. Concrete's many possible uses and possibilities for manufacturing make it ideal for lunar construction. The objectives of this research are summarized as follows: i) study the possibility of concrete production on the Moon or other planets, ii) study the effect of high temperature on the tensile behavior of concrete, and iii) study the effect of high temperature on the tensile behavior of carbon fiber reinforced with inorganic polymer composites. Literature review indicates that production of concrete on the Moon or other planets is feasible using the indigenous materials. Results of this study has shown that both the tensile strength and static elastic modulus of concrete decreased with a rise in temperature from 200 to 500 C. The addition of silica fume to concrete showed higher resistance to high temperatures. Carbon fiber reinforced inorganic polymer (CFRIP) composites seemed to perform well up to 300 C. However, a significant reduction in strength was observed of about 40% at 400 C and up to 80% when the specimens were exposed to 700 C.

  6. Lipid and carotenoid production by Rhodotorula glutinis under irradiation/high-temperature and dark/low-temperature cultivation.

    Science.gov (United States)

    Zhang, Zhiping; Zhang, Xu; Tan, Tianwei

    2014-04-01

    The capacity of lipid and carotenoid production by Rhodotorula glutinis was investigated under different irradiation conditions, temperatures and C/N ratios. The results showed that dark/low-temperature could enhance lipid content, while irradiation/high-temperature increased the yields of biomass and carotenoid. The optimum C/N ratio for production was between 80 and 100. A two-stage cultivation strategy was used for lipid and carotenoid production in a 5L fermenter. In the first stage, the maximum biomass reached 28.1g/L under irradiation/high-temperature. Then, the cultivation condition was changed to dark/low-temperature, and C/N ratio was adjusted to 90. After the second stage, the biomass, lipid content and carotenoid reached 86.2g/L, 26.7% and 4.2mg/L, respectively. More significantly, the yields of biomass and lipid were 43.1% and 11.5%, respectively. Lipids contained 79.7% 18C and 16.8% 16C fatty acids by GC analysis. HPLC quantified the main carotenoids were β-carotene (68.4%), torularhodin (21.5%) and torulene (10.1%). Crown Copyright © 2014. Published by Elsevier Ltd. All rights reserved.

  7. Thermal Cycling and High Temperature Reverse Bias Testing of Control and Irradiated Gallium Nitride Power Transistors

    Science.gov (United States)

    Patterson, Richard L.; Boomer, Kristen T.; Scheick, Leif; Lauenstein, Jean-Marie; Casey, Megan; Hammoud, Ahmad

    2014-01-01

    The power systems for use in NASA space missions must work reliably under harsh conditions including radiation, thermal cycling, and exposure to extreme temperatures. Gallium nitride semiconductors show great promise, but information pertaining to their performance is scarce. Gallium nitride N-channel enhancement-mode field effect transistors made by EPC Corporation in a 2nd generation of manufacturing were exposed to radiation followed by long-term thermal cycling and testing under high temperature reverse bias conditions in order to address their reliability for use in space missions. Result of the experimental work are presented and discussed.

  8. The High Temperature Tensile and Creep Behaviors of High Entropy Superalloy.

    Science.gov (United States)

    Tsao, Te-Kang; Yeh, An-Chou; Kuo, Chen-Ming; Kakehi, Koji; Murakami, Hideyuki; Yeh, Jien-Wei; Jian, Sheng-Rui

    2017-10-04

    This article presents the high temperature tensile and creep behaviors of a novel high entropy alloy (HEA). The microstructure of this HEA resembles that of advanced superalloys with a high entropy FCC matrix and L1 2 ordered precipitates, so it is also named as "high entropy superalloy (HESA)". The tensile yield strengths of HESA surpass those of the reported HEAs from room temperature to elevated temperatures; furthermore, its creep resistance at 982 °C can be compared to those of some Ni-based superalloys. Analysis on experimental results indicate that HESA could be strengthened by the low stacking-fault energy of the matrix, high anti-phase boundary energy of the strengthening precipitate, and thermally stable microstructure. Positive misfit between FCC matrix and precipitate has yielded parallel raft microstructure during creep at 982 °C, and the creep curves of HESA were dominated by tertiary creep behavior. To the best of authors' knowledge, this article is the first to present the elevated temperature tensile creep study on full scale specimens of a high entropy alloy, and the potential of HESA for high temperature structural application is discussed.

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

    Science.gov (United States)

    Yamakawa, K.; Shimomura, Y.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  11. Multiple cracks initiation and propagation behavior of stainless steel in high temperature water environment

    International Nuclear Information System (INIS)

    Kamaya, Masayuki; Chiba, Goro; Nakajima, Nobuo; Totsuka, Nobuo

    2001-01-01

    Environmentally assisted crack initiation behavior is greatly affected by applied stress and environmental factors, such as water temperature, contained impurities and so on. On the other hand, crack initiation behavior also influences crack propagation. A typical example of this influence can be observed as the interference effects of multiple cracks, such as the coalescence of approaching crack tips or the arrest phenomena in the relaxation zone of an adjacent crack. To understand these effects of crack initiation on crack propagation behavior is very important to predict the lifetime of components, in which quite a few cracks tend to occur. This study aimed at revealing the crack initiation behavior and the influence of this behavior on propagation. At first, to evaluate the effect of applied stress on crack initiation behavior, sensitized stainless steel was subjected to a four-point bending test in a high temperature water environment at the constant potentials of ECP +50 mV and ECP +150 mV. Secondly, a crack initiation and growth simulation model was developed, in which the interference effect of multiple cracks is evaluated by the finite element method, based on the experimental results. Using this model, the relationship between crack initiation and propagation was studied. From the model, it was revealed that the increasing number of the cracks accelerates crack propagation and reduces life. (author)

  12. Utilization of High-Temperature Slags From Metallurgy Based on Crystallization Behaviors

    Science.gov (United States)

    Sun, Yongqi; Zhang, Zuotai

    2018-05-01

    Here, following the principle of modifying crystallization behaviors, including avoidance and optimization, we review recent research on the utilization of hot slags. Because of the high-temperature property (1450-1650°C), the utilization of hot slags are much different from that of other wastes. We approach this issue from two main directions, namely, material recycling and heat utilization. From the respect of material recycling, the utilization of slags mainly follows total utilization and partial utilization, whereas the heat recovery from slags follows two main paths, namely, physical granulation and chemical reaction. The effective disposal of hot slags greatly depends on clarifying the crystallization behaviors, and thus, we discuss some optical techniques and their applicable scientific insights. For the purpose of crystallization avoidance, characterizing the glass-forming ability of slags is of great significance, whereas for crystallization modification, the selection of chemical additives and control of crystallization conditions comprise the central routes.

  13. Oxidation behavior of TD-NiCr in a dynamic high temperature environment

    Science.gov (United States)

    Tenney, D. R.; Young, C. T.; Herring, H. W.

    1974-01-01

    The oxidation behavior of TD-NiCr has been studied in static and high-speed flowing air environments at 1100 and 1200 C. It has been found that the stable oxide morphologies formed on the specimens exposed to the static and dynamic environments were markedly different. The faceted crystal morphology characteristic of static oxidation was found to be unstable under high-temperature, high-speed flow conditions and was quickly replaced by a porous NiO 'mushroom' type structure. Also, it was found that the rate of formation of CrO3 from Cr2O3 was greatly enhanced by high gas velocity conditions. The stability of Cr2-O3 was found to be greatly improved by the presence of an outer NiO layer, even though the NiO layer was very porous. An oxidation model is proposed to explain the observed microstructures and overall oxidation behavior of TD-NiCr alloys.

  14. High temperature homogenization improves impact toughness of vitamin E-diffused, irradiated UHMWPE.

    Science.gov (United States)

    Oral, Ebru; O'Brien, Caitlin; Doshi, Brinda; Muratoglu, Orhun K

    2017-06-01

    Diffusion of vitamin E into radiation cross-linked ultrahigh molecular weight polyethylene (UHMWPE) is used to increase stability against oxidation of total joint implant components. The dispersion of vitamin E throughout implant preforms has been optimized by a two-step process of doping and homogenization. Both of these steps are performed below the peak melting point of the cross-linked polymer (homogenization of antioxidant-doped, radiation cross-linked UHMWPE could improve its toughness. We found that homogenization at 300°C for 8 h resulted in an increase in the impact toughness (74 kJ/m 2 compared to 67 kJ/m 2 ), the ultimate tensile strength (50 MPa compared to 43 MPa) and elongation at break (271% compared to 236%). The high temperature treatment did not compromise the wear resistance or the oxidative stability as measured by oxidation induction time. In addition, the desired homogeneity was achieved at a much shorter duration (8 h compared to >240 h) by using high temperature homogenization. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc. J Orthop Res 35:1343-1347, 2017. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc.

  15. Study on structural recovery of graphite irradiated with swift heavy ions at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pellemoine, F., E-mail: pellemoi@frib.msu.edu [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Avilov, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Bender, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Ewing, R.C. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Fernandes, S. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Lang, M. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996-2300 (United States); Li, W.X. [Dept. of Geological Sciences, Stanford University, Stanford, CA 94305-2115 (United States); Mittig, W. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Schein, M. [Facility for Rare Isotope Beams, Michigan State University, East Lansing, MI 48824 (United States); Severin, D. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Tomut, M. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Laboratory of Magnetism and Superconductivity, National Institute for Materials Physics NIMP, Bucharest (Romania); Trautmann, C. [Dept. of Materials Research, GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany); Dept. of Materials Science, Technische Universität Darmstadt, Darmstadt (Germany); and others

    2015-12-15

    Thin graphite foils bombarded with an intense high-energy (8.6 MeV/u) gold beam reaching fluences up to 1 × 10{sup 15} ions/cm{sup 2} lead to swelling and electrical resistivity changes. As shown earlier, these effects are diminished with increasing irradiation temperature. The work reported here extends the investigation of beam induced changes of these samples by structural analysis using synchrotron X-ray diffraction and transmission electron microscope. A nearly complete recovery from swelling at irradiation temperatures above about 1500 °C is identified.

  16. Molecular weight distribution of electron and γ-ray irradiated PEEK measured by very high temperature GPC

    International Nuclear Information System (INIS)

    Nakahara, H.

    1996-01-01

    Poly(ether ether ketone)(PEEK) films were irradiated with electron beam in air and in helium. Gel fractions of the PEEK samples were determined as the ratio of the weight of insoluble fraction/total weigh by extracting the samples with 1-chloronaphthalene (1-CN) at 260degC. While unirradiated PEEK samples were dissolved in 1-CN completely, PEEK samples highly (10 - 50 MGy) irradiated in air were almost insoluble in the solvent. The weight-average molecular weight M w of soluble fractions of the samples were measured by very high temperature gel permeation chromatography (VHTGPC): it was found that the M w decreases with increasing dose. On the other hand, PEEK samples irradiated in helium gave gel fractions at lower doses (0 - 5 MGy) than in air. The PEEK films were also irradiated with 60 Co γ-rays in the dose range, i.e. from 0 to 5 MGy. The γ-irradiated PEEK samples were completely dissolved in 1-CN at 260degC. Their M w measured by VHTGPC decreases with increasing dose. (author)

  17. Fusion neutron irradiation induced ordering and defect production in Cu3Au at high temperatures

    International Nuclear Information System (INIS)

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

    1987-08-01

    We irradiate three Cu 3 Au alloys different degrees of initial long-range order at temperatures between 300K and 434K. The resistivity of samples is monitored during irradiation and related to the long-term order parameter by the Muto relation. The results show that the ordering rate, which is proportional to the concentration of freely migrating vacancies, increases at the beginning and then decreases later with fluence. The decrease is a result of the continuous production of sinks in the form of dislocation loops. The effect of sinks on vacancy annihilation in some cases causes a reversed temperature dependence of ordering rate. The free vacancy production rate and the rate of sink production are determined using an ordering kinetics theory. The results of the 14 MeV neutron irradiations are compared to those obtained in other neutron spectra and particle irradiations. The estimated free vacancy production rate is also compared to the primary defect production rate measured at 4.2K in disordered samples

  18. High Temperature Degradation Behavior and its Mechanical Properties of Inconel 617 alloy for Intermediate Heat Exchanger of VHTR

    International Nuclear Information System (INIS)

    Jo, Tae Sun; Kim, Se Hoon; Kim, Young Do; Park, Ji Yeon

    2008-01-01

    Inconel 617 alloy is a candidate material of intermediate heat exchanger (IHX) and hot gas duct (HGD) for very high temperature reactor (VHTR) because of its excellent strength, creep-rupture strength, stability and oxidation resistance at high temperature. Among the alloying elements in Inconel 617, chromium (Cr) and aluminum (Al) can form dense oxide that act as a protective surface layer against degradation. This alloy supports severe operating conditions of pressure over 8 MPa and 950 .deg. C in He gas with some impurities. Thus, high temperature stability of Inconel 617 is very important. In this work, the oxidation behavior of Inconel 617 alloy was studied by exposure at high temperature and was discussed the high temperature degradation behavior with microstructural changes during the surface oxidation

  19. High temperature deformation behavior and microstructural evolutions of a high Zr containing WE magnesium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Asqardoust, Sh.; Zarei-Hanzaki, A. [School of Metallurgical & Materials Engineering, University of Tehran, Tehran (Iran, Islamic Republic of); Fatemi, S.M., E-mail: mfatemi@ut.ac.ir [Shahid Rajaee Teacher Training University, Tehran (Iran, Islamic Republic of); Moradjoy-Hamedani, M. [School of Metallurgical & Materials Engineering, University of Tehran, Tehran (Iran, Islamic Republic of)

    2016-06-05

    Magnesium alloys containing RE elements (WE grade) are considered as potential materials for high temperature structural applications. To this end, it is crucial to study the flow behavior and the microstructural evolution of these alloys at high temperatures. In present work, the hot compression testing was employed to investigate the deformation behavior of a rolled WE54 magnesium alloy at elevated temperatures. The experimental material failed to deform to target strain of 0.6 at 250 and 300 °C, while the straining was successfully performed at 350 °C. A flow softening was observed at 350 °C, which was related to the depletion of RE strengthener elements, particularly Y atoms, from the solid solution and dynamic precipitation of β phases. It was suggested that the Zener pinning effect of the latter precipitates might retard the occurrence of dynamic recrystallization. As the temperature increased to 450 and 500 °C, the RE elements dissolved in the matrix and thus dynamic recrystallization could considerably progress in the microstructure. The comparative study of specimens cut along transverse ad normal direction (TD and ND specimens) implied that the presence of RE elements might effectively reduce the yield anisotropy in WE54 rolled alloy. Microstructural observations indicated a higher fraction of dynamically-recrystallized grains for the ND specimens. This was discussed relying on the different shares of deformation mechanism during compressing the TD and ND specimens. - Highlights: • Deformation behavior of a high Zr WE alloy was addressed at low strain rate. • Dynamic precipitation was realized at 350 °C. • The occurrence of DRX was retarded due to Zener pinning effect. • A higher DRX fraction was obtained in ND specimens comparing with TD ones.

  20. SCC growth behavior of stainless steel weld heat-affected zone in hydrogenated high temperature water

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Terachi, Takumi; Miyamoto, Tomoki; Arioka, Koji

    2010-01-01

    It is known that the SCC growth rate of stainless steels in high-temperature water is accelerated by cold-work (CW). The weld heat-affected-zone (HAZ) of stainless steels is also deformed by weld shrinkage. However, only little have been reported on the SCC growth of weld HAZ of SUS316 and SUS304 in hydrogenated high-temperature water. Thus, in this present study, SCC growth experiments were performed using weld HAZ of stainless steels, especially to obtain data on the dependence of SCC growth on (1) temperature and (2) hardness in hydrogenated water at temperatures from 250degC to 340degC. And then, the SCC growth behaviors were compared between weld HAZ and CW stainless steels. The following results have been obtained. Significant SCC growth were observed in weld HAZ (SUS316 and SUS304) in hydrogenated water at 320degC. The SCC growth rates of the HAZ are similar to that of 10% CW non-sensitized SUS316, in accordance with that the hardness of weld HAZ is also similar to that of 10% CW SUS316. Temperature dependency of SCC growth of weld HAZ (SUS316 and SUS304) is also similar to that of 10% CW non-sensitized SUS316. That is, no significant SCC were observed in the weld HAZ (SUS316 and SUS304) in hydrogenated water at 340degC. This suggests that SCC growth behaviors of weld HAZ and CW stainless steels are similar and correlated with the hardness or yield strength of the materials, at least in non-sensitized regions. And the similar temperature dependence between the HAZ and CW stainless steels suggests that the SCC growth behaviors are also attributed to the common mechanism. (author)

  1. Study on the surface sulfidization behavior of smithsonite at high temperature

    Science.gov (United States)

    Lv, Jin-fang; Tong, Xiong; Zheng, Yong-xing; Xie, Xian; Wang, Cong-bing

    2018-04-01

    Surface sulfidization behavior of smithsonite at high temperature was investigated by X-ray powder diffractometer (XRD) along with thermodynamic calculation, X-ray photoelectron spectroscopy (XPS) and electron probe microanalysis (EPMA). The XRD and thermodynamic analyses indicated that the smithsonite was decomposed into zincite at high temperatures. After introducing a small amount of pyrite, artificial sulfides were formed at surface of the obtained zincite. The XPS analyses revealed that the sulfide species including zinc sulfide and zinc disulfide were generated at the zincite surface. The EPMA analyses demonstrated that the film of sulfides was unevenly distributed at the zincite surface. The average concentration of elemental sulfur at the sample surface increased with increasing of pyrite dosage. A suitable mole ratio of FeS2 to ZnCO3 for the surface thermal modification was determined to be about 0.3. These findings can provide theoretical support for improving the process during which the zinc recovery from refractory zinc oxide ores is achieved by xanthate flotation.

  2. Metal release behavior of surface oxidized stainless steels into flowing high temperature pure water

    International Nuclear Information System (INIS)

    Fujiwara, Kazuo; Tomari, Haruo; Nakayama, Takenori; Shimogori, Kazutoshi; Ishigure, Kenkichi; Matsuura, Chihiro; Fujita, Norihiko; Ono, Shoichi.

    1987-01-01

    In order to clarify the effect of oxidation treatment of Type 304 SS on the inhibition of metal release into high temperature pure water, metal release rate of individual alloying element into flowing deionized water containing 50 ppb dissolved oxygen was measured as the function of exposure time on representative specimens oxidized in air and steam. The behavior of metal release was also discussed in relation to the structure of surface films. Among the alloying elements the amount of Fe ion, Cr ion and Fe crud in high temperature pure water tended to saturate with the exposure time and that of Ni ion and Co ion tended to increase monotonously with the exposure time for all specimens tested. And the treatment of steam-oxidation was the most effective to decrease the metal release of alloying elements and the treatment by air-oxidation also decreased the metal release. These tendencies were confirmed to correlate well with the structure of the surface films as it was in the results in the static autoclave test. (author)

  3. Behavior of radioactive organic iodide in an atmosphere of High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Nakashima, Mikio; Sagawa, Chiaki; Masaki, Nobuyuki; Hirabayashi, Takakuni; Aratono, Yasuyuki

    1990-06-01

    Formation and decomposition behavior of radioactive organic iodide have been studied in an atmosphere of High Temperature Gas-cooled Reactor (High Temperature Engineering Test Reactor, HTTR). Na 125 I was chosen for radioactive iodine source instead of CsI diffusing from coated fuel particles. Na 125 I adsorbed on graphite was heated in pure He and He containing O 2 or H 2 O atmosphere. The results obtained are as follows. It was proved that organic iodide was formed with organic radicals released from graphite even in He atmosphere. Thus, the interchange rate of inorganic iodide with organic iodide was remarkably decreased with prolonged preheat-treatment period at 1000degC. Organic iodide formed was easily decomposed by its recirculation into hot reaction tube kept at 900degC. When organic iodide was passed through powdered graphite bed, more than 70% was decomposed at 90degC. Oxygen and water vapour intermixed in He suppressed the interchange rate of inorganic iodide with organic iodide. These results suggest that organic iodide rarely exists in the pressure vessel under normal operating condition of HTTR, and, under hypothetical accident condition of HTTR, organic iodide fraction never exceeds the value used for a safety assessment of light water reactor. (author)

  4. Microstructure, tensile properties and fracture behavior of high temperature Al–Si–Mg–Cu cast alloys

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, A.M.A., E-mail: madel@uqac.ca [Center for Advanced Materials, Qatar University, Doha (Qatar); Department of Metallurgical and Materials Engineering, Faculty of Petroleum and Mining Engineering, Suez Canal University, Box 43721, Suez (Egypt); Samuel, F.H. [Université du Québec à Chicoutimi, Chicoutimi, QC, Canada G7H 2B1 (Canada); Al Kahtani, Saleh [Industrial Engineering Program, Mechanical Engineering Department, College of Engineering, Salman bin Abdulaziz University, Al Kharj (Saudi Arabia)

    2013-08-10

    The high temperature tensile behavior of 354 aluminum cast alloy was investigated in the presence of Zr and Ni. The cast alloys were given a solutionizing treatment followed by artificial aging at 190 °C for 2 h. High temperature tensile tests were conducted at various temperatures from 25 °C to 300 °C. Optical microscopy and electron probe micro-analyzer were used to study the microstructure of different intermetallic phases formed. The fractographic observations of fracture surface were analyzed by scanning electron microscopy to understand the fracture mechanism. The results revealed that the intermetallics phases of (Al, Si){sub 3}(Zr, Ti), Al{sub 3}CuNi and Al{sub 9}NiFe are the main feature in the microstructures of alloys with Zr and Ni additions. The results also indicated that the tensile strength of alloy decreases with an increase in temperature. The combined addition of 0.2 wt% Zr and 0.2 wt% Ni leads to a 30% increase in the tensile properties at 300 °C compared to the base alloy. Zr and Ni bearing phases played a vital role in the fracture mechanism of the alloys studied.

  5. Microstructure, tensile properties and fracture behavior of high temperature Al–Si–Mg–Cu cast alloys

    International Nuclear Information System (INIS)

    Mohamed, A.M.A.; Samuel, F.H.; Al Kahtani, Saleh

    2013-01-01

    The high temperature tensile behavior of 354 aluminum cast alloy was investigated in the presence of Zr and Ni. The cast alloys were given a solutionizing treatment followed by artificial aging at 190 °C for 2 h. High temperature tensile tests were conducted at various temperatures from 25 °C to 300 °C. Optical microscopy and electron probe micro-analyzer were used to study the microstructure of different intermetallic phases formed. The fractographic observations of fracture surface were analyzed by scanning electron microscopy to understand the fracture mechanism. The results revealed that the intermetallics phases of (Al, Si) 3 (Zr, Ti), Al 3 CuNi and Al 9 NiFe are the main feature in the microstructures of alloys with Zr and Ni additions. The results also indicated that the tensile strength of alloy decreases with an increase in temperature. The combined addition of 0.2 wt% Zr and 0.2 wt% Ni leads to a 30% increase in the tensile properties at 300 °C compared to the base alloy. Zr and Ni bearing phases played a vital role in the fracture mechanism of the alloys studied

  6. Straining electrode behavior and corrosion resistance of nickel base alloys in high temperature acidic solution

    International Nuclear Information System (INIS)

    Yamanaka, Kazuo

    1992-01-01

    Repassivation behavior and IGA resistance of nickel base alloys containing 0∼30 wt% chromium was investigated in high temperature acid sulfate solution. (1) The repassivation rate was increased with increasing chromium content. And so the amounts of charge caused by the metal dissolution were decreased with increasing chromium content. (2) Mill-annealed Alloy 600 suffered IGA at low pH environment below about 3.5 at the fixed potentials above the corrosion potential in 10%Na 2 SO 4 +H 2 SO 4 solution at 598K. On the other hand, thermally-treated Alloy 690 was hard to occur IGA at low pH environments which mill-annealed Alloy 600 occurred IGA. (3) It was considered that the reason, why nickel base alloys containing high chromium content such as Alloy 690 (60%Ni-30%Cr-10%Fe) had high IGA/SCC resistance in high temperature acidic solution containing sulfate ion, is due to both the promotion of the repassivation and the suppression of the film dissolution by the formation of the dense chromium oxide film

  7. Fission gas release and grain growth in THO2-UO2 fuel irradiated at high temperature

    International Nuclear Information System (INIS)

    Goldberg, I.; Waldman, L.A.; Giovengo, J.F.; Campbell, W.R.

    1979-01-01

    Data are presented on fission gas release and grain growth in ThO 2 -UO 2 fuels irradiated as part of the LWBR fuel element development program. These data for rods that experienced peak linear power outputs ranging from 15 to 22 KW/ft supplement fission gas release data previously reported for 51 rods containing ThO 2 and ThO 2 -UO 2 fuel irradiated at peak linear powers predominantly below 14 KW/ft. Fission gas release was relatively high (up to 15.0 percent) for the rods operated at high power in contrast to the relatively low fission gas release (0.1 to 5.2 percent) measured for the rods operated at lower power. Metallographic examination revealed extensive equiaxed grain growth in the fuel at the high power axial locations of the three rods

  8. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H; Blackstone, R [Stichting Energieonderzoek Centrum Nederland, Petten; Loelgen, R

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10/sup 21/ n cm/sup -2/ DNE in the temperature range 600 to 1200/sup 0/C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material.

  9. Irradiation-induced creep in fuel compacts for high-temperature reactor applications

    International Nuclear Information System (INIS)

    Veringa, H.; Blackstone, R.; Loelgen, R.

    1977-01-01

    Restrained shrinkage experiments at neutron fluences up to 3 x 10 21 n cm -2 DNE in the temperature range 600 to 1200 0 C were performed on three different dummy coated-particle fuel compacts in the high-flux reactor at Petten. The data were evaluated to obtain the steady-state radiation creep coefficient of the compacts. It was found that, for the materials investigated, the creep coefficient is temperature dependent, but no clear relationship with Young's modulus could be established. Under certain conditions this irradiation-induced plasticity influences the elastic properties, with the concomitant increase of the creep coefficient. This effect coincides with the formation and further opening up of cracks due to stresses caused by irradiation-induced shrinkage of matrix material. (author)

  10. Compaction of irradiated fuel can wastes by high temperature melting in cold crucibles

    International Nuclear Information System (INIS)

    Piccinato, R.; Ruty, J.P.; Caraballo, R.; Jacquet-Francillon, N.

    1993-01-01

    The fusion of hull wastes obtained from the reprocessing of various irradiated fuels is an alternative method to the cementation process used for the conditioning of such wastes. This new process, based on the direct fusion of hulls, has been carried out at CEA Marcoule with an inactive industrial prototype and qualified with an active laboratory prototype. The report shows the results obtained with the lab prototype on stainless steel and zircaloy hulls

  11. High temperature tensile testing of modified 9Cr-1Mo after irradiation with high energy protons

    International Nuclear Information System (INIS)

    Toloczko, M.B.; Hamilton, M.L.; Maloy, S.A.

    2003-01-01

    This study examines the effect of tensile test temperatures ranging from 50 to 600 deg. C on the tensile properties of a modified 9Cr-1Mo ferritic steel after high energy proton irradiation at about 35-67 deg. C to doses from 1 to 3 dpa and 9 dpa. For the specimens irradiated to doses between 1 and 3 dpa, it was observed that the yield strength and ultimate strength decreased monotonically as a function of tensile test temperature, whereas the uniform elongation (UE) remained at approximately 1% for tensile test temperatures up to 250 deg. C and then increased for tensile test temperatures up to and including 500 deg. C. At 600 deg. C, the UE was observed to be less than the values at 400 and 500 deg. C. UE of the irradiated material tensile tested at 400-600 deg. C was observed to be greater than the values for the unirradiated material at the same temperatures. Tensile tests on the 9 dpa specimens followed similar trends

  12. Irradiation behaviour of advanced fuel elements for the helium-cooled high temperature reactor (HTR)

    International Nuclear Information System (INIS)

    Nickel, H.

    1990-05-01

    The design of modern HTRs is based on high quality fuel. A research and development programme has demonstrated the satisfactory performance in fuel manufacturing, irradiation testing and accident condition testing of irradiated fuel elements. This report describes the fuel particles with their low-enriched UO 2 kernels and TRISO coating, i.e. a sequence of pyrocarbon, silicon carbide, and pyrocarbon coating layers, as well as the spherical fuel element. Testing was performed in a generic programme satisfying the requirements of both the HTR-MODUL and the HTR 500. With a coating failure fraction less than 2x10 -5 at the 95% confidence level, the results of the irradiation experiments surpassed the design targets. Maximum accident temperatures in small, modular HTRs remain below 1600deg C, even in the case of unrestricted core heatup after depressurization. Here, it was demonstrated that modern TRISO fuels retain all safety-relevant fission products and that the fuel does not suffer irreversible changes. Isothermal heating tests have been extended to 1800deg C to show performance margins. Ramp tests to 2500deg C demonstrate the limits of present fuel materials. A long-term programm is planned to improve the statistical significance of presently available results and to narrow remaining uncertainty limits. (orig.) [de

  13. High Temperature Uniaxial Compression and Stress-Relaxation Behavior of India-Specific RAFM Steel

    Science.gov (United States)

    Shah, Naimish S.; Sunil, Saurav; Sarkar, Apu

    2018-05-01

    India-specific reduced activity ferritic martensitic steel (INRAFM), a modified 9Cr-1Mo grade, has been developed by India as its own structural material for fabrication of the Indian Test Blanket Module (TBM) to be installed in the International Thermonuclear Energy Reactor (ITER). The extensive study on mechanical and physical properties of this material has been currently going on for appraisal of this material before being put to use in the ITER. High temperature compression, stress-relaxation, and strain-rate change behavior of the INRAFM steel have been investigated. The optical microscopic and scanning electron microscopic characterizations were carried out to observe the microstructural changes that occur during uniaxial compressive deformation test. Comparable true plastic stress values at 300 °C and 500 °C and a high drop in true plastic stress at 600 °C were observed during the compression test. Stress-relaxation behaviors were investigated at 500 °C, 550 °C, and 600 °C at a strain rate of 10-3 s-1. The creep properties of the steel at different temperatures were predicted from the stress-relaxation test. The Norton's stress exponent (n) was found to decrease with the increasing temperature. Using Bird-Mukherjee-Dorn relationship, the temperature-compensated normalized strain rate vs stress was plotted. The stress exponent (n) value of 10.05 was obtained from the normalized plot. The increasing nature of the strain rate sensitivity (m) with the test temperature was found from strain-rate change test. The low plastic stability with m 0.06 was observed at 600 °C. The activation volume (V *) values were obtained in the range of 100 to 300 b3. By comparing the experimental values with the literature, the rate-controlling mechanisms at the thermally activated region of high temperature were found to be the nonconservative movement of jogged screw dislocations and thermal breaking of attractive junctions.

  14. Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2016-01-15

    Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

  15. Long-Term Cyclic Oxidation Behavior of Wrought Commercial Alloys at High Temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Li, Bingtao [Iowa State Univ., Ames, IA (United States)

    2003-01-01

    The oxidation resistance of a high-temperature alloy is dependent upon sustaining the formation of a protective scale, which is strongly related to the alloying composition and the oxidation condition. The protective oxide scale only provides a finite period of oxidation resistance owing to its eventual breakdown, which is especially accelerated under thermal cycling conditions. This current study focuses on the long-term cyclic oxidation behavior of a number of commercial wrought alloys. The alloys studied were Fe- and Ni-based, containing different levels of minor elements, such as Si, Al, Mn, and Ti. Oxidation testing was conducted at 1000 and 1100 C in still air under both isothermal and thermal cycling conditions (1-day and 7-days). The specific aspects studied were the oxidation behavior of chromia-forming alloys that are used extensively in industry. The current study analyzed the effects of alloying elements, especially the effect of minor element Si, on cyclic oxidation resistance. The behavior of oxide scale growth, scale spallation, subsurface changes, and chromium interdiffusion in the alloy were analyzed in detail. A novel model was developed in the current study to predict the life-time during cyclic oxidation by simulating oxidation kinetics and chromium interdiffusion in the subsurface of chromia-forming alloys.

  16. High-temperature strength of Nb-1%Zr alloy for irradiation-capsules inner-shell

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Nakata, Hirokatsu; Tanaka, Mitsuo; Fukaya, Kiyoshi.

    1978-04-01

    Coated fuel particles in capsules will be irradiated at about 1600 0 C in JMTR. Nb-1%Zr alloy was chosen for inner shell material of the capsules because of its sufficient strength at 1000 0 C and low induced radioactivity. Nb-1%Zr ingot produced by electron beam melting was formed into seamless tubes by hollowing and swaging, followed by annealing. Creep test in helium flow and tensile test in vacuum were made to examine mechanical strength of the Nb-1%Zr tubes at 1000 0 C. Following are the results; 1) 0.2% yield stress at 1000 0 C is about 6 kg/mm 2 . 2) 3000 hr creep rupture stress at 1000 0 C is about 6 kg/mm 2 . (auth.)

  17. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, Toshikazu; Kanda, Keiji; Mishima, Kaichiro; Tamai, Tadaharu; Hayashi, Masatoshi; Snelgrove, James L.; Stahl, David; Matos, James E.; Travelli, Armando; Case, F. Neil; Posey, John C.

    1983-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel cladding material. The release of fission products from the fuel plate at temperature below 500 deg. C was found negligible. The first rapid release of fission products was observed with the occurrence of blistering at 561±1 deg. C on the plates. The next release at 585. C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 deg. C of U-Al x . The released material was mostly xenon, but small amounts of iodine and cesium were observed. (author)

  18. Release of fission products from irradiated aluminide fuel at high temperature

    International Nuclear Information System (INIS)

    Shibata, T.; Kanda, K.; Mishima, K.

    1982-01-01

    Irradiated uranium aluminide fuel plates of 40% U-235 enrichment were heated for the determination of fission products released under flowing helium gas at temperatures up to and higher than the melting point of fuel-cladding material. The release of fission products from the fuel plate at temperature below 500 0 C was found negligible. The firist rapid release of fission products was observed with the occurrence of blistering at 561 +- 1 0 C on the plates. The next release at 585 0 C might be caused by melting of the cladding material of 6061-Al alloy. The last release of fission product gases was occurred at the eutectic temperature of 640 0 C of U-Al/sub x/. The released material was mostly xenon, but small amounts of iodine and cesium were observed

  19. A new disordering mechanism in A15 type compounds submitted to low temperature irradiation or to quenching from high temperatures

    International Nuclear Information System (INIS)

    Fluekiger, R.

    1984-05-01

    A new diffusion mechanism describing the changes of the long range order parameter in A15 type compounds after both quenching from high temperatures or low temperature irradiation with high energy particles is presented. It is based on the occupation of nonequilibrium or 'virtual' sites centered halfway between two neighbouring A atoms on 6c sites, arising from the instability of a single 6c vacancy recently found by Welch and coworkers by pair potential calculations. After low temperature irradiation, the occupation of this interstitial site creates the necessary conditions for A B site exchanges over several interatomic distances by focused replacement collision sequences. Due to the occupation of a certain concentration of virtual sites, atomic 'overlapping' is not only possible between A atoms on the chains or between A and B atoms (due to deviations from perfect ordering),but also between B atoms on BBB sequences. The latter are retained after low temperature irradiation only and are responsible for the observed lattice expansion and static displacement. (orig.) [de

  20. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  1. Oxidation Behavior of AlN/h-BN Nano Composites at High Temperature

    International Nuclear Information System (INIS)

    Jin Haiyun; Huang Yinmao; Feng Dawei; He Bo; Yang Jianfeng

    2011-01-01

    Both AlN/ nano h-BN composites and AlN/ micro h-BN composites were fabricated. The high temperature oxidation behaviors were investigated at 1000deg. C and 1300deg. C using a cycle-oxidation method. The results showed that there were little changes of both nano composites and monolithic AlN ceramic at temperature of 1000deg. C. And at 1300deg. C, the oxidation dynamics curve of composites could be divided into two courses: a slowly weight increase and a rapid weight decrease, but the oxidation behavior of nano composites was better than micro composites. It was due to that the uniform distribution of oxidation production (Al 18 B 4 O 33 ) surround the AlN grains in nano composites and the oxidation proceeding was retarded. The XRD analysis and SEM observations showed that there was no BN remained in the composites surface after 1300deg. C oxidation and the micropores remain due to the vaporizing of B 2 O 3 oxidized by BN.

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

    Directory of Open Access Journals (Sweden)

    Yonghao Leong

    2016-09-01

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

  3. High-temperature behavior of a deformed Fermi gas obeying interpolating statistics.

    Science.gov (United States)

    Algin, Abdullah; Senay, Mustafa

    2012-04-01

    An outstanding idea originally introduced by Greenberg is to investigate whether there is equivalence between intermediate statistics, which may be different from anyonic statistics, and q-deformed particle algebra. Also, a model to be studied for addressing such an idea could possibly provide us some new consequences about the interactions of particles as well as their internal structures. Motivated mainly by this idea, in this work, we consider a q-deformed Fermi gas model whose statistical properties enable us to effectively study interpolating statistics. Starting with a generalized Fermi-Dirac distribution function, we derive several thermostatistical functions of a gas of these deformed fermions in the thermodynamical limit. We study the high-temperature behavior of the system by analyzing the effects of q deformation on the most important thermostatistical characteristics of the system such as the entropy, specific heat, and equation of state. It is shown that such a deformed fermion model in two and three spatial dimensions exhibits the interpolating statistics in a specific interval of the model deformation parameter 0 < q < 1. In particular, for two and three spatial dimensions, it is found from the behavior of the third virial coefficient of the model that the deformation parameter q interpolates completely between attractive and repulsive systems, including the free boson and fermion cases. From the results obtained in this work, we conclude that such a model could provide much physical insight into some interacting theories of fermions, and could be useful to further study the particle systems with intermediate statistics.

  4. Review on fatigue behavior of high-strength concrete after high temperature

    Science.gov (United States)

    Zhao, Dongfu; Jia, Penghe; Gao, Haijing

    2017-06-01

    The fatigue of high-strength concrete after high temperature has begun to attract attention. But so far the researches work about the fatigue of high-strength concrete after high temperature have not been reported. This article based on a large number of literature. The research work about the fatigue of high-strength concrete after high temperature are reviewed, analysed and expected, which can provide some reference for the experimental study of fatigue damage analysis.

  5. Fission product behavior in high-temperature water: CsI vs MoO4

    Science.gov (United States)

    Kanjana, K.; Silva, K.; Channuie, J.

    2017-09-01

    Fission product behaviors of Cs, a major element released in a severe nuclear accident, still remain unclear. The question frequently addressed is whether Cs released will be in the form of Cs2MoO4 or CsOH. This is a challenging issue since it has been demonstrated that the reaction between Cs2MoO4 and water leading to CsOH production is thermodynamically favored. The present research aims at investigation of CsOH generation through this chemical channel. A high-temperature setup with a flow system based on the cooling system of a water-cooled nuclear reactor has been assembled. The reaction between aqueous solutions of CsI and Na2MoO4 in a high-corrosion-resistant hot cell (Hastelloy) has been studied up to 80 °C in deoxygenated system. The products have been characterized using FTIR and XRD. The results have shown that there is no reaction between CsI and Na2MoO4 under the experimental conditions.

  6. Constitutive modeling of creep behavior in single crystal superalloys: Effects of rafting at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Ya-Nan, E-mail: fanyn12@mails.tsinghua.edu.cn; Shi, Hui-Ji, E-mail: shihj@mail.tsinghua.edu.cn; Qiu, Wen-Hui

    2015-09-17

    Rafting and creep modeling of single crystal superalloys at high temperatures are important for the safety assessment and life prediction in practice. In this research, a new model has been developed to describe the rafting evolution and incorporated into the Cailletaud single crystal plasticity model to simulate the creep behavior. The driving force of rafting is assumed to be the relaxation of the strain energy, and it is calculated with the local stress state, a superposition of the external and misfit stress tensors. In addition, the isotropic coarsening is introduced by the cube root dependence of the microstructure periodicity on creep time based on Ostwal ripening. Then the influence of rafting on creep deformation is taken into account as the Orowan stress in the single crystal plasticity model. The capability of the proposed model is validated with creep experiments of CMSX-4 at 950 °C and 1050 °C. It is able to predict the rafting direction at complex loading conditions and evaluate the channel width during rafting. For [001] tensile creep tests, good agreement has been shown between the model predictions and experimental results at different temperatures and stress levels. The creep acceleration can be captured with this model and is attributed to the microstructure degradation caused by the precipitate coarsening.

  7. Mechanochemical preparation of nanocrystalline TiO2 powders and their behavior at high temperatures

    International Nuclear Information System (INIS)

    Gajovic, A.; Furic, K.; Tomasic, N.; Popovic, S.; Skoko, Z.; Music, S.

    2005-01-01

    Nanocrystalline TiO 2 powders were prepared by high-energy ball-milling using zirconia vial and balls. The changes of microstructure caused by material processing were studied using Raman spectroscopy, X-ray powder diffraction (XRD), transmission electron microscopy (TEM) and selected area electron diffraction (SAED). The milling of the starting TiO 2 powder (anatase + rutile in traces) induced phase transitions to high-pressure polymorph, TiO 2 II, and rutile. We found that the phase transition to TiO 2 II was initiated at the surface of the small particles, while transition to rutile started in their center. Changes in crystallite size during milling process were obtained by the Scherrer method, while the particle size changes were monitored by TEM. The kinetics of phase changes, a decrease in crystallite/particle size, as well as zirconia contamination depended on the powder-to-ball weight ratio. The starting powder and some selected ball-milled samples were investigated in situ by Raman spectroscopy and XRD at high temperatures (up to 1300 deg. C) to examine their behavior during the sintering process. A difference in the results obtained by these two techniques was explained in frame of basic physical properties characterizing both methods. The morphology of the final sinters was monitored by scanning electron microscopy (SEM)

  8. Neutronic behavior of thorium fuel cycles in a very high temperature hybrid system

    International Nuclear Information System (INIS)

    Rodriguez Garcia, Lorena; Milian Perez, Daniel; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel; Velasco, Abanades

    2013-01-01

    Nuclear energy needs to guarantee four important issues to be successful as a sustainable energy source: nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. Pebble bed reactors (PBR), which are very high temperature systems together with fuel cycles based in Thorium, they could offer the opportunity to meet the sustainability demands. Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. This paper shows the main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a variety of fuel cycles with Thorium (Th+U 233 , Th+Pu 239 and Th+U). The parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles using the well-known MCNPX computational code. (author)

  9. Thermomechanical behavior of NiTiPdPt high temperature shape memory alloy springs

    International Nuclear Information System (INIS)

    Nicholson, D E; Vaidyanathan, R; Padula II, S A; Noebe, R D; Benafan, O

    2014-01-01

    Transformation strains in high temperature shape memory alloys (HTSMAs) are generally smaller than for conventional NiTi alloys and can be purposefully limited in cases where stability and repeatability at elevated temperatures are desired. Yet such alloys can still be used in actuator applications that require large strokes when used in the form of springs. Thus there is a need to understand the thermomechanical behavior of shape memory alloy spring actuators, particularly those consisting of alternative alloys. In this work, a modular test setup was assembled with the objective of acquiring stroke, stress, temperature, and moment data in real time during joule heating and forced convective cooling of Ni 19.5 Ti 50.5 Pd 25 Pt 5 HTSMA springs. The spring actuators were subjected to both monotonic axial loading and thermomechanical cycling. The role of rotational constraints (i.e., by restricting rotation or allowing for free rotation at the ends of the springs) on stroke performance was also assessed. Finally, recognizing that evolution in the material microstructure can result in changes in HTSMA spring geometry, the effect of material microstructural evolution on spring performance was examined. This was done by taking into consideration the changes in geometry that occurred during thermomechanical cycling. This work thus provides insight into designing with HTSMA springs and predicting their thermomechanical performance. (paper)

  10. Low cycle fatigue and creep fatigue behavior of alloy 617 at high temperature

    International Nuclear Information System (INIS)

    Cabet, Celine; Carroll, Laura; Wright, Richard

    2013-01-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger (IHX) application of the very high temperature nuclear reactor (VHTR), expected to have an outlet temperature as high as 950 C. Acceptance of Alloy 617 in Section III of the ASME Code for nuclear construction requires a detailed understanding of the creep-fatigue behavior. Initial creep-fatigue work on Alloy 617 suggests a more dominant role of environment with increasing temperature and/or hold times evidenced through changes in creep-fatigue crack growth mechanisms and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle fatigue specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens and the addition of a hold time at peak tensile strain degraded the cycle life. This suggests that creep-fatigue interaction occurs and that the environment may be partially responsible for accelerating failure. (authors)

  11. Neutronic behavior of thorium fuel cycles in a very high temperature hybrid system

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Garcia, Lorena; Milian Perez, Daniel; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

    2013-07-01

    Nuclear energy needs to guarantee four important issues to be successful as a sustainable energy source: nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. Pebble bed reactors (PBR), which are very high temperature systems together with fuel cycles based in Thorium, they could offer the opportunity to meet the sustainability demands. Thorium is a potentially valuable energy source since it is about three to four times as abundant as Uranium. It is also a widely distributed natural resource readily accessible in many countries. This paper shows the main advantages of the use of a hybrid system formed by a Pebble Bed critical nuclear reactor and two Pebble Bed Accelerator Driven Systems (ADSs) using a variety of fuel cycles with Thorium (Th+U{sup 233}, Th+Pu{sup 239} and Th+U). The parameters related to the neutronic behavior like deep burn, nuclear fuel breeding, Minor Actinide stockpile, power density profiles and other are used to compare the fuel cycles using the well-known MCNPX computational code. (author)

  12. SCC crack propagation behavior in 316L weld metal under high temperature water

    International Nuclear Information System (INIS)

    Nakade, Katsuyuki; Hirasaki, Toshifumi; Suzuki, Shunichi; Takamori, Kenro; Kumagai, Katsuhiko; Tanaka, Yoshihiko; Umeoka, Kuniyoshi

    2008-01-01

    Intergranular stress corrosion cracking (SCC) of 316L weld metal is of concern to the BWR plants. PLR pipes in commercial BWR plants have shown SCC in almost HAZ area in high temperature water, whereas, SCC has been arrested around fusion boundary for long time in the actual PLR pipe. The SCC behavior could be characterized in terms of dendrite direction, which was defined as the angle between dendrite growth direction and macro-SCC direction. In this study, the relationship between dendrite growth direction and macro-SCC direction was clearly showed on the fracture surface. The relative large difference of SCC susceptibility of 316L HAZ and weld metal was observed on the fracture surface. In the case of 0 degree, SCC has rapidly propagated into the weld metal parallel to the dendrite structure. In the case of more than 30 degree SCC direction, SCC was arrested around fusion area, and 60 degree SCC was drastically arrested around the fusion area. The large inclined dendrite structure for SCC is highly resistant to SCC. (author)

  13. High pressure study of high temperatures superconductors: Material base, universal Tc-behavior, and charge transfer

    International Nuclear Information System (INIS)

    Chu, C.W.; Hor, P.H.; Lin, J.G.; Xiong, Q.; Huang, Z.J.; Meng, R.L.; Xue, Y.Y.; Jean, Y.C.

    1991-01-01

    The superconducting transition temperature (T c ) has been measured in YBa 2 Cu 3 O 6.7 , YBa 2 Cu 3 O 7 , Y 2 Ba 4 Cu 7 O 15 , YBa 2 Cu 4 O 8 , Tl 2 Ba 2 Ca n-1 Cu n O n+4-δ , La 2-x Sr x CuO 4 , and La 2-x Ba x CuO 4 under high pressures. The pressure effect on the positron lifetime (τ) has also been determined in the first four compounds. Based on these and other high pressure data, the authors suggest that (1) all known cuprate high temperature superconductors (HTS's) may be no more than mere modifications of either 214-T, 214-T', 123, or a combination of 214-T' and 123, (2) a nonmonotonic T c -behavior may govern the T c -variation of all hole cuprate HTS's and (3) pressure can induce charge transfer leading to a T c -change. The implications of these suggestions will also be discussed

  14. Development of crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation (1) (Contract research)

    International Nuclear Information System (INIS)

    Izumo, Hironobu; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ishitsuka, Etsuo; Chimi, Yasuhiro; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; Nakano, Junichi; Kaji, Yoshiyuki; Tsukada, Takashi

    2009-04-01

    To evaluate integrity of irradiation-assisted stress corrosion cracking (IASCC) on in-core structural materials used in light water reactors (LWRs), useful knowledge regarding IASCC has been obtained mainly by post-irradiation examinations (PIEs). In the core of commercial LWRs, however, the actual IASCC occurs under the effects of irradiation on both materials and high-temperature water environment. Therefore, it is necessary to confirm the suitability of the knowledge by PIE with comparison to IASCC behaviors during in-core SCC tests. Fundamental techniques for in-core crack growth and crack initiation tests have been developed already at the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). For the in-core crack growth test technique, to evaluate the effects of neutron irradiation on stainless steels irradiated to low neutron fluences, it is indispensable to develop new loading technique which is applicable to compact tension (CT) specimens with thickness of 0.5 inch (0.5T), from the viewpoint of validity based on the fracture mechanics. Based on the present technical investigation for the in-core loading technique, it is expected that a target load of 7.6 kN approximately can apply to a 0.5T-CT specimen by adopting a loading unit of a lever type instead of the previous uni-axial tension type. For the in-core crack initiation test technique, moreover, construction of a loading unit adopting linear variable differential transformers (LVDTs) has been investigated and technical issues have examined. (author)

  15. Post-irradiation examinations and high-temperature tests on undoped large-grain UO{sub 2} discs

    Energy Technology Data Exchange (ETDEWEB)

    Noirot, J., E-mail: jean.noirot@cea.fr [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Pontillon, Y. [CEA, DEN, DEC, Cadarache, F-13108 St. Paul Lez Durance (France); Yagnik, S. [EPRI, P.O. Box 10412, Palo Alto, CA 94303-0813 (United States); Turnbull, J.A. [Independent Consultant (United Kingdom)

    2015-07-15

    Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants –in the form of thin circular discs – were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/t{sub HM}. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO{sub 2} discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/t{sub HM}. Detailed characterizations of some of these irradiated large-grain UO{sub 2} discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO{sub 2} discs irradiated under the same conditions. Examination of the high burn-up large-grain UO{sub 2} discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO{sub 2} discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/t{sub HM} discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel’s microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms

  16. Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

    International Nuclear Information System (INIS)

    Cabrera-Salcedo, A.

    2012-01-01

    During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

  17. Study of behavior of concrete and cement based composite materials exposed to high temperatures

    OpenAIRE

    Bodnárová, L.; Horák, D.; Válek, J.; Hela, R.; Sitek, L. (Libor)

    2013-01-01

    The paper describes possibilities of observation of behaviour of concrete and cement based composite material exposed to high temperatures. Nowadays, for large-scale tests of behaviour of concrete exposed to high temperatures, testing devices of certified fire testing stations in the Czech Republic and surrounding states are used. These tests are quite expensive. For experimental verification of smaller test specimens, a testing device was built at the Technical University in Brno, wher...

  18. Design, Fabrication, Test Report of the Material Capsule(08M-10K) with Double Thermal Media for High-temperature Irradiation

    International Nuclear Information System (INIS)

    Cho, Man Soon; Choo, K. N.; Kang, Y. H.; Sohn, J. M.; Shin, Y. T.; Park, S. J.; Kim, B. G.; Oh, S. Y.

    2010-01-01

    To overcome the restriction of the irradiation test at a high temperature of the existing material capsule with Al thermal media, a capsule suitable for the irradiation at the high temperature was developed and the performance test was undertaken. The 08M-10K capsule was designed and fabricated as that with double thermal media to verify the structural and external integrity in the high-temperature irradiation higher than 500 .deg. C. The thermal performance test was undertaken at the out-pile test facility, and the soundness of the double thermal media was confirmed with the naked eye after disassembling the capsule. Though the temperatures of the specimens reach 500±20 .deg. C as a result maintaining the capsule during 5 hours after setting the specimens temperatures in the target range, the high-temperature thermal media with double structure was confirmed to maintain the soundness. And the specimens and the thermal media were heated to 600 .deg. C for about 3 minutes, but the thermal media were maintained sound. Especially, the Al thermal media were heated for 5 hours in range of 500±20 .deg. C and for 3 minutes at the temperature of 600 .deg. C. However, the thermal media were confirmed to maintain the soundness. Whether a capsule has only Al thermal media or the high-temperature thermal media with double structure, any capsule can be used in the range of 500±20 .deg. C as the result of this experiment maintaining the specimens high-temperature

  19. Preparation and High-temperature Anti-adhesion Behavior of a Slippery Surface on Stainless Steel.

    Science.gov (United States)

    Zhang, Pengfei; Huawei, Chen; Liu, Guang; Zhang, Liwen; Zhang, Deyuan

    2018-03-29

    Anti-adhesion surfaces with high-temperature resistance have a wide application potential in electrosurgical instruments, engines, and pipelines. A typical anti-wetting superhydrophobic surface easily fails when exposed to a high-temperature liquid. Recently, Nepenthes-inspired slippery surfaces demonstrated a new way to solve the adhesion problem. A lubricant layer on the slippery surface can act as a barrier between the repelled materials and the surface structure. However, the slippery surfaces in previous studies rarely showed high-temperature resistance. Here, we describe a protocol for the preparation of slippery surfaces with high-temperature resistance. A photolithography-assisted method was used to fabricate pillar structures on stainless steel. By functionalizing the surface with saline, a slippery surface was prepared by adding silicone oil. The prepared slippery surface maintained the anti-wetting property for water, even when the surface was heated to 300 °C. Also, the slippery surface exhibited great anti-adhesion effects on soft tissues at high temperatures. This type of slippery surface on stainless steel has applications in medical devices, mechanical equipment, etc.

  20. Analysis of irradiation-induced stresses in coating layers of coated fuel particles for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Kikuchi, Teruo; Fukuda, Kousaku; Sato, Sadao; Toyota, Junji; Shiozawa, Shusaku; Sawa, Kazuhiro; Kashimura, Satoru.

    1991-07-01

    Irradiation-induced stresses in coating layers of coated fuel particles were analyzed by the MICROS-2 code for the fuels of the High Temperature Engineering Test Reactor (HTTR) under its operating conditions. The analyses were made on the standard core fuel (A-type) and the test fuels comprising the advanced SiC-coated particle fuel (B-1 type) and the ZrC-coated particle fuel (B-2 type). For the B-1 type fuel, the stresses were relieved due to the thicker buffer and SiC layers than for the A type fuel. The slightly decreased thickness of the fourth layer for the B-1 type than for the A type fuel had no significant effect on the stresses. As for the B-2 type fuel, almost the same results as for the B-1 type were obtained under an assumption that the ZrC layer as well as the SiC layer undergoes negligible dimension change within the analysis conditions. The obtained results indicated that the B-1 and B-2 type fuels are better than the A type fuel in terms of integrity against the irradiation-induced stresses. Finally, research subjects for development of the analysis code on the fuel behavior are discussed. (author)

  1. Understanding the Irradiation Behavior of Zirconium Carbide

    International Nuclear Information System (INIS)

    Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

    2013-01-01

    -induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

  2. Off-Axis Ratcheting Behavior of Unidirectional Carbon/Epoxy Laminate under Asymmetric Cyclic Loading at High Temperature

    Science.gov (United States)

    2011-11-01

    ply unidirectional carbon/epoxy laminates [0]12 were fabricated from the prepreg tape of P3252-20 (TORAY). They were laid up by hand and cured in...Off-Axis Ratcheting Behavior of Unidirectional Carbon/Epoxy Laminate under Asymmetric Cyclic Loading at High Temperature Takafumi Suzuki 1 and...Development of an engineering model for predicting the off-axis ratcheting behavior of a unidirectional CFRP laminate has been attempted. For this purpose

  3. Effects of pre-irradiation annealing at high temperature on optical absorption and electron paramagnetic resonance of natural pumpellyite mineral

    International Nuclear Information System (INIS)

    Javier-Ccallata, Henry; Filho, Luiz Tomaz; Sartorelli, Maria L.; Watanabe, Shigueo

    2013-01-01

    Highlights: •Natural pumpellyite mineral presents superposition bands around 900 and 1060 nm due Fe 2+ and Fe 3+ . •High temperature annealing influences the EPR and OA spectra. •The behavior of EPR line for 800 and 900 °C can be attributed to forbidden dd transitions due the Fe 3+ . -- Abstract: Natural silicate mineral of pumpellyite, Ca 2 MgAl 2 (SiO 4 )(Si 2 O 7 )(OH) 2 ·(H 2 O), point group A2/m, has been studied concerning high temperature annealing and γ-radiation effects on Optical Absorption (OA) and Electron Paramagnetic Resonance (EPR) properties. Chemical analysis revealed that besides Si, Al, Ca and Mg, other oxides i.e., Fe, Mn, Na, K, Ti and P are present in the structure as impurities. OA measurements of natural and annealed pumpellyite revealed several bands in the visible region due to spin forbidden transitions of Fe 2+ and Fe 3+ . The behaviour of bands around 900 and 1060 nm, with pre-annealing and γ radiation dose, indicating a transition Fe 2+ → e − + Fe 3+ . On the other hand, EPR measurements reveal six lines of Mn 2+ , and satellites due to hyperfine interaction, superimposed on the signal of Fe 3+ around of g = 2. For heat treatment from 800 °C the signal grows significantly and for 900 °C a strong signal of Fe 3+ hides all Mn 2+ lines. The strong growth of this signal indicates that the transitions are due to Fe 3+ dipole–dipole interactions

  4. Effects of pre-irradiation annealing at high temperature on optical absorption and electron paramagnetic resonance of natural pumpellyite mineral

    Energy Technology Data Exchange (ETDEWEB)

    Javier-Ccallata, Henry, E-mail: henrysjc@gmail.com [Escuela de Ingeniería Electrónica y Telecomunicaciones, Universidad Alas Peruanas Filial Arequipa, Urb. D. A. Carrión G-14, J. L. Bustamante y Rivero, Arequipa (Peru); Laboratório de Sistemas Nanoestruturados, Departamento de Física, Universidade Federal de Santa Catarina, Florianópolis, Santa Catarina (Brazil); Filho, Luiz Tomaz [Departamento de Física Nuclear, Instituto de Física, Universidade de São Paulo, Rua do Matão, travessa R, 187, CEP 05508-900 São Paulo, SP (Brazil); Faculdade de Tecnologia e Ciências Exatas, Universidade São Judas Tadeu, Rua Taquari 546, São Paulo, SP (Brazil); Sartorelli, Maria L. [Laboratório de Sistemas Nanoestruturados, Departamento de Física, Universidade Federal de Santa Catarina, Florianópolis, Santa Catarina (Brazil); Watanabe, Shigueo [Departamento de Física Nuclear, Instituto de Física, Universidade de São Paulo, Rua do Matão, travessa R, 187, CEP 05508-900 São Paulo, SP (Brazil)

    2013-09-15

    Highlights: •Natural pumpellyite mineral presents superposition bands around 900 and 1060 nm due Fe{sup 2+}and Fe{sup 3+}. •High temperature annealing influences the EPR and OA spectra. •The behavior of EPR line for 800 and 900 °C can be attributed to forbidden dd transitions due the Fe{sup 3+}. -- Abstract: Natural silicate mineral of pumpellyite, Ca{sub 2}MgAl{sub 2}(SiO{sub 4})(Si{sub 2}O{sub 7})(OH){sub 2}·(H{sub 2}O), point group A2/m, has been studied concerning high temperature annealing and γ-radiation effects on Optical Absorption (OA) and Electron Paramagnetic Resonance (EPR) properties. Chemical analysis revealed that besides Si, Al, Ca and Mg, other oxides i.e., Fe, Mn, Na, K, Ti and P are present in the structure as impurities. OA measurements of natural and annealed pumpellyite revealed several bands in the visible region due to spin forbidden transitions of Fe{sup 2+} and Fe{sup 3+}. The behaviour of bands around 900 and 1060 nm, with pre-annealing and γ radiation dose, indicating a transition Fe{sup 2+} → e{sup −} + Fe{sup 3+}. On the other hand, EPR measurements reveal six lines of Mn{sup 2+}, and satellites due to hyperfine interaction, superimposed on the signal of Fe{sup 3+} around of g = 2. For heat treatment from 800 °C the signal grows significantly and for 900 °C a strong signal of Fe{sup 3+} hides all Mn{sup 2+} lines. The strong growth of this signal indicates that the transitions are due to Fe{sup 3+} dipole–dipole interactions.

  5. Static and Dynamic Friction Behavior of Candidate High Temperature Airframe Seal Materials

    Science.gov (United States)

    Dellacorte, C.; Lukaszewicz, V.; Morris, D. E.; Steinetz, B. M.

    1994-01-01

    The following report describes a series of research tests to evaluate candidate high temperature materials for static to moderately dynamic hypersonic airframe seals. Pin-on-disk reciprocating sliding tests were conducted from 25 to 843 C in air and hydrogen containing inert atmospheres. Friction, both dynamic and static, was monitored and serves as the primary test measurement. In general, soft coatings lead to excessive static friction and temperature affected friction in air environments only.

  6. Corrosion Behavior of Nickel-Plated Alloy 600 in High Temperature Water

    International Nuclear Information System (INIS)

    Kim, Ji Hyun; Hwang, Il Soon

    2008-01-01

    In this paper, electrochemical and microstructural characteristics of nickel-plated Alloy 600 wee investigated in order to identify the performance of electroless Ni-plating on Alloy 600 in high-temperature aqueous condition with the comparison of electrolytic nickel-plating. For high temperature corrosion test of nickel-plated Alloy 600, specimens were exposed for 770 hours to typical PWR primary water condition. During the test, open circuit potentials (OCP's) of all specimens were measured using a reference electrode. Also, resistance to flow accelerated corrosion (FAC) test was examined in order to check the durability of plated layers in high-velocity flow environment at high temperature. After exposures to high flow rate aqueous condition, the integrity of surfaces was confirmed by using both scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS). For the field application, a remote process for electroless nickel-plating was demonstrated using a plate specimen with narrow gap on a laboratory scale. Finally, a practical seal design was suggested for more convenient application

  7. Some effects of gas adsorption on the high temperature volatile release behavior of a terrestrial basalt, tektite and lunar soil

    Science.gov (United States)

    Graham, D. G.; Muenow, D. W.; Gibson, E. K., Jr.

    1979-01-01

    Mass pyrograms obtained from high-temperature, mass psectrometric pyrolysis of a glassy theoleiitic submarine basalt and a tektite, ground in air to less than 64 microns, have shown N2 and SO release patterns very similar to those from the pyrolysis of mature lunar soil fines. The N2 and CO release behavior from the terrestrial samples reproduces the biomodal, high-temperature (approximately 700 and 1050 C) features from the lunar samples. Unground portions of the basalt and tektite show no release of N2 and CO during pyrolysis. Grinding also alters the release behavior and absolute amounts of H2O and CO2. It is suggested that adsorption of atmospheric gases in addition to solar wind implantation of ions may account for the wide range of values in previously reported concentrations of carbon and nitrogen from lunar fines.

  8. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  9. High temperature annealing of minority carrier traps in irradiated MOCVD n(+)p InP solar cell junctions

    Science.gov (United States)

    Messenger, S. R.; Walters, R. J.; Summers, G. P.

    1993-01-01

    Deep level transient spectroscopy was used to monitor thermal annealing of trapping centers in electron irradiated n(+)p InP junctions grown by metalorganic chemical vapor deposition, at temperatures ranging from 500 up to 650K. Special emphasis is given to the behavior of the minority carrier (electron) traps EA (0.24 eV), EC (0.12 eV), and ED (0.31 eV) which have received considerably less attention than the majority carrier (hole) traps H3, H4, and H5, although this work does extend the annealing behavior of the hole traps to higher temperatures than previously reported. It is found that H5 begins to anneal above 500K and is completely removed by 630K. The electron traps begin to anneal above 540K and are reduced to about half intensity by 630K. Although they each have slightly different annealing temperatures, EA, EC, and ED are all removed by 650K. A new hole trap called H3'(0.33 eV) grows as the other traps anneal and is the only trap remaining at 650K. This annealing behavior is much different than that reported for diffused junctions.

  10. IGSCC growth behaviors of Alloy 690 in hydrogenated high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Arioka, K.; Yamada, T.; Miyamoto, T.; Terachi, T. [INSS, (Japan)

    2011-07-01

    The rate of growth of stress corrosion cracking (SCC) was measured for cold worked and thermally treated and solution treated Alloy 690 (UNS N06690, CW TT690, CW ST690) in hydrogenated pressurized water reactor (PWR) primary water under static load condition. Three important patterns were observed: First, Intergranular stress corrosion cracking (IGSCC) was observed on both TT and ST690 even in static load condition if materials were heavily cold worked although the rate of SCC growth was much slower than that of CW mill annealed Alloy 600. Furthermore much rapid SCC growth was recognized in 20% CW TT690 than that of 20% CW ST690. This is quite different result in the literature in high temperature caustic solution. Second, in order to assess the role of creep, rates of creep crack growth were measured in air, argon, and hydrogen gas environments using 20% CW TT690, and 20% CW MA600 in the range of temperatures between 360 and 460 C; intergranular creep cracking (IG creep cracking) was observed on the test materials even in air. Similar slope of 1/T-type temperature dependencies on IGSCC and IG creep crack growth were observed on 20% CW TT690. Similar fracture morphologies and similar 1/T-type temperature dependencies suggest that creep is important in the growth of IGSCC of CW TT690 in high temperature water. Third, cavities and pores were observed at grain boundaries near tips of SCC and creep although the size of the cavities and pores of SCC were much smaller than that of creep cracks. Also the population and size of cavities seem to decrease with decreasing test temperature. These results suggest that the difference in the size and population of cavities might be related with the difference in crack growth rate. And the cavities seem to be formed result from collapse of vacancies at grain boundaries as the crack embryo. This result suggests that diffusion of condensation of vacancies in high stressed fields occurs in high temperature water and gas environments

  11. Nb-Based Nb-Al-Fe Alloys: Solidification Behavior and High-Temperature Phase Equilibria

    Science.gov (United States)

    Stein, Frank; Philips, Noah

    2018-03-01

    High-melting Nb-based alloys hold significant promise for the development of novel high-temperature materials for structural applications. In order to understand the effect of alloying elements Al and Fe, the Nb-rich part of the ternary Nb-Al-Fe system was investigated. A series of Nb-rich ternary alloys were synthesized from high-purity Nb, Al, and Fe metals by arc melting. Solidification paths were identified and the liquidus surface of the Nb corner of the ternary system was established by analysis of the as-melted microstructures and thermal analysis. Complementary analysis of heat-treated samples yielded isothermal sections at 1723 K and 1873 K (1450 °C and 1600 °C).

  12. Effect of Ca and Y additions on oxidation behavior of magnesium alloys at high temperatures

    Institute of Scientific and Technical Information of China (English)

    FAN Jianfeng; YANG Changlin; XU Bingshe

    2012-01-01

    Oxidation and ignition of magnesium alloys at elevated temperature were successfully retarded by additions of Y and Ca.which could be melted at 1173 K in air without any protection.Thermogravimetric measurements in dry air revealed that the oxidation dynamics curves of Mg-2.5Ca alloy and Mg-3.5Y-0.79Ca alloy at high temperatures followed the parabolic-line law or the ubic-line law.X-ray diffraction (XRD) and scanning electron microscopy (SEM) analysis indicated that the oxide film on the surface of Mg-3.5Y-0.79Ca and Mg-2.5Ca alloys exhibited a duplex structure.which agreed with the results of thermodynamic analysis.By comparison,the ignition-proof effect of the combination addition of Y and Ca was better than that of the single addition of Ca.

  13. Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

    1998-01-01

    Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

  14. Modeling of High Temperature Oxidation Behavior of FeCrAl Alloy by using Artificial Neural Network

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Joon; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Refractory alloys are candidate materials for replacing current zirconium-base cladding of light water reactors and they retain significant creep resistance and mechanical strength at high temperatures up to 1500 ℃ due to their high melting temperature. Thermal neutron cross sections of refractory metals are higher than that of zirconium, however the loss of neutron can be overcome by reducing cladding thickness which can be facilitated with enhanced mechanical properties. However, most refractory metals show the poor oxidation resistance at a high temperature. Oxidation behaviors of the various compositions of FeCrAl alloys in high temperature conditions were modeled by using Bayesian neural network. The automatic relevance determination (ARD) technique represented the influence of the composition of alloying elements on the oxidation resistance of FeCrAl alloys. This model can be utilized to understand the tendency of oxidation behavior along the composition of each element and prove the applicability of neural network modeling for the development of new cladding material of light water reactors.

  15. Experimental study of carbon materials behavior under high temperature and VUV radiation: Application to Solar Probe+ heat shield

    International Nuclear Information System (INIS)

    Eck, J.; Sans, J.-L.; Balat-Pichelin, M.

    2011-01-01

    The aim of the Solar Probe Plus (SP+) mission is to understand how the solar corona is heated and how the solar wind is accelerated. To achieve these goals, in situ measurements are necessary and the spacecraft has to approach the Sun as close as 9.5 solar radii. This trajectory induces extreme environmental conditions such as high temperatures and intense Vacuum Ultraviolet radiation (VUV). To protect the measurement and communication instruments, a heat shield constituted of a carbon material is placed on the top of the probe. In this study, the physical and chemical behavior of carbon materials is experimentally investigated under high temperatures (1600-2100 K), high vacuum (10 -4 Pa) and VUV radiation in conditions near those at perihelion for SP+. Thanks to several in situ and ex situ characterizations, it was found that VUV radiation induced modification of outgassing and of mass loss rate together with alteration of microstructure and morphology.

  16. Correlation between Mechanical Behavior and Actuator-type Performance of Ni-Ti-Pd High-temperature Shape Memory Alloys

    Science.gov (United States)

    Bigelow, Glen S.; Padula, Santo A., II; Garg, Anita; Noebe, Ronald D.

    2007-01-01

    High-temperature shape memory alloys in the NiTiPd system are being investigated as lower cost alternatives to NiTiPt alloys for use in compact solid-state actuators for the aerospace, automotive, and power generation industries. A range of ternary NiTiPd alloys containing 15 to 46 at.% Pd has been processed and actuator mimicking tests (thermal cycling under load) were used to measure transformation temperatures, work behavior, and dimensional stability. With increasing Pd content, the work output of the material decreased, while the amount of permanent strain resulting from each load-biased thermal cycle increased. Monotonic isothermal tension testing of the high-temperature austenite and low temperature martensite phases was used to partially explain these behaviors, where a mismatch in yield strength between the austenite and martensite phases was observed at high Pd levels. Moreover, to further understand the source of the permanent strain at lower Pd levels, strain recovery tests were conducted to determine the onset of plastic deformation in the martensite phase. Consequently, the work behavior and dimensional stability during thermal cycling under load of the various NiTiPd alloys is discussed in relation to the deformation behavior of the materials as revealed by the strain recovery and monotonic tension tests.

  17. WESF cesium capsule behavior at high temperature or during thermal cycling

    International Nuclear Information System (INIS)

    Tingey, G.L.; Gray, W.J.; Shippell, R.J.; Katayama, Y.B.

    1985-06-01

    Double-walled stainless steel (SS) capsules prepared for storage of radioactive 137 Cs from defense waste are now being considered for use as sources for commercial irradiation. Cesium was recovered at B-plant from the high-level radioactive waste generated during processing of defense nuclear fuel. It was then purified, converted to the chloride form, and encapsulated at the Hanford Waste Encapsulation and Storage Facility (WESF). The molten cesium chloride salt was encapsulated by pouring it into the inner of two concentric SS cylinders. Each cylinder was fitted with a SS end cap that was welded in place by inert gas-tungsten arc welding. The capsule configuration and dimensions are shown in Figure 1. In a recent review of the safety of these capsules, Tingey, Wheelwright, and Lytle (1984) indicated that experimental studies were continuing to produce long-term corrosion data, to reaffirm capsule integrity during a 90-min fire where capsule temperatures reached 800 0 C, to monitor mechanical properties as a function of time, and to assess the effects of thermal cycling due to periodic transfer of the capsules from a water storage pool to the air environment of an irradiator facility. This report covers results from tests that simulated the effects of the 90-min fire and from thermal cycling actual WESF cesium capsules for 3845 cycles over a period of six months. 11 refs., 39 figs., 9 tabs

  18. Phase Transformation and Creep Behavior in Ti50Pd30Ni20 High Temperature Shape Memory Alloy in Compression

    Science.gov (United States)

    Kumar, Parikshith K.; Desai, Uri; Monroe, James; Lagoudas, Dimitris C.; Karaman, Ibrahim; Noebe, Ron; Bigelow, Glenn

    2010-01-01

    The creep behavior and the phase transformation of Ti50Pd30Ni20 High Temperature Shape Memory Alloy (HTSMA) is investigated by standard creep tests and thermomechanical tests. Ingots of the alloy are induction melted, extruded at high temperature, from which cylindrical specimens are cut and surface polished. A custom high temperature test setup is assembled to conduct the thermomechanical tests. Following preliminary monotonic tests, standard creep tests and thermally induced phase transformation tests are conducted on the specimen. The creep test results suggest that over the operating temperatures and stresses of this alloy, the microstructural mechanisms responsible for creep change. At lower stresses and temperatures, the primary creep mechanism is a mixture of dislocation glide and dislocation creep. As the stress and temperature increase, the mechanism shifts to predominantly dislocation creep. If the operational stress or temperature is raised even further, the mechanism shifts to diffusion creep. The thermally induced phase transformation tests show that actuator performance can be affected by rate independent irrecoverable strain (transformation induced plasticity + retained martensite) as well as creep. The rate of heating and cooling can adversely impact the actuators performance. While the rate independent irrecoverable strain is readily apparent early in the actuators life, viscoplastic strain continues to accumulate over the lifespan of the HTSMA. Thus, in order to get full actuation out of the HTSMA, the heating and cooling rates must be sufficiently high enough to avoid creep.

  19. Calculation of high-temperature insulation parameters and heat transfer behaviors of multilayer insulation by inverse problems method

    Directory of Open Access Journals (Sweden)

    Huang Can

    2014-08-01

    Full Text Available In the present paper, a numerical model combining radiation and conduction for porous materials is developed based on the finite volume method. The model can be used to investigate high-temperature thermal insulations which are widely used in metallic thermal protection systems on reusable launch vehicles and high-temperature fuel cells. The effective thermal conductivities (ECTs which are measured experimentally can hardly be used separately to analyze the heat transfer behaviors of conduction and radiation for high-temperature insulation. By fitting the effective thermal conductivities with experimental data, the equivalent radiation transmittance, absorptivity and reflectivity, as well as a linear function to describe the relationship between temperature and conductivity can be estimated by an inverse problems method. The deviation between the calculated and measured effective thermal conductivities is less than 4%. Using the material parameters so obtained for conduction and radiation, the heat transfer process in multilayer thermal insulation (MTI is calculated and the deviation between the calculated and the measured transient temperatures at a certain depth in the multilayer thermal insulation is less than 6.5%.

  20. Experimental study and modelling of the high temperature mechanical behavior of oxide dispersion strengthened ferritic steels

    International Nuclear Information System (INIS)

    Steckmeyer, A.

    2012-01-01

    The strength of metals, and therefore their maximum operating temperature, can be improved by oxide dispersion strengthening (ODS). Numerous research studies are carried out at the French Atomic Energy Commission (CEA) in order to develop a cladding tube material for Gen IV nuclear power reactors. Oxide dispersion strengthened steels appear to be the most promising candidates for such application, which demands a minimum operating temperature of 650 C. The present dissertation intends to improve the understanding of the mechanical properties of ODS steels, in terms of creep lifetime and mechanical anisotropy. The methodology of this work includes mechanical tests between room temperature and 900 C as well as macroscopic and polycrystalline modelling. These tests are carried out on a Fe-14Cr1W0,26Ti + 0,3 Y 2 O 3 ODS ferritic steel processed at CEA by mechanical alloying and hot extrusion. The as-received material is a bar with a circular section. The mechanical tests reveal the high mechanical strength of this steel at high temperature. A strong influence of the strain rate on the ductility and the mechanical strength is also observed. A macroscopic mechanical model has been developed on the basis of some experimental statements such as the high kinematic contribution to the flow stress. This model has a strong ability to reproduce the mechanical behaviour of the studied material. Two different polycrystalline models have also been developed in order to reproduce the mechanical anisotropy of the material. They are based on its specific grain morphology and crystallographic texture. The discrepancy between the predictions of both models and experimental results reveal the necessity to formulate alternate assumptions on the deformation mechanisms of ODS ferritic steels. (author) [fr

  1. The thermochemical behavior of some binary shape memory alloys by high temperature direct synthesis calorimetry

    International Nuclear Information System (INIS)

    Meschel, S.V.; Pavlu, J.; Nash, P.

    2011-01-01

    Research highlights: → We studied 14 shape memory alloys. → The enthalpies of formation and structure characteristics are summarized. → Theoretical predictions by ab initio calculations compare better with experimental measurements than Miedema's semi empirical model. - Abstract: The standard enthalpies of formation of some shape memory alloys have been measured by high temperature direct synthesis calorimetry at 1373 K. The following results (in kJ/mol of atoms) are reported: CoCr (-0.3 ± 2.9); CuMn (-3.7 ± 3.2); Cu 3 Sn (-10.4 ± 3.1); Fe 2 Tb (-5.5 ± 2.4); Fe 2 Dy (-1.6 ± 2.9); Fe 17 Tb 2 (-2.1 ± 3.1); Fe 17 Dy 2 (-5.3 ± 1.7); FePd 3 (-16.0 ± 2.7); FePt (-23.0 ± 1.9); FePt 3 (-20.7 ± 2.3); NiMn (-24.9 ± 2.6); TiNi (-32.7 ± 1.0); TiPd (-60.3 ± 2.5). The results are compared with some earlier experimental values obtained by calorimetry and by EMF technique. They are also compared with predicted values on the basis of the semi empirical model of Miedema and co-workers and with ab initio calculations when available. We will also assess the available information regarding the structures of these alloys.

  2. Corrosion Behaviors of Structural Materials in High Temperature S-CO{sub 2} Environments

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-15

    The isothermal corrosion tests of several types of stainless steels, Ni-based alloys, and ferritic-martensitic steels (FMS) were carried out at the temperature of 550 and 650 .deg. C in SFR S-CO{sub 2} environment (200 bar) for 1000 h. The weight gain was greater in the order of FMSs, stainless steels, and Ni-based alloys. For the FMSs (Fe-based with low Cr content), a thick outer Fe oxide, a middle (Fe,Cr)-rich oxide, and an inner (Cr,Fe)-rich oxide were formed. They showed significant weight gains at both 550 and 650 .deg. C. In the case of austenitic stainless steels (Fe-based) such as SS 316H and 316LN (18 wt.% Cr), the corrosion resistance was dependent on test temperatures except SS 310S (25 wt.% Cr). After corrosion test at 650 .deg. C, a large increase in weight gain was observed with the formation of outer thick Fe oxide and inner (Cr,Fe)-rich oxide. However, at 550 .deg. C, a thin Cr-rich oxide was mainly developed along with partially distributed small and nodular shaped Fe oxides. Meanwhile, for the Ni-based alloys (16-28 wt.% Cr), a very thin Cr-rich oxide was developed at both test temperatures. The superior corrosion resistance of high Cr or Ni-based alloys in the high temperature S-CO{sub 2} environment was attributed to the formation of thin Cr-rich oxide on the surface of the materials.

  3. Electrochemical polarization behavior of sensitized SUS 304 stainless steel in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Kushiya, K [Tohoku Univ., Sendai (Japan); Sugimoto, K; Ejima, T

    1978-11-01

    Anodic polarization curves for a solution-treated or sensitized SUS 304 stainless steel and solution-treated Fe-Ni-Cr ternary alloys containing 10%Ni and 6 to 14%Cr have been measured in deaerated 0.5 mol/l Na/sub 2/SO/sub 4/ solutions of pH 2.0 to 5.9 at 298, 523 and 553 K. Corrosion potentials for U-bend SCC test specimens of sensitized SUS 304 stainless steel have also been monitored for a long time in the same solutions as those used for the polarization measurements except that they were aerated. It was found that the differences in the current densities in the passive state, i sub(pass), between the solution treated steel and the sensitized one and also between the ternary alloy with higher Cr content and the one with lower Cr content become large with increasing temperature and decreasing pH. This means that the difference in the values of i sub(pass) between grain bodies and Cr-depleted zones along grain boundaries of sensitized steel becomes larger and susceptibility to intergranular corrosion of the sensitized steel in the passivation region becomes higher with increasing temperature and decreasing pH. Since corrosion potentials for the U-bend SCC test specimens in air-satulated solutions lie in the passive region of anodic polarization curves for the sensitized steel in deaerated solutions, the intergranular stress-corrosion cracking of the sensitized steel in high temperature water with dissolved oxygen is considered to be caused by the preferential corrosion in the Cr-depleted zone.

  4. Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-11-15

    In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

  5. Long-term creep behavior of high-temperature gas turbine materials under constant and variable stress

    International Nuclear Information System (INIS)

    Granacher, J.; Preussler, T.

    1987-01-01

    Within the framework of the documented research project, extensive creep rupture tests were carried out with characteristic, high-temperature gas turbine materials for establishment of improved design data. In the range of the main application temperatures and in stress ranges down to application-relevant values the tests extended over a period of about 40,000 hours. In addition, long-term annealing tests were carried out in the most important temperature ranges for the measurement of the density-dependent straim, which almost always manifested itself as a material contraction. Furthermore, hot tensile tests were carried out for the description of the elastoplastic short-term behavior. Several creep curves were derived from the results of the different tests with a differentiated evaluation method. On the basis of these creep curves, creep equations were set up for a series of materials which are valid in the entire examined temperature range and stress range and up to the end of the secondary creep range. Also, equations for the time-temperature-dependent description of the material contraction behavior were derived. With these equations, the high-temperature deformation behavior of the examined materials under constant creep stress can be described simply and application-oriented. (orig.) With 109 figs., 19 tabs., 77 refs [de

  6. Microwave heating behavior and microwave absorption properties of barium titanate at high temperatures

    Directory of Open Access Journals (Sweden)

    K. Kashimura

    2016-06-01

    Full Text Available The temperature dependence of the microwave absorption behavior of BaTiO3 particles was investigated over various frequencies and temperatures of 25-1000 ∘C. First, using both the coaxial transmission line method and the cavity perturbation method by a network analyzer, the real and imaginary parts of the relative permittivity of BaTiO3 ( ε r ′ and ε r ″ , respectively were measured, in order to improve the reliability of the data obtained at 2.45 GHz. The imaginary parts of the relative permittivity as measured by the two methods were explored by their heating behaviors. Furthermore, the temperature dependence of the microwave absorption behavior of BaTiO3 particles was investigated for frequencies of 2.0-13.5 GHz and temperatures of 25-1000 ∘C using the coaxial transmission line method.

  7. Effects of electron beam irradiation on mechanical properties at low and high temperature of fiber reinforced composites using PEEK as matrix material

    International Nuclear Information System (INIS)

    Sasuga, Tsuneo; Seguchi, Tadao; Sakai, Hideo; Odajima, Toshikazu; Nakakura, Toshiyuki; Masutani, Masahiro.

    1987-11-01

    Carbon fiber reinforced composite (PEEK-CF) using polyarylether-ether-ketone (PEEK) as a matrix material was prepared and the electron beam radiation effects on the mechanical properties at low and high temperature and the effects of annealing after irradiation were studied. Cooling down to 77 K, the flexural strength of PEEK-CF increased to about 20 % than that at room temperature. The data of flexural strength for the irradiated specimens showed some scattering, but the strength and modulus at 77 K were changed scarcely up to 120 MGy. The flexural strength and modulus in the unirradiated specimen decreased with increasing of measurement temperature, and the strength at 140 deg C, which is the just below temperature of the glass transition of PEEK, was to 70 % of the value at room temperature. For the irradiated specimens, the strength and modulus increased with dose and the values at 140 deg C for the specimen irradiated with 120 MGy were nearly the same with the unirradiated specimen measured at room temperature. The improvement of mechanical properties at high temperature by irradiation was supported by a viscoelastic measurement in which the glass transition shifted to the higher temperature by the radiation-induced crosslinking. A glass fiber reinforced PEEK composite (PEEK-GF) was prepared and its irradiation effects by electron beam was studied. Unirradiated PEEK-GF showed the same performance with that for GFRP of epoxide resin as matrix material, but by irradiation the flexual strength and modulus decreased with dose. It was revealed that this composite was destroyed by delamination because inter laminar shear strength (ILSS) decreased with dose and analysis of the profile of S-S curve showed typical delamination. Fractoglaphy by electron microscopy supported the delamination which is caused by the lowering of adhesion on interface between the fiber and matrix with increase of dose. (author)

  8. Failure Mechanical Behavior of Australian Strathbogie Granite at High Temperatures: Insights from Particle Flow Modeling

    Directory of Open Access Journals (Sweden)

    Sheng-Qi Yang

    2017-05-01

    Full Text Available Thermally induced damage has an important influence on rock mechanics and engineering, especially for high-level radioactive waste repositories, geological carbon storage, underground coal gasification, and hydrothermal systems. Additionally, the wide application of geothermal heat requires knowledge of the geothermal conditions of reservoir rocks at elevated temperature. However, few methods to date have been reported for investigating the micro-mechanics of specimens at elevated temperatures. Therefore, this paper uses a cluster model in particle flow code in two dimensions (PFC2D to simulate the uniaxial compressive testing of Australian Strathbogie granite at various elevated temperatures. The peak strength and ultimate failure mode of the granite specimens at different elevated temperatures obtained by the numerical methods are consistent with those obtained by experimentation. Since the tensile force is always concentrated around the boundary of the crystal, cracks easily occur at the intergranular contacts, especially between the b-b and b-k boundaries where less intragranular contact is observed. The intergranular and intragranular cracking of the specimens is almost constant with increasing temperature at low temperature, and then it rapidly and linearly increases. However, the inflection point of intergranular micro-cracking is less than that of intragranular cracking. Intergranular cracking is more easily induced by a high temperature than intragranular cracking. At an elevated temperature, the cumulative micro-crack counts curve propagates in a stable way during the active period, and it has no unstable crack propagation stage. The micro-cracks and parallel bond forces in the specimens with elevated temperature evolution and axial strain have different characteristics than those at lower temperature. More branch fractures and isolated wider micro-cracks are generated with increasing temperature when the temperature is over 400

  9. Study of a chromia-forming alloy behavior as interconnect material for High Temperature Vapor Electrolysis

    International Nuclear Information System (INIS)

    Guillou, S.

    2011-01-01

    In High Temperature Vapor Electrolysis (HTVE) system, the materials chosen for the inter-connectors should have a good corrosion behaviour in air and in H 2 -H 2 O mixtures at 800 C, and keep a high electronic conductivity over long durations as well. In this context, the first goal of this study was to evaluate a commercial ferritic alloy (the K41X alloy) as interconnect for HTVE application. Oxidation tests in furnace and in microbalance have therefore been carried out in order to determine oxidation kinetics. Meanwhile, the Area Specific Resistance (ASR) was evaluated by Contact Resistance measurements performed at 800 C. The second objective was to improve our comprehension of chromia-forming alloys oxidation mechanism, in particular in H 2 /H 2 O mixtures. For that purpose, some specific tests have been conducted: tracer experiments, coupled with the characterization of the oxide scale by PEC (Photo-Electro-Chemistry). This approach has also been applied to the study of a LaCrO 3 perovskite oxide coating on the K41X alloy. This phase is indeed of high interest for HTVE applications due to its high conductivity properties. This latter study leads to further understanding on the role of lanthanum as reactive element, which effect is still under discussion in literature.In both media at 800 C, the scale is composed of a Cr 2 O 3 /(Mn,Cr) 3 O 4 duplex scale, covered in the case of H 2 -H 2 O mixture by a thin scale made of Mn 2 TiO 4 spinel. In air, the growth mechanism is found to be cationic, in agreement with literature. The LaCrO 3 coating does not modify the direction of scale growth but lowers the growth kinetics during the first hundreds hours. Moreover, with the coating, the scale adherence is favored and the conductivity appears to be slightly higher. In the H 2 -H 2 O mixture, the growth mechanism is found to be anionic. The LaCrO 3 coating diminishes the oxidation kinetics. Although the scale thickness is about the same in both media, the ASR parameter

  10. High-temperature extrusion behavior of a superplastic zirconia-based ceramic

    International Nuclear Information System (INIS)

    Kellett, B.J.; Carry, C.; Mocellin, A.

    1990-01-01

    Workability of 3-mol%-yttria-stabilized tetragonal ZrO 2 has been gauged through a series of extrusion experiments performed under vacuum with graphite dies at 1500 degrees C and 35 MPa piston stress. It is shown that dense and smooth extrustions can be obtained from solid billets when graphite paper is used as a lubricant. Sigmoidal dies and conical dies with cone angles of 18.4 degrees, 26.6 degrees, and 45 degrees and diameter ratios of 1.5, 2, and 3 were used to explore extrusion behavior. Observed piston velocities correspond to what may be predicted from the experimental uniaxial constitutive creep equation and a simple slab analysis. A precise analysis, however, is not attempted because of lack of steady-state behavior of the material itself

  11. Characterization and corrosion behavior of F6NM stainless steel treated in high temperature water

    Science.gov (United States)

    Li, Zheng-yang; Cai, Zhen-bing; Yang, Wen-jin; Shen, Xiao-yao; Xue, Guo-hong; Zhu, Min-hao

    2018-03-01

    F6NM martensitic stainless steel was exposed to 350 °C water condition for 500, 1500, and 2500 h to simulate pressurized water reactor (PWR) condition. The characterization and corrosion behavior of the oxide film were investigated. Results indicate that the exposed steel surface formed a double-layer oxide film. The outer oxide film is Fe-rich and contains two type oxide particles. However, the inner oxide film is Cr-rich, and two oxide films, whose thicknesses increase with increasing exposure time. The oxide film reduces the corrosion behavior because the outer oxide film has many crack and pores. Finally, the mechanism and factors affecting the formation of the oxide film were investigated.

  12. A review of creep behavior of high temperature composites in relation to molybdenum disilicide composites

    International Nuclear Information System (INIS)

    Sadananda, K.; Feng, C.R.

    1993-01-01

    A brief review of creep behavior of composites is presented. It is shown that even for a two component system, creep of a composite depends on complex combination of several factors, including the constitutive behavior of the component phases at stress and temperature, and mechanical, chemical, diffusional and thermodynamic stability of the two-phase interfaces. The existing theoretical models based on continuum mechanics are presented. These models are evaluated using the extensive experimental data on molydisilicide--silicon carbide composites by the authors. The analysis shows that the rule of mixture based on isostrain and isostress provides two limiting bounds wherein all other predictions fall. For molydisilicide, the creep is predominantly governed by the creep of the majority phase, i.e. the matrix while fibers deform predominately elastically

  13. Effect of simultaneous ion irradiation on microstructural change of SiC/SiC composites at high temperature

    International Nuclear Information System (INIS)

    Taguchi, T.; Wakai, E.; Igawa, N.; Nogami, S.; Snead, L.L.; Hasegawa, A.; Jitsukawa, S.

    2002-01-01

    The effect of simultaneous triple ion irradiation of He, H and Si on microstructural evolution of two kinds of SiC/SiC composites (HNS composite (using Hi-Nicalon type S SiC fiber) and TSA composite (using Tyranno SA SiC fiber)) at 1000 deg. C has been investigated. The microstructure observations of SiC/SiC composites irradiated to 10 dpa were examined by transmission electron microscopy. He bubbles were hardly formed in matrix of TSA composite, but many helium bubbles and some cracks were observed at grain boundaries of matrix of HNS composite. He bubbles and cracks were not, on the other hand, observed in the both fiber fabrics of HNS and TSA composites. Debonding between fiber and carbon layer following irradiation region was not observed in the both composites. Under these irradiation conditions, TSA composite showed the better microstructural stability against ion beams irradiation than one of HNS composite

  14. Stability of Y-Ti-O nanoparticles in ODS alloys during heat treatment and high temperature swift heavy ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Skuratov, V.A. [FLNR, JINR, Dubna (Russian Federation); National Research Nuclear University MEPhI, Moscow (Russian Federation); Dubna State University, Dubna (Russian Federation); Sohatsky, A.S.; Kornieieva, K. [FLNR, JINR, Dubna (Russian Federation); O' Connell, J.H.; Neethling, J.H. [CHRTEM, NMMU, Port Elizabeth (South Africa); Nikitina, A.A.; Ageev, V.S. [JSC VNIINM, Moscow (Russian Federation); Zdorovets, M. [Institute of Nuclear Physics, Astana (Kazakhstan); Ural Federal University, Yekaterinburg (Russian Federation); Volkov, A.D. [Nazarbayev University, Astana (Kazakhstan)

    2016-12-15

    Aim of this report is to compare the morphology of swift (167 and 220 MeV) Xe ion induced latent tracks in Y{sub 2}Ti{sub 2}O{sub 7} nanoparticles during post-irradiation heat treatment and after irradiation at different temperatures in pre-thinned TEM foils and TEM targets prepared from hundreds microns thick irradiated oxide dispersion strengthened (ODS) steel. No difference in track parameters was found in room temperature irradiated nanoparticles in pre-thinned and conventional samples. Microstructural data gathered from pre-thinned foils irradiated in the temperature range 350-650 C or annealed at similar temperatures demonstrate that amorphous latent tracks interact with the surrounding matrix, changing the track and nanoparticle morphology, while such effect is not observed in conventional ODS material treated at the same conditions. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. Friction behavior of cobalt base and nickel base hardfacing materials in high temperature sodium

    International Nuclear Information System (INIS)

    Mizobuchi, Syotaro; Kano, Shigeki; Nakayama, Kohichi; Atsumo, Hideo

    1980-01-01

    A friction behavior of the hardfacing materials such as cobalt base alloy ''Stellite'' and nickel base alloy ''Colmonoy'' used in the sliding components of a sodium cooled fast breeder reactor was investigated in various sodium environments. Also, friction tests on these materials were carried out in argon environment. And they were compared with those in sodium environment. The results obtained are as follows: (1) In argon, the cobalt base hardfacing alloy showed better friction behavior than the nickel base hardfacing alloy. In sodium, the latter was observed to have the better friction behavior being independent of the sodium temperature. (2) The friction coefficient of each material tends to become lower by pre-exposure in sodium. Particularly, this tendency was remarkable for the nickel base hardfacing alloy. (3) The friction coefficient between SUS 316 and one of these hardfacing materials was higher than that between latter materials. Also, some elements of hardfacing alloys were recognized to transfer on the friction surface of SUS 316 material. (4) It was observed that each tested material has a greater friction coefficient with a decrease of the oxygen content in sodium. (author)

  16. X-ray imaging for studying behavior of liquids at high pressures and high temperatures using Paris-Edinburgh press

    International Nuclear Information System (INIS)

    Kono, Yoshio; Kenney-Benson, Curtis; Park, Changyong; Shen, Guoyin; Shibazaki, Yuki; Wang, Yanbin

    2015-01-01

    Several X-ray techniques for studying structure, elastic properties, viscosity, and immiscibility of liquids at high pressures have been integrated using a Paris-Edinburgh press at the 16-BM-B beamline of the Advanced Photon Source. Here, we report the development of X-ray imaging techniques suitable for studying behavior of liquids at high pressures and high temperatures. White X-ray radiography allows for imaging phase separation and immiscibility of melts at high pressures, identified not only by density contrast but also by phase contrast imaging in particular for low density contrast liquids such as silicate and carbonate melts. In addition, ultrafast X-ray imaging, at frame rates up to ∼10 5 frames/second (fps) in air and up to ∼10 4 fps in Paris-Edinburgh press, enables us to investigate dynamics of liquids at high pressures. Very low viscosities of melts similar to that of water can be reliably measured. These high-pressure X-ray imaging techniques provide useful tools for understanding behavior of liquids or melts at high pressures and high temperatures

  17. X-ray imaging for studying behavior of liquids at high pressures and high temperatures using Paris-Edinburgh press

    Energy Technology Data Exchange (ETDEWEB)

    Kono, Yoshio; Kenney-Benson, Curtis; Park, Changyong; Shen, Guoyin [HPCAT, Geophysical Laboratory, Carnegie Institution of Washington, 9700 S. Cass Ave., Argonne, Illinois 60439 (United States); Shibazaki, Yuki [Frontier Research Institute for Interdisciplinary Sciences, Tohoku University, Aramaki aza Aoba 6-3, Aoba-ku, Sendai 980-8578 (Japan); Wang, Yanbin [GeoSoilEnviroCARS, Center for Advanced Radiation Sources, The University of Chicago, 5640 S. Ellis Avenue, Chicago, Illinois 60637 (United States)

    2015-07-15

    Several X-ray techniques for studying structure, elastic properties, viscosity, and immiscibility of liquids at high pressures have been integrated using a Paris-Edinburgh press at the 16-BM-B beamline of the Advanced Photon Source. Here, we report the development of X-ray imaging techniques suitable for studying behavior of liquids at high pressures and high temperatures. White X-ray radiography allows for imaging phase separation and immiscibility of melts at high pressures, identified not only by density contrast but also by phase contrast imaging in particular for low density contrast liquids such as silicate and carbonate melts. In addition, ultrafast X-ray imaging, at frame rates up to ∼10{sup 5} frames/second (fps) in air and up to ∼10{sup 4} fps in Paris-Edinburgh press, enables us to investigate dynamics of liquids at high pressures. Very low viscosities of melts similar to that of water can be reliably measured. These high-pressure X-ray imaging techniques provide useful tools for understanding behavior of liquids or melts at high pressures and high temperatures.

  18. Single-source-precursor Synthesis and High-temperature Behavior of SiC Ceramics Containing Boron

    Science.gov (United States)

    Gui, Miaomiao; Fang, Yunhui; Yu, Zhaoju

    2014-12-01

    In this paper, a hyperbranched polyborocarbosilane (HPBCS) was prepared by a one-pot synthesis with Cl2Si(CH3)CH2Cl, Cl3SiCH2Cl and BCl3 as the starting materials. The obtained HPBCS was characterized by GPC, FT-IR and NMR, and was confirmed to have hyperbranched structures. The thermal property of the resulting HPBCS was investigated by TGA. The ceramic yield of the HPBCS is about 84% and that of the counterpart hyperbranched hydridopolycarbosilane is only 45%, indicating that the introduction of boron into the preceramic polymer significantly improved the ceramic yield. With the polymer-derived ceramic route, the final ceramics were annealed at 1800 °C in argon atmosphere for 2 h in order to characterize the microstructure and to evaluate the high-temperature behavior. The final ceramic microstructure was studied by XRD and SEM, indicating that the introduction of boron dramatically inhibits SiC crystallization. The boron-containing SiC ceramic shows excellent high-temperature behavior against decomposition and crystallization at 1800 °C.

  19. High-temperature oxidation behavior of Ti3AlC2 in air

    Institute of Scientific and Technical Information of China (English)

    XU Xue-wen; LI Yang-xian; ZHU Jiao-qun; MEI Bing-chu

    2006-01-01

    Not only the isothermal oxidation behaviors at 900-1 300 ℃ for 20 h in air of bulk Ti3AlC2 with 2.8% TiC which was sintered by hot pressing with the additive of silicon,but also the cyclic oxidation behavior at 1 100-1 300 °C for 30 cycles,were investigated by using TG,XRD,SEM. The isothermal and cyclic oxidation behaviors generally follow a parabolic rate law. The parabolic rate constants of the former increased from 1.39×10-10 kg2/(m4·s) at 900 ℃ to 5.56×10-9 kg2/(m4·s) at 1 300 ℃. The calculated activation energy is 136.45 kJ/mol. The oxidation products are á-Al2O3 and little TiO2 at 900-1 000 ℃,however when the temperature is raised up to 1 200 ℃,TiO2 partially reacts to Al2TiO5,and the reaction is completed at 1 300 ℃. This demonstrates that Ti3AlC2 has excellent oxidation resistance and good thermal shock because the dense continuous oxide scale consists of mass á-Al2O3 and little TiO2 and/or Al2TiO5. Generally,the oxide scale is grown by the inward diffusion of O2- and the outward diffusion of Ti4+ and Al3+.

  20. Constitutive equations for describing high-temperature inelastic behavior of structural alloys

    International Nuclear Information System (INIS)

    Robinson, D.N.; Pugh, C.E.; Corum, J.M.

    1976-01-01

    This paper addresses constitutive equations for the description of inelastic behavior of LMFBR structural alloys at elevated temperatures. Both elastic-plastic (time-independent) and creep (time-dependent) deformations are considered for types 304 and 316 stainless steel and 2 1 / 4 Cr--1 Mo steel. The constitutive equations identified for interim use in design analyses are described along with the assumptions and data on which they are based. Areas where improvements are needed are identified, and some alternate theories that are being pursued are outlined

  1. High temperature oxidation behavior of AISI 304L stainless steel—Effect of surface working operations

    International Nuclear Information System (INIS)

    Ghosh, Swati; Kumar, M. Kiran; Kain, Vivekanand

    2013-01-01

    Highlights: ► Surface working resulted in thinner oxide on the surface. ► Oxides on machined/ground surfaces richer in Cr, higher in specific resistivity. ► Additional ionic transport process at the metal-oxide for ground sample established. ► Presence of fragmented grains and martensite influenced oxide nature/morphology. - Abstract: The oxidation behavior of grade 304L stainless steel (SS) subjected to different surface finishing (machining and grinding) operations was followed in situ by contact electric resistance (CER) and electrochemical impedance spectroscopy (EIS) measurements using controlled distance electrochemistry (CDE) technique in high purity water (conductivity −1 ) at 300 °C and 10 MPa in an autoclave connected to a recirculation loop system. The results highlight the distinct differences in the oxidation behavior of surface worked material as compared to solution annealed material in terms of specific resistivity and low frequency Warburg impedance. The resultant oxide layer was characterized for (a) elemental analyses by glow discharge optical emission spectroscopy (GDOES) and (b) morphology by scanning electron microscopy (SEM). Oxide layers with higher specific resistivity and chromium content were formed in case of machined and ground conditions. Presence of an additional ionic transport process has also been identified for the ground condition at the metal/oxide interface. These differences in electrochemical properties and distinct morphological features of the oxide layer as a result of surface working were attributed to the prevalence of heavily fragmented grain structure and presence of martensite.

  2. Constitutive Modeling of the High-Temperature Flow Behavior of α-Ti Alloy Tube

    Science.gov (United States)

    Lin, Yanli; Zhang, Kun; He, Zhubin; Fan, Xiaobo; Yan, Yongda; Yuan, Shijian

    2018-05-01

    In the hot metal gas forming process, the deformation conditions, such as temperature, strain rate and deformation degree, are often prominently changed. The understanding of the flow behavior of α-Ti seamless tubes over a relatively wide range of temperatures and strain rates is important. In this study, the stress-strain curves in the temperature range of 973-1123 K and the initial strain rate range of 0.0004-0.4 s-1 were measured by isothermal tensile tests to conduct a constitutive analysis and a deformation behavior analysis. The results show that the flow stress decreases with the decrease in the strain rate and the increase of the deformation temperature. The Fields-Backofen model and Fields-Backofen-Zhang model were used to describe the stress-strain curves. The Fields-Backofen-Zhang model shows better predictability on the flow stress than the Fields-Backofen model, but there exists a large deviation in the deformation condition of 0.4 s-1. A modified Fields-Backofen-Zhang model is proposed, in which a strain rate term is introduced. This modified Fields-Backofen-Zhang model gives a more accurate description of the flow stress variation under hot forming conditions with a higher strain rate up to 0.4 s-1. Accordingly, it is reasonable to adopt the modified Fields-Backofen-Zhang model for the hot forming process which is likely to reach a higher strain rate, such as 0.4 s-1.

  3. The Degradation Behavior of SiCf/SiO2 Composites in High-Temperature Environment

    Science.gov (United States)

    Yang, Xiang; Cao, Feng; Qing, Wang; Peng, Zhi-hang; Wang, Yi

    2018-04-01

    SiCf/SiO2 composites had been fabricated efficiently by Sol-Gel method. The oxidation behavior, thermal shock property and ablation behavior of SiCf/SiO2 composites was investigated. SiCf/SiO2 composites showed higher oxidation resistance in oxidation atmosphere, the flexural strength retention ratio was larger than 90.00%. After 1300 °C thermal shock, the mass retention ratio was 97.00%, and the flexural strength retention ratio was 92.60%, while after 1500 °C thermal shock, the mass retention ratio was 95.37%, and the flexural strength retention ratio was 83.34%. After 15 s ablation, the mass loss rate was 0.049 g/s and recession loss rate was 0.067 mm/s. The SiO2 matrix was melted in priority and becomes loosen and porous. With the ablation going on, the oxides were washed away by the shearing action of the oxyacetylene flame. The evaporation of SiO2 took away large amount of heat, which is also beneficial to the protection for SiCf/SiO2 composites.

  4. Effects of zinc injection on electrochemical corrosion and cracking behavior of stainless steels in borated and lithiated high temperature water

    International Nuclear Information System (INIS)

    Wu Xinqiang; Liu Xiahe; Han Enhou; Ke Wei

    2014-01-01

    Zinc (Zn) injection water chemistry (ZWC) adopted in primary coolant system in pressurized water reactors (PWRs) is to reduce the radiation buildup as well as retard the corrosion degradation in high temperature pressurized water through improving the characteristics of oxide scales formed on components materials. However, Zn injection involved corrosion and cracking behavior and related mechanisms are still under discussion. The understanding of Zn-bearing oxide scale characteristics and their protective property is of significance to clarify the environmentally assisted material failure problems in PWRs power plants. In the present work, in-situ potentiodynamic polarization curves and electrochemical impedance spectra measurements in high temperature borated and lithiated water as well as ex-situ X-ray photoelectron spectroscopy analyses have been done to investigate the effects of temperature (R.T.-603 K), pH T value at 573 K (6.9-7.4) and Zn-injection concentration (0-150 ppb) on electrochemical corrosion behavior and oxide scale characteristics of nuclear-grade stainless steels. The protective property of oxide scales under Zn-free and Zn-injected conditions degraded with increasing temperature, with Cr-rich oxide layer playing a key role on retarding further corrosion. The composition of oxide scales appeared slightly pH T dependent: rich in chromites and ferrites at pH T =6.9 and pH T =7.4, respectively. The corrosion rate decreased significantly in the high pH T value solution with Zn injection due to the formation of thin and compact oxide scales. The ≤50 ppb Zn injection could significantly affect the formation of Zn-bearing oxides on the surfaces, while >50 ppb Zn injection showed no obvious influence on the oxide scales. A modified point defect model was proposed to discuss the effects of injected Zn concentrations on the oxide scales in high temperature water. A 10 ppb Zn injection obviously decreased the intergranular cracking susceptibility of

  5. Polyurethane-Based Ionogels Exhibiting Durable Thermoresponsive Optical Behavior Under High-Temperature Conditions.

    Science.gov (United States)

    Sato, Tomoya; England, Matt W; Wang, Liming; Urata, Chihiro; Kakiuchida, Hiroshi; Hozumi, Atsushi

    2018-01-01

    Polyurethane (PU)-based transparent and flexible ionogels, showing unusual thermo-responsive optical properties, were successfully prepared by mixing PU-precursor and a hydrophobic ionic liquid, 1-ethyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide (EMIM-TFSI). Although the initial ionogels were transparent at room temperature, significant increases in opacity were observed with increasing temperature up to 120°C, because of macroscopic phase separation of the PU-matrix and hydrophobic EMIM-TFSI. In addition, the optical transition temperature could be arbitrarily controlled simply by varying the mixing ratio of EMIM-TFSI within the PU-matrix. As confirmed by UV-Vis spectra acquired at different temperatures, this thermo-responsive optical behavior was found to be reversible, repeatable and durable even after 30 cycles of a thermal-stress testing between 30 and 100°C.

  6. Oxidation behavior of 304 stainless steel exposed to steam at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H.; Ryu, J. R.; Park, G. H. [Kyunghee Univ., Yongin (Korea, Republic of); Yoo, T. G. [FNC Technology, Seoul (Korea, Republic of)

    2003-10-01

    An experiment was conducted on 304 stainless steel(SUS304L) at the LOCA(Lost of Coolant Accident) requirement temperature, 800 .deg. C to 1100 deg. C. SUS304L was used as clothing material and structural frame of LWR. Oxidation behavior of SUS304L by temperature and time was examined after the mechanical and chemical polishing of SUS304L plate. After oxidation, change in weight showed a linear pattern for the first 20 minutes and a parabolic pattern afterwards. Then, fine structure and oxidation layer of SUS304L plate were observed through OM photographing and oxidation characteristics of SUS304L were found through hardness measurement by depth of each plate and XRD(X-Ray Diffraction) photographing.

  7. Theoretical study of thermopower behavior of LaFeO3 compound in high temperature region

    Science.gov (United States)

    Singh, Saurabh; Shastri, Shivprasad S.; Pandey, Sudhir K.

    2018-04-01

    The electronic structure and thermopower (α) behavior of LaFeO3 compound were investigated by combining the ab-initio electronic structures and Boltzmann transport calculations. LSDA plus Hubbard U (U = 5 eV) calculation on G-type anti-ferromagnetic (AFM) configuration gives an energy gap of ˜2 eV, which is very close to the experimentally reported energy gap. The calculated values of effective mass of holes (mh*) in valance band (VB) are found ˜4 times that of the effective mass of electrons (me*) in conduction band (CB). The large effective masses of holes are responsible for the large and positive thermopower exhibited by this compound. The calculated values of α using BoltzTraP code are found to be large and positive in the 300-1200 K temperature range, which is in agreement with the experimentally reported data.

  8. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  9. High-Temperature Oxidation Behavior and Kinetics of Forged 12Cr-MoVW Steel

    Directory of Open Access Journals (Sweden)

    Kim Yong Hwan

    2017-06-01

    Full Text Available The oxidation kinetics of forged 12Cr-MoVW steel was investigated in an air (N2+O2 atmosphere at 873-1073 K (Δ50 K using thermogravimetric analysis. The oxidized samples were characterized using X-ray diffraction, and the surface and cross-sectional morphologies were examined using scanning electron microscopy coupled with energy-dispersive X-ray spectroscopy. The forged 12Cr-MoVW steel samples exhibited parabolic behavior and a low oxidation rate compared with their as-cast counterparts. A protective oxide layer was uniformly formed at relatively low temperature (≤973 K for the forged samples, which thus exhibited better oxidation resistance than the as-cast ones. These oxides are considered solid-solution compounds such as (Fe, Cr2O3.

  10. High temperature low cycle fatigue behavior of Ni-base superalloy M963

    International Nuclear Information System (INIS)

    He, L.Z.; Zheng, Q.; Sun, X.F.; Guan, H.R.; Hu, Z.Q.; Tieu, A.K.; Lu, C.; Zhu, H.T.

    2005-01-01

    The cyclic stress-strain response and the low cycle fatigue life behavior of solution treated Ni-base superalloy M963 were studied. Fully reversed strain-controlled tests were performed at temperature range from 700 to 950 deg. C in air at a constant total strain rate. The dislocation characteristics and failed surface observation were evaluated through scanning electron microscopy and transmission electron microscopy, respectively. The alloy exhibited the cyclic hardening, softening, or stable cyclic stress response, which was dependent on the temperature and total strain range. The fracture surface observation revealed that fatigue crack initiation was transgranular and closely related to the total strain range; however, fatigue crack propagation exhibited a strong dependence on testing temperature. The dramatic reduction in fatigue life and intergranular cracking observed at 900 and 950 deg. C were attributed to oxidation

  11. High-temperature low cycle fatigue behavior of a gray cast iron

    Energy Technology Data Exchange (ETDEWEB)

    Fan, K.L., E-mail: 12klfan@tongji.edu.cn; He, G.Q.; She, M.; Liu, X.S.; Lu, Q.; Yang, Y.; Tian, D.D.; Shen, Y.

    2014-12-15

    The strain controlled low cycle fatigue properties of the studied gray cast iron for engine cylinder blocks were investigated. At the same total strain amplitude, the low cycle fatigue life of the studied material at 523 K was higher than that at 423 K. The fatigue behavior of the studied material was characterized as cyclic softening at any given total strain amplitude (0.12%–0.24%), which was attributed to fatigue crack initiation and propagation. Moreover, this material exhibited asymmetric hysteresis loops due to the presence of the graphite lamellas. Transmission electron microscopy analysis suggested that cyclic softening was also caused by the interactions of dislocations at 423 K, such as cell structure in ferrite, whereas cyclic softening was related to subgrain boundaries and dislocation climbing at 523 K. Micro-analysis of specimen fracture appearance was conducted in order to obtain the fracture characteristics and crack paths for different strain amplitudes. It showed that the higher the temperature, the rougher the crack face of the examined gray cast iron at the same total strain amplitude. Additionally, the microcracks were readily blunted during growth inside the pearlite matrix at 423 K, whereas the microcracks could easily pass through pearlite matrix along with deflection at 523 K. The results of fatigue experiments consistently showed that fatigue damage for the studied material at 423 K was lower than that at 523 K under any given total strain amplitude. - Highlights: • The low cycle fatigue behavior of the HT250 for engine cylinder blocks was investigated. • TEM investigations were conducted to explain the cyclic deformation response. • The low cycle fatigue cracks of HT250 GCI were studied by SEM. • The fatigue life of the examined material at 523 K is higher than that at 423 K.

  12. High temperature creep behavior in the (α + β) phase temperature range of M5 alloy

    International Nuclear Information System (INIS)

    Trego, G.

    2011-01-01

    The isothermal steady-state creep behavior of a M5 thin sheet alloy in a vacuum environment was investigated in the (α + β) temperature, low-stress (1-10 MPa) range. To this aim, the simplest approach consists in identifying α and β creep flow rules in their respective single-phase temperature ranges and extrapolating them in the two-phase domain. However, the (α + β) experimental behavior may fall outside any bounds calculated using such creep flow data. Here, the model was improved for each phase by considering two microstructural effects: (i) Grain size: Thermo-mechanical treatments applied on the material yielded various controlled grain size distributions. Creep tests in near-α and near-β ranges evidenced a strong grain-size effect, especially in the diffusional creep regime. (ii) Chemical contrast between the two phases in the (α + β) range: From thermodynamic calculations and microstructural investigations, the β phase is enriched in Nb and depleted in O (the reverse being true for the α phase). Thus, creep tests were performed on model Zr-Nb-O thin sheets with Nb and O concentrations representative of each phase in the considered temperature range. New α and β creep flow equations were developed from this extended experimental database and used to compute, via a finite element model, the creep rates of the two-phase material. The 3D morphology of phases (β grains nucleated at α grain boundaries) was explicitly introduced in the computations. The effect of phase morphology on the macroscopic creep flow was shown using this specific morphology, compared to other typical morphologies and to experimental data. (author) [fr

  13. SiC-based neutron detector in quasi-realistic working conditions: efficiency and stability at room and high temperature under fast neutron irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Ferone, Raffaello; Issa, Fatima; Ottaviani, Laurent; Biondo, Stephane; Vervisch, Vanessa [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231,13397 Marseille Cedex 20, (France); Szalkai, Dora; Klix, Axel [KIT- Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology Karlsruhe 76344, (Germany); Vermeeren, Ludo [SCK-CEN, Boeretang 200, B-2400 Mol, (Belgium); Saenger, Richard [Schlumberger, Clamart, (France); Lyoussi, Abadallah [CEA, DEN, Departement d' Etudes des Reacteurs, Service de Physique Experimentale, Laboratoire Dosimetrie Capteurs Instrumentation, 13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    In the framework of the European I SMART project, we have designed and made new SiC-based nuclear radiation detectors able to operate in harsh environments and to detect both fast and thermal neutrons. In this paper, we report experimental results of fast neutron irradiation campaign at high temperature (106 deg. C) in quasi-realistic working conditions. Our device does not suffer from high temperature, and spectra do show strong stability, preserving features. These experiments, as well as others in progress, show the I SMART SiC-based device skills to operate in harsh environments, whereas other materials would strongly suffer from degradation. Work is still demanded to test our device at higher temperatures and to enhance efficiency in order to make our device fully exploitable from an industrial point of view. (authors)

  14. Investigation of high temperature corrosion behavior on 304L austenite stainless steel in corrosive environments

    Energy Technology Data Exchange (ETDEWEB)

    Sahri, M. I.; Othman, N. K.; Samsu, Z.; Daud, A. R. [School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi Selangor (Malaysia)

    2014-09-03

    In this work, 304L stainless steel samples were exposed at 700 °C for 10hrs in different corrosive environments; dry oxygen, molten salt, and molten salt + dry oxygen. The corrosion behavior of samples was analyzed using weight change measurement technique, optical microscope (OM) and Scanning Electron Microscope (SEM) equipped with Energy Dispersive X-ray (EDX). The existence phases of corroded sample were determined using X-ray Diffraction (XRD). The lowest corrosion rate was recorded in dry oxygen while the highest was in molten salt + dry oxygen environments with the value of 0.0062 mg/cm{sup 2} and −13.5225 mg/cm{sup 2} respectively. The surface morphology of sample in presence of salt mixture showed scale spallation. Oxide scales of Fe{sub 3}O{sub 4}, Fe{sub 2}O{sub 3} were the main phases developed and detected by XRD technique. Cr{sub 2}O{sub 3} was not developed in every sample as protective layers but chromate-rich oxide was developed. The cross-section analysis found the oxide scales were in porous, thick and non-adherent that would not an effective barrier to prevent from further degradation of alloy. EDX analysis also showed the Cr-element was low compared to Fe-element at the oxide scale region.

  15. High Temperature Oxidation Behavior of Zirconium Alloy with Nano structured Oxide Layer in Air Environment

    International Nuclear Information System (INIS)

    Park, Y. J.; Kim, J. W.; Park, J. W.; Cho, S. O.

    2016-01-01

    If the temperature of the cladding materials increases above 1000 .deg. C, which can be caused by a loss of coolant accident (LOCA), Zr becomes an auto-oxidation catalyst and hence produces a huge amount of hydrogen gas from water. Therefore, many investigations are being carried out to prevent (or reduce) the hydrogen production from Zr-based cladding materials in the nuclear reactors. Our team has developed an anodization technique by which nanostructured oxide can be formed on various flat metallic elements such as Al, Ti, and Zr-based alloy. Anodization is a simple electrochemical technique and requires only a power supply and an electrolyte. In this study, Zr-based alloys with nanostructured oxide layers were oxidized by using Thermogravimetry analysis (TGA) and compared with the pristine one. It reveals that the nanostructured oxide layer can prevent oxidation of substrate metal in air. Oxidation behavior of the pristine Zr-Nb-Sn alloy and the Zr-Nb-Sn alloy with nanostructured oxide layer evaluated by measuring weight gain (TGA). In comparison with the pristine Zr-Nb-Sn alloy, weight gain of the Zr-Nb-Sn alloy with nanostructured oxide layer is lower than 10% even for 12 hours oxidation in air.

  16. Oxidation Behavior of Some Cr Ferritic Steels for High Temperature Fuel Cells

    International Nuclear Information System (INIS)

    Mohamed, H.E.

    2012-01-01

    The oxidation behavior of three high Cr ferritic steels designated 1Al, RA and 5Al with different levels of Al, Si, Mn and Hf has been investigated in the present work. These steels have been developed as candidates for Solid Oxide Fuel Cell (SOFC) interconnect. Specimens of these alloys have been subjected to isothermal as well as cyclic oxidation in air. Isothermal oxidation tests are conducted in the temperature range 800 - 1000 degree C for time periods up to 1000 h. cyclic oxidation tests were carried out at 800 and 1000 degree C for twenty 25 - h cycles giving a total cyclic exposure time of 500 h. The growth rate of the oxide scales was found to follow a parabolic law over a certain oxidation period which changed with alloy composition and oxidation temperature. The value of the parabolic rate constant increased with increasing oxidation temperature. At 800 and 900 degree C alloy 1Al exhibited higher oxidation resistance compared to the other two alloys. Alloy RA showed spalling behavior when oxidized at 900 degree C and the extent of spalling increased with increasing the oxidation temperature to 1000 degree C. Alloy 5Al oxidized at 1000 degree C showed the highest oxidation resistance among the investigated alloys. Alloy 1Al and RA showed similar scale morphology and composition. X- ray diffraction analysis revealed that the scales developed on these alloys consist of Cr 2 O 3 with an outer layer of MnCr 2 O 4 and a minor amount of FeCr 2 O 4 spinels. Alloy 5Al developed scale consisting of γ- Al 2 O 3 at 800 degree C and γ and α- Al 2 O 3 at 900 degree C. Oxidation of alloy 5Al at 1000 degree C led to formation of a scale consisting mainly of the protective phase α Al 2 O 3 . The presence of 0.84 wt% Al and 0.95 wt % Si in alloy 1Al enhanced its oxidation resistance compared to alloy RA which contains only 0.29 wt% Si and is Al - free. This enhancement was attributed to formation of internal oxidation zone in alloy 1Al just beneath the oxide / alloy

  17. High temperature oxidation behavior of austenitic stainless steel AISI 304 in steam of nanofluids contain nanoparticle ZrO2

    International Nuclear Information System (INIS)

    Prajitno, Djoko Hadi; Syarif, Dani Gustaman

    2014-01-01

    The objective of this study is to evaluate high temperature oxidation behavior of austenitic stainless steel SS 304 in steam of nanofluids contain nanoparticle ZrO 2 . The oxidation was performed at high temperatures ranging from 600 to 800°C. The oxidation time was 60 minutes. After oxidation the surface of the samples was analyzed by different methods including, optical microscope, scanning electron microscope (SEM) and X-ray diffraction (XRD). X-ray diffraction examination show that the oxide scale formed during oxidation of stainless steel AISI 304 alloys is dominated by iron oxide, Fe 2 O 3 . Minor element such as Cr 2 O 3 is also appeared in the diffraction pattern. Characterization by optical microscope showed that cross section microstructure of stainless steel changed after oxidized with the oxide scale on the surface stainless steels. SEM and x-ray diffraction examination show that the oxide of ZrO 2 appeared on the surface of stainless steel. Kinetic rate of oxidation of austenite stainless steel AISI 304 showed that increasing oxidation temperature and time will increase oxidation rate

  18. High temperature oxidation behavior of austenitic stainless steel AISI 304 in steam of nanofluids contain nanoparticle ZrO2

    Energy Technology Data Exchange (ETDEWEB)

    Prajitno, Djoko Hadi, E-mail: djokohp@batan.go.id; Syarif, Dani Gustaman, E-mail: djokohp@batan.go.id [Research Center for Nuclear Materials and Radiometry, Jl. Tamansari 71, Bandung 40132 (Indonesia)

    2014-03-24

    The objective of this study is to evaluate high temperature oxidation behavior of austenitic stainless steel SS 304 in steam of nanofluids contain nanoparticle ZrO{sub 2}. The oxidation was performed at high temperatures ranging from 600 to 800°C. The oxidation time was 60 minutes. After oxidation the surface of the samples was analyzed by different methods including, optical microscope, scanning electron microscope (SEM) and X-ray diffraction (XRD). X-ray diffraction examination show that the oxide scale formed during oxidation of stainless steel AISI 304 alloys is dominated by iron oxide, Fe{sub 2}O{sub 3}. Minor element such as Cr{sub 2}O{sub 3} is also appeared in the diffraction pattern. Characterization by optical microscope showed that cross section microstructure of stainless steel changed after oxidized with the oxide scale on the surface stainless steels. SEM and x-ray diffraction examination show that the oxide of ZrO{sub 2} appeared on the surface of stainless steel. Kinetic rate of oxidation of austenite stainless steel AISI 304 showed that increasing oxidation temperature and time will increase oxidation rate.

  19. Final report on neutron irradiation at low temperature to investigate plastic instability and at high temperature to study caviation

    DEFF Research Database (Denmark)

    Singh, B.N; Eldrup, Morten Mostgaard; Golubov, D.J.

    2005-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at333 and 573 K. Electrical resistivity......, based on the production bias model (PBM) were carried out to study the details of evolution of cavitieswith and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been...... studied using the Kinetic Monte Carlo (KMC) code during arealistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops...

  20. Final report on neutron irradiation at low temperature to investigate plastic instability and a high temperature to study cavitation

    International Nuclear Information System (INIS)

    Singh, B.N.; Eldrup, M.; Golubov, S.I.; Edwards, D.J.; Jung, P.

    2005-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at 333 and 573 K. Electrical resistivity and tensile properties were measured both in the unirradiated and irradiated condition. Some additional specimens of pure iron were irradiated at 333 K to doses of 10-3, 10-2 and 10-1 dpa and tensile tested at 333 K. To investigate the effect of helium on cavity nucleation and growth, specimens of pure iron and EUROFER 97 were implanted with different amounts of helium at 323 K and subsequently neutron irradiated to doses of 10-3, 10-2 and 10-1 dpa at 323 K. Defect microstructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations, based on the production bias model (PBM) were carried out to study the details of evolution of cavities with and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been studied using the Kinetic Monte Carlo (KMC) code during a realistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops.The present report describes both experimental procedure and calculational methodology employed in the present work. The main results of all these investigations, both experimental and theoretical, are highlighted with appropriate examples. Finally, a brief summary is given of the main results conclusions. (au)

  1. Fatigue Crack Propagation Behavior of RC Beams Strengthened with CFRP under High Temperature and High Humidity Environment

    Directory of Open Access Journals (Sweden)

    Dongyang Li

    2017-01-01

    Full Text Available Numerical and experimental methods were applied to investigate fatigue crack propagation behavior of reinforced concrete (RC beams strengthened with a new type carbon fiber reinforced polymer (CFRP named as carbon fiber laminate (CFL subjected to hot-wet environment. J-integral of a central crack in the strengthened beam under three-point bending load was calculated by ABAQUS. In finite element model, simulation of CFL-concrete interface was based on the bilinear cohesive zone model under hot-wet environment and indoor atmosphere. And, then, fatigue crack propagation tests were carried out under high temperature and high humidity (50°C, 95% R · H environment pretreatment and indoor atmosphere (23°C, 78% R · H to obtain a-N curves and crack propagation rate, da/dN, of the strengthened beams. Paris-Erdogan formula was developed based on the numerical analysis and environmental fatigue tests.

  2. COMPARATIVE ANALYSIS OF THE BEHAVIOR OF COAXIAL AND FRONTAL COUPLINGS – WITH PERMANENT MAGNETS – IN HIGH TEMPERATURE ENVIRONMENTS

    Directory of Open Access Journals (Sweden)

    Marcel Oanca

    2004-12-01

    Full Text Available This paper presents a comparative analysis of the behavior of coaxial and frontal couplings – with permanent magnets – in high temperature environments specific to iron and steel industry. The comparative analysis is made at the level of the specific forces developed in the most difficult environments. The maximum temperature was limited for reasons of thermal stability of the Nd-Fe-B permanent magnets. In this context it was studied, by the help of the PDE-ase soft that uses the finite element method, the way magnetic induction modifies, the specific forces developed and the distribution of temperature within the coaxial and frontal couplers with permanent magnets, for variations of the distance between the magnets (air gap within the limits 2-20 mm.

  3. Characterization technique for inhomogeneous 4H-SiC Schottky contacts: A practical model for high temperature behavior

    Science.gov (United States)

    Brezeanu, G.; Pristavu, G.; Draghici, F.; Badila, M.; Pascu, R.

    2017-08-01

    In this paper, a characterization technique for 4H-SiC Schottky diodes with varying levels of metal-semiconductor contact inhomogeneity is proposed. A macro-model, suitable for high-temperature evaluation of SiC Schottky contacts, with discrete barrier height non-uniformity, is introduced in order to determine the temperature interval and bias domain where electrical behavior of the devices can be described by the thermionic emission theory (has a quasi-ideal performance). A minimal set of parameters, the effective barrier height and peff, the non-uniformity factor, is associated. Model-extracted parameters are discussed in comparison with literature-reported results based on existing inhomogeneity approaches, in terms of complexity and physical relevance. Special consideration was given to models based on a Gaussian distribution of barrier heights on the contact surface. The proposed methodology is validated by electrical characterization of nickel silicide Schottky contacts on silicon carbide (4H-SiC), where a discrete barrier distribution can be considered. The same method is applied to inhomogeneous Pt/4H-SiC contacts. The forward characteristics measured at different temperatures are accurately reproduced using this inhomogeneous barrier model. A quasi-ideal behavior is identified for intervals spanning 200 °C for all measured Schottky samples, with Ni and Pt contact metals. A predictable exponential current-voltage variation over at least 2 orders of magnitude is also proven, with a stable barrier height and effective area for temperatures up to 400 °C. This application-oriented characterization technique is confirmed by using model parameters to fit a SiC-Schottky high temperature sensor's response.

  4. Irradiation behaviors of coated fuel particles, (4)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Ogawa, Toru; Ikawa, Katsuichi; Iwamoto, Kazumi; Ishimoto, Kiyoshi

    1981-09-01

    Loose coated fuel particles prepared in confirmity to a preliminary design for the multi-purpose VHTR in fiscal 1972 - 1974 were irradiated by 73F - 12A capsule in JMTR. Main purpose for this irradiation experiment was to examine irradiation stability of the candidate TRISO coated fuel particles for the VHTR. Also the coated particles possessing low-density kernel (90%TD), highly anisotropic OLTI-PyC and ZrC coating layer were loaded with the candidate particles in this capsule. The coated particles were irradiated up to 1.5 x 10 21 n/cm 2 of fast neutron fluence (E > 0.18 MeV) and 3.2% FIMA of burnup. In the post irradiation examination it was observed that among three kinds of TRISO particles exposed to irradiation corresponding to the normal operating condition of the VHTR ones possessing poor characteristics of the coating layers did not show a good stability. The particles irradiated under abnormally high temperature condition (> 1800 0 C) revealed 6.7% of max. EOL failure fraction (95% confidence limit). Most of these particles were failed by the ameoba effect. Furthermore, among four kinds of the TRISO particles exposed to irradiation corresponding to the transient condition of the VHTR (--1500 0 C) the two showed a good stability, while the particles possessing highly anisotropic OLTI-PyC or poorly characteristic coating layers were not so good. (author)

  5. Relationship Between Unusual High-Temperature Fatigue Crack Growth Threshold Behavior in Superalloys and Sudden Failure Mode Transitions

    Science.gov (United States)

    Telesman, J.; Smith, T. M.; Gabb, T. P.; Ring, A. J.

    2017-01-01

    An investigation of high temperature cyclic fatigue crack growth (FCG) threshold behavior of two advanced nickel disk alloys was conducted. The focus of the study was the unusual crossover effect in the near-threshold region of these type of alloys where conditions which produce higher crack growth rates in the Paris regime, produce higher resistance to crack growth in the near threshold regime. It was shown that this crossover effect is associated with a sudden change in the fatigue failure mode from a predominant transgranular mode in the Paris regime to fully intergranular mode in the threshold fatigue crack growth region. This type of a sudden change in the fracture mechanisms has not been previously reported and is surprising considering that intergranular failure is typically associated with faster crack growth rates and not the slow FCG rates of the near-threshold regime. By characterizing this behavior as a function of test temperature, environment and cyclic frequency, it was determined that both the crossover effect and the onset of intergranular failure are caused by environmentally driven mechanisms which have not as yet been fully identified. A plausible explanation for the observed behavior is proposed.

  6. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  7. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    International Nuclear Information System (INIS)

    Phillpot, Simon; Tulenko, James

    2011-01-01

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  8. Effect of pre-oxidation on high temperature sulfidation behavior of FeCr and FeCrAl alloys

    Directory of Open Access Journals (Sweden)

    Pillis Marina Fuser

    2004-01-01

    Full Text Available High temperature corrosion of structural alloys in sulfur bearing environments is many orders of magnitude higher than in oxidizing environments. Efforts to increase sulfidation resistance of these alloys include addition of alloying elements. Aluminum additions to iron-chromium alloys bring about increase in sulfidation resistance. This paper reports the effect of pre-oxidation on the sulfidation behavior of Fe-20Cr and Fe-20Cr-5Al alloys in H2-2% H2S environment at 800 °C. The surfaces of sulfidized specimens were also examined. Pre-oxidation of the two alloys results in an incubation period during subsequent sulfidation. After this incubation period, the Fe-20Cr alloy showed sulfidation behavior similar to that when the alloy was not pre-oxidized. The incubation period during sulfidation of the Fe-20Cr-5Al alloy was significantly longer, over 45 h, compared to 2 h for the Al free alloy. Based on the microscopic and gravimetric data a mechanism for sulfidation of these alloys with pre-oxidation has been proposed.

  9. High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

    Energy Technology Data Exchange (ETDEWEB)

    Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

    1965-06-28

    Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

  10. In-reactor behaviour of centrifugally atomized U3Si dispersion fuel irradiated at high temperature in HANARO

    International Nuclear Information System (INIS)

    Kim, Ki Hwan; Park, Jong Man; Yoo, Byeong Ok; Park, Dae Kyu; Lee, Choong Sung; Kim, Chang Kyu

    2002-01-01

    The irradiation test on full-size U 3 Si dispersion fuel elements, prepared by centrifugal atomization and conventional comminution method, has been performed up to about 77 at.% U-235 in maximum burn-up at CT hole position having the highest power condition in the HANARO reactor, in order to examine the irradiation performance of the atomized U 3 Si for the driver fuels of HANARO. The in-reactor interaction of the atomized U 3 Si dispersion fuel meats is generally assumed to be acceptable with the range of 5-15 μm in average thickness. The atomized spherical particles have more uniform and thinner reaction layer than the comminuted irregular particles. The U 3 Si particles have relatively fine and uniform size distribution of fission gas bubbles, irrespective of the powdering method. The bubble population in the atomized particles appears to be finer and more homogeneous with the characteristics of narrower bubble size distribution than that of the comminuted fuel. The atomized U 3 Si dispersion fuel elements exhibit sound swelling behaviours of 5 % in ΔV/V m even at ∼77 at.% U-235 burn-up, which meets with the safety criterion of the fuel rod, 20vol.% for HANARO. The atomized U3Si dispersion fuel elements show smaller swelling than the comminuted fuel elements

  11. High temperature in combination with UV irradiation enhances horizontal transfer of stx2 gene from E. coli O157:H7 to non-pathogenic E. coli.

    Directory of Open Access Journals (Sweden)

    Wan-Fu Yue

    Full Text Available Shiga toxin (stx genes have been transferred to numerous bacteria, one of which is E. coli O157:H7. It is a common belief that stx gene is transferred by bacteriophages, because stx genes are located on lambdoid prophages in the E. coli O157:H7 genome. Both E. coli O157:H7 and non-pathogenic E. coli are highly enriched in cattle feedlots. We hypothesized that strong UV radiation in combination with high temperature accelerates stx gene transfer into non-pathogenic E. coli in feedlots.E. coli O157:H7 EDL933 strain were subjected to different UV irradiation (0 or 0.5 kJ/m(2 combination with different temperature (22, 28, 30, 32, and 37 °C treatments, and the activation of lambdoid prophages was analyzed by plaque forming unit while induction of Stx2 prophages was quantified by quantitative real-time PCR. Data showed that lambdoid prophages in E. coli O157:H7, including phages carrying stx2, were activated under UV radiation, a process enhanced by elevated temperature. Consistently, western blotting analysis indicated that the production of Shiga toxin 2 was also dramatically increased by UV irradiation and high temperature. In situ colony hybridization screening indicated that these activated Stx2 prophages were capable of converting laboratory strain of E. coli K12 into new Shiga toxigenic E. coli, which were further confirmed by PCR and ELISA analysis.These data implicate that high environmental temperature in combination with UV irradiation accelerates the spread of stx genes through enhancing Stx prophage induction and Stx phage mediated gene transfer. Cattle feedlot sludge are teemed with E. coli O157:H7 and non-pathogenic E. coli, and is frequently exposed to UV radiation via sunlight, which may contribute to the rapid spread of stx gene to non-pathogenic E. coli and diversity of shiga toxin producing E. coli.

  12. Irradiation environment and materials behavior

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1992-01-01

    Irradiation environment is unique for materials used in a nuclear energy system. Material itself as well as irradiation and environmental conditions determine the material behaviour. In this review, general directions of research and development of materials in an irradiation environment together with the role of materials science are discussed first, and then recent materials problems are described for energy systems which are already existing (LWR), under development (FBR) and to be realized in the future (CTR). Topics selected are (1) irradiation embrittlement of pressure vessel steels for LWRs, (2) high fluence performance of cladding and wrapper materials for fuel subassemblies of FBRs and (3) high fluence irradiation effects in the first wall and blanket structural materials of a fusion reactor. Several common topics in those materials issues are selected and discussed. Suggestions are made on some elements of radiation effects which might be purposely utilized in the process of preparing innovative materials. (J.P.N.) 69 refs

  13. Neutron irradiation and high temperature effects on amorphous Fe-based nano-coatings on steel - A macroscopic assessment

    Science.gov (United States)

    Simos, N.; Zhong, Z.; Dooryhee, E.; Ghose, S.; Gill, S.; Camino, F.; Şavklıyıldız, İ.; Akdoğan, E. K.

    2017-06-01

    The study revealed that loss of ductility in an amorphous Fe-alloy coating on a steel substrate composite structure was essentially prevented from occurring, following radiation with modest neutron doses of ∼2 × 1018 n/cm2. At the higher neutron dose of ∼2 × 1019, macroscopic stress-strain analysis showed that the amorphous Fe-alloy nanostructured coating, while still amorphous, experienced radiation-induced embrittlement, no longer offering protection against ductility loss in the coating-substrate composite structure. Neutron irradiation in a corrosive environment revealed exemplary oxidation/corrosion resistance of the amorphous Fe-alloy coating, which is attributed to the formation of the Fe2B phase in the coating. To establish the impact of elevated temperatures on the amorphous-to-crystalline transition in the amorphous Fe-alloy, electron microscopy was carried out which confirmed the radiation-induced suppression of crystallization in the amorphous Fe-alloy nanostructured coating.

  14. Neutron irradiation and high temperature effects on amorphous Fe-based nano-coatings on steel – A macroscopic assessment

    International Nuclear Information System (INIS)

    Simos, N.; Zhong, Z.; Dooryhee, E.; Ghose, S.; Gill, S.

    2017-01-01

    Here, this study revealed that loss of ductility in an amorphous Fe-alloy coating on a steel substrate composite structure was essentially prevented from occurring, following radiation with modest neutron doses of ~2 x 10 18 n/cm 2 . At the higher neutron dose of ~2 x 10 19 , macroscopic stress-strain analysis showed that the amorphous Fe-alloy nanostructured coating, while still amorphous, experienced radiation-induced embrittlement, no longer offering protection against ductility loss in the coating-substrate composite structure. Neutron irradiation in a corrosive environment revealed exemplary oxidation/corrosion resistance of the amorphous Fe-alloy coating, which is attributed to the formation of the Fe 2 B phase in the coating. To establish the impact of elevated temperatures on the amorphous-to-crystalline transition in the amorphous Fe-alloy, electron microscopy was carried out which confirmed the radiation-induced suppression of crystallization in the amorphous Fe-alloy nanostructured coating.

  15. Advances in high temperature chemistry

    CERN Document Server

    Eyring, Leroy

    1969-01-01

    Advances in High Temperature Chemistry, Volume 2 covers the advances in the knowledge of the high temperature behavior of materials and the complex and unfamiliar characteristics of matter at high temperature. The book discusses the dissociation energies and free energy functions of gaseous monoxides; the matrix-isolation technique applied to high temperature molecules; and the main features, the techniques for the production, detection, and diagnosis, and the applications of molecular beams in high temperatures. The text also describes the chemical research in streaming thermal plasmas, as w

  16. Swelling of uranium dioxide and deformation behavior of the fuel element at high temperature irradiation

    International Nuclear Information System (INIS)

    Gontar, A.S.; Gutnik, V.S.; Nelidov, M.V.; Nikolaev, Yu.V.

    1993-01-01

    As post-reactor investigations showed, significant difference of swelling rates is connected with the fact that swelling of UO 2 with the equiaxial structure is mainly the result of fission gas bubbles accumulation along grain boundaries, and, in the case of the column structure, with formation of fine bubbles inside grains. The data given testify to usefulness of such investigations to predict TFE lifetime. As proven in this study one can see rates of radial deformation of fuel element cladding of a multi-cell TFE as a result of UO 2 swelling. They were calculated using the code SDS. Typical sizes were taken for calculation: cladding diameter--20 mm, cladding temperature at the central section of the fuel element--1,900 K, energy generation rate--145 W/cm 3 . These parameters provide output electric power of the TFE 600 W at the active zone length--400 mm

  17. Fabrication processes of C/Sic composites for high temperature components in energy systems and investigation of their oxidation behavior

    International Nuclear Information System (INIS)

    El-Hakim, E.

    2004-01-01

    Carbon fibre-reinforced ceramic matrix composite are promising candidate materials for high temperature applications such as structural components in energy systems, fusion reactors and advanced gas turbine engines. C/C composites has low oxidation resistance at temperatures above 500degree. To overcome this low oxidation resistance a coating should be applied. Tenax HTA 5131 carbon fibres impregnated with phenolic resin and reinforced silicon carbide were modified by the addition of a coating layer of boron oxide, (suspended in Dyansil-40) for improving anti-oxidation properties of the composites.The oxidation behavior of carbon-silicon carbide composites coated with B 2 O 3 , as an protective layer former, in dry air has been studied in the temperature range 800- 1000 degree for 8 hrs and 16 hrs. The results show that the oxidation rates of the uncoated composites samples are higher than those of the coated composites. The uncoated samples exhibit the highest oxidation rate during the initial stages of oxidation. The composite coated with B 2 O 3 had a significantly improved oxidation resistance due to the formation of a barrier layer for oxygen diffusion. This improvement in the oxidation resistance is attributed to the blocking of the active sites for oxygen diffusion. The oxidation resistance of the coated composite is highly improved; the weight loss percentage of casted samples is 4.5-16% after 16-hrs oxidation in air while the weight loss of uncoated samples is about 60%. The results are supported by scanning electron microscopy

  18. Microstructure and High-temperature Wear Behavior of Hot-dipped Aluminized Coating on Different Substrate Materials

    Directory of Open Access Journals (Sweden)

    ZHOU De-qin

    2018-02-01

    Full Text Available The aluminized 45 and H13 steel were prepared via hot-dipped aluminizing and subsequently high-temperature diffusion treatment. The phase, morphology and composition of aluminized coating were characterized by XRD,SEM and EDS methods. Comparative study was performed on unlubricated sliding wear behavior of plating under different substrates on a pin-on-disc wear tester, and the wear mechanism was explored. The results show that the coating is composed of ductile phases FeAl and Fe3Al. Kikendall porosity parallel to the surface exists around the interface of the two phases; because of the carbide particles agglomeration, the bond between the coating and H13 steel is apparently inferior to that in the case of 45 steel; the aluminized 45 steel possesses an excellent wear resistance under 50-200N at 400℃, whereas mild-to-severe wear transition occurs when the temperature increases to 600℃. The wear rate of the aluminized H13 steel reaches the lowest at 400℃, then slightly increases at 600℃. The wear mechanisms of Fe-Al coating are mainly predominated by oxidative mild wear, whereas the extrusion wear prevails in the process for aluminized 45 steel at 600℃.

  19. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  20. High-temperature oxidation behavior of dense SiBCN monoliths: Carbon-content dependent oxidation structure, kinetics and mechanisms

    International Nuclear Information System (INIS)

    Li, Daxin; Yang, Zhihua; Jia, Dechang; Wang, Shengjin; Duan, Xiaoming; Zhu, Qishuai; Miao, Yang; Rao, Jiancun; Zhou, Yu

    2017-01-01

    Highlights: •The scale growth for all investigated monoliths at 1500 °C cannot be depicted by a linear or parabolic rate law. •The carbon-rich monoliths oxidize at 1500 °C according to a approximately linear weight loss equation. •The excessive carbon in SiBCN monoliths deteriorates the oxidation resistance. •The oxidation resistance stems from the characteristic oxide structures and increased oxidation resistance of BN(C). -- Abstract: The high temperature oxidation behavior of three SiBCN monoliths: carbon-lean SiBCN with substantial Si metal, carbon-moderate SiBCN and carbon-rich SiBCN with excessive carbon, was investigated at 1500 °C for times up to15 h. Scale growth for carbon-lean and −moderate monoliths at 1500 °C cannot be described by a linear or parabolic rate law, while the carbon-rich monoliths oxidize according to a approximately linear weight loss equation. The microstructures of the oxide scale compose of three distinct layers. The passivating layer of carbon and boron containing amorphous SiO 2 and increased oxidation resistance of BN(C) both benefit the oxidation resistance.

  1. Effect of the carbon dioxide pressure on the electrochemical behavior of 3Cr low alloyed steel at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Jia, Zhijun, E-mail: jiazhijunwin@163.com [Department of Chemical Engineering, University of Tsinghua, Beijing 100084 (China); Key Laboratory of Corrosion and Protection of Chinese Ministry of Education, University of Science and Technology Beijing, Beijing 100083 (China); Li, Xiaogang; Du, Cuiwei; Liu, Zhiyong; Gao, Jin [Key Laboratory of Corrosion and Protection of Chinese Ministry of Education, University of Science and Technology Beijing, Beijing 100083 (China)

    2012-10-15

    Electrochemical corrosion behavior of 3Cr steel in CO{sub 2}-containing solution at a high temperature was investigated by various electrochemical measurements and analysis as well as thermodynamic calculations of ionic concentrations and equilibrium electrode potentials. A conceptual model was developed to illustrate the electrochemical corrosion mechanism of 3Cr steel in the CO{sub 2}-containing sodium chloride solution. Comparing the corrosion potentials of 3Cr steel in the test solution under different CO{sub 2} pressures with the conceptual model, it is found that anodic reactions of the 3Cr steel contain a direct dissolution of Fe, and the formation of corrosion scales, FeCO{sub 3} and Cr(OH){sub 3}, by Fe+HCO{sub 3}{sup -}=FeCO{sub 3}+H{sup +}+2e and Cr + 3OH{sup -} = Cr(OH){sub 3}. With the CO{sub 2} pressure increasing, the corrosion potential has a positive shift. It indicates that the CO{sub 2} pressure has a greater effect on the cathodic reaction than that of anodic reaction. And the corrosion current has positive linear relationship with the increase of CO{sub 2} pressure. It is attributed to the concentration increasing of the reactants of the cathodic reaction. According to analysis of the electrochemical impedance spectroscopy, the scale forming reactions dominate the corrosion process when the CO{sub 2} pressure is lower than 0.6 MPa and the dissolution of Fe, followed by the consecutive mechanism with adsorbed intermediate products, takes up the dominant part in the anodic process when the CO{sub 2} pressure exceeds 0.6 MPa. -- Highlights: Black-Right-Pointing-Pointer A conceptual model is developed to illustrate the corrosion mechanism. Black-Right-Pointing-Pointer A good reference electrode which is used at high temperature is made. Black-Right-Pointing-Pointer Corrosion current has positive linear relationship with the increase of CO{sub 2} pressure. Black-Right-Pointing-Pointer CO{sub 2} pressure has a greater effect on cathodic reaction than

  2. High temperature tribological behaviors of (WAl)C–Co ceramic composites with the additions of fluoride solid lubricants

    International Nuclear Information System (INIS)

    Cheng, Jun; Qiao, Zhuhui; Yin, Bing; Hao, Junying; Yang, Jun; Liu, Weimin

    2015-01-01

    The tribological behaviors of the (W 0.67 Al 0.33 )C 0.67 –Co/fluoride (CaF 2 , BaF 2 , CaF 2 /BaF 2 ) composites against SiC ball from room temperature to 600 °C were investigated. A marked increase in the friction coefficient resulting from fluoride oxidation was observed as the temperature increased. The composites containing BaF 2 or (Ca, Ba)F 2 displayed better integrated wear resistance over a wide temperature range compared with (W 0.67 Al 0.33 )C 0.67 –Co/CaF 2 . The high temperature tribological characteristics of the three composites were distinct, which originated from the composition difference on the worn surfaces. First, the SiO 2 /SiC film formed on the worn surfaces of the composites with BaF 2 or (Ca, Ba)F 2 was favorable for their wear resistance. Second, the oxidation of WC matrix was an important factor influencing the wear resistance of the composites. When mixture oxides of WO 2 and WO 3 appeared on the surface, wear is severe. In addition, single WO 3 formed on the worn surfaces, appeared more adhesive to the underlying substrate and decreased the wear rate. - Highlights: • The composites containing BaF 2 or (Ca, Ba)F 2 exhibit better wear resistance. • The tribological behaviors are strongly correlated to surface composition. • The stoichiometry difference in the tungsten oxides leads to distinct wear rate. • The friction coefficient of the composites increases with the testing temperature

  3. High temperature tribological behaviors of (WAl)C–Co ceramic composites with the additions of fluoride solid lubricants

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jun; Qiao, Zhuhui, E-mail: zhqiao@licp.cas.cn; Yin, Bing; Hao, Junying, E-mail: jyhao@licp.cas.cn; Yang, Jun; Liu, Weimin

    2015-08-01

    The tribological behaviors of the (W{sub 0.67}Al{sub 0.33})C{sub 0.67}–Co/fluoride (CaF{sub 2}, BaF{sub 2}, CaF{sub 2}/BaF{sub 2}) composites against SiC ball from room temperature to 600 °C were investigated. A marked increase in the friction coefficient resulting from fluoride oxidation was observed as the temperature increased. The composites containing BaF{sub 2} or (Ca, Ba)F{sub 2} displayed better integrated wear resistance over a wide temperature range compared with (W{sub 0.67}Al{sub 0.33})C{sub 0.67}–Co/CaF{sub 2}. The high temperature tribological characteristics of the three composites were distinct, which originated from the composition difference on the worn surfaces. First, the SiO{sub 2}/SiC film formed on the worn surfaces of the composites with BaF{sub 2} or (Ca, Ba)F{sub 2} was favorable for their wear resistance. Second, the oxidation of WC matrix was an important factor influencing the wear resistance of the composites. When mixture oxides of WO{sub 2} and WO{sub 3} appeared on the surface, wear is severe. In addition, single WO{sub 3} formed on the worn surfaces, appeared more adhesive to the underlying substrate and decreased the wear rate. - Highlights: • The composites containing BaF{sub 2} or (Ca, Ba)F{sub 2} exhibit better wear resistance. • The tribological behaviors are strongly correlated to surface composition. • The stoichiometry difference in the tungsten oxides leads to distinct wear rate. • The friction coefficient of the composites increases with the testing temperature.

  4. Ruthenium release at high temperature from irradiated PWR fuels in various oxidising conditions. Main findings from the VERCORS program

    International Nuclear Information System (INIS)

    Ducros, G.; Pontillon, Y.; Malgouyres, P.P.; Taylor, P.; Dutheillet, Y.

    2005-01-01

    Fission product release and transport in case of PWR severe accident is a major topic in reactor safety assessment due to the potential radiological consequences for surrounding populations and the environment. In this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the VERCORS analytical test program which was performed by the ''Commissariat a l'Energie Atomique'' (CEA). It is usually considered as complementary to the PHEBUS FP in-pile integral experimental program. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions).The influence of the nature of the fuel (UO 2 versus MOX, burn-up) and the fuel morphology (initially intact or fragmented fuels) have also been investigated. These led to an extended data base allowing on the one hand to study mechanisms which promote fission products release, and on the other hand to enhance models implemented in severe accident codes. Among all the fission products investigated, ruthenium is of specific concern because of its high radiological effects due essentially to the combination of both its short and long half-life isotopes (i.e. 103 Ru and 106 Ru respectively), but also by its ability to generate volatile gaseous oxides (RuO 3 , RuO 4 ) in very oxidising conditions, in particular in the case of air ingress accidents. Important uncertainties still remain on the release and transport of this element in such situations, and investigations on this open issue are notably carried out in the SARNET European framework. The present communication gives a general overview of the VERCORS program and presents more deeply the main findings concerning the ruthenium release. Its global behaviour is analysed on the basis of several comparative tests: same UO 2 sample (35 and 50 GWd/t) under hydrogen or steam conditions, similar MOX sample (40 GWd/t) under

  5. Viscoplastic behavior of uranium dioxide at high temperature; Comportement viscoplastique du dioxyde d'uranium a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sauter, F

    2001-02-01

    This work is a part of a project led by EDF the purpose of which is the development of more predictive models to describe the thermomechanical behavior of fuel assembly. First, we recall the baselines of the Power Water Reactors then we deal with the viscoplastic behavior of uranium dioxide (UO{sub 2}). This knowledge enables an accurate description of the stress relaxation during Pellet Cladding Interactions. The pellets we have used in the last part are similar to the industrial ones. They exhibit a yield point during strain hardening tests and a sigma creep curve. In order to describe these characteristics, we have adapted different kind of approaches: thermodynamical - the Distribution of Non Linear Relaxations, approaches based on dislocation glide inspired by Alexander and Haasen and introduced in the Pilvin polycrystalline model. We recall the purpose of internal variables in the thermodynamics of system far from equilibrium then in case of a viscoplastic flow controlled by dislocation glide, we establish a link between densities of dislocations and internal variables in the D.N.L.R. approach. As vacancy diffusion in the grain boundary has a contribution to the viscoplastic strain, a similar is presented in appendix. These models are able to reproduce the behavior of UO{sub 2} pellets in strain hardening, stress relaxation and creep tests. Much possible progress has been revealed by the analysis of the tests. Further more, we propose a model for yield point and sigma creep curve. We also have extended these results to the behavior of irradiated pellets and stressed the influence of damage. (author)

  6. Processing, Microstructure and Creep Behavior of Mo-Si-B-Based Intermetallic Alloys for Very High Temperature Structural Applications

    Energy Technology Data Exchange (ETDEWEB)

    Vijay Vasudevan

    2008-03-31

    This research project is concerned with developing a fundamental understanding of the effects of processing and microstructure on the creep behavior of refractory intermetallic alloys based on the Mo-Si-B system. In the first part of this project, the compression creep behavior of a Mo-8.9Si-7.71B (in at.%) alloy, at 1100 and 1200 C was studied, whereas in the second part of the project, the constant strain rate compression behavior at 1200, 1300 and 1400 C of a nominally Mo-20Si-10B (in at.%) alloy, processed such as to yield five different {alpha}-Mo volume fractions ranging from 5 to 46%, was studied. In order to determine the deformation and damage mechanisms and rationalize the creep/high temperature deformation data and parameters, the microstructure of both undeformed and deformed samples was characterized in detail using x-ray diffraction, scanning electron microscopy (SEM) with back scattered electron imaging (BSE) and energy dispersive x-ray spectroscopy (EDS), electron back scattered diffraction (EBSD)/orientation electron microscopy in the SEM and transmission electron microscopy (TEM). The microstructure of both alloys was three-phase, being composed of {alpha}-Mo, Mo{sub 3}Si and T2-Mo{sub 5}SiB{sub 2} phases. The values of stress exponents and activation energies, and their dependence on microstructure were determined. The data suggested the operation of both dislocation as well as diffusional mechanisms, depending on alloy, test temperature, stress level and microstructure. Microstructural observations of post-crept/deformed samples indicated the presence of many voids in the {alpha}-Mo grains and few cracks in the intermetallic particles and along their interfaces with the {alpha}-Mo matrix. TEM observations revealed the presence of recrystallized {alpha}-Mo grains and sub-grain boundaries composed of dislocation arrays within the grains (in Mo-8.9Si-7.71B) or fine sub-grains with a high density of b = 1/2<111> dislocations (in Mo-20Si-10B), which

  7. Effect of microstructure on high-temperature mechanical behavior of nickel-base superalloys for turbine disc applications

    Science.gov (United States)

    Sharpe, Heather Joan

    2007-05-01

    Engineers constantly seek advancements in the performance of aircraft and power generation engines, including, lower costs and emissions, and improved fuel efficiency. Nickel-base superalloys are the material of choice for turbine discs, which experience some of the highest temperatures and stresses in the engine. Engine performance is proportional to operating temperatures. Consequently, the high-temperature capabilities of disc materials limit the performance of gas-turbine engines. Therefore, any improvements to engine performance necessitate improved alloy performance. In order to take advantage of improvements in high-temperature capabilities through tailoring of alloy microstructure, the overall objectives of this work were to establish relationships between alloy processing and microstructure, and between microstructure and mechanical properties. In addition, the projected aimed to demonstrate the applicability of neural network modeling to the field of Ni-base disc alloy development and behavior. The first phase of this work addressed the issue of how microstructure varies with heat treatment and by what mechanisms these structures are formed. Further it considered how superalloy composition could account for microstructural variations from the same heat treatment. To study this, four next-generation Ni-base disc alloys were subjected to various controlled heat-treatments and the resulting microstructures were then quantified. These quantitative results were correlated to chemistry and processing, including solution temperature, cooling rate, and intermediate hold temperature. A complex interaction of processing steps and chemistry was found to contribute to all features measured; grain size, precipitate distribution, grain boundary serrations. Solution temperature, above a certain threshold, and cooling rate controlled grain size, while cooling rate and intermediate hold temperature controlled precipitate formation and grain boundary serrations. Diffusion

  8. High-light damage in air-dry thalli of the old forest lichen Lobaria pulmonaria - interactions of irradiance, exposure duration and high temperature

    International Nuclear Information System (INIS)

    Gauslaa, Y.; Solhaug, K.A.

    1999-01-01

    High-light damage in air-dry thalli of Lobaria pulmonaria were measured in the laboratory as reductions in maximal PSII efficiency (FV/FM) after a 48 h recovery in a hydrated state at low light to account for permanent damage. Thalli treated with the lowest light dose (90 mol photons m −2 ) recovered normal FV/FM-values with increasing irradiances (400–700 nm) up to 1000 µmol photons m −2 s −1 . Doubling this dose lowered the threshold level for damage from 1000 to 320 µmol photons m −2 s −1 , and reduced FV/FM at 1000 µmol photons m −2 s −1 by more than 50%. A second doubling of the dose to 360 mol photons m −2 caused damage at 200 µmol photons m −2 s −1 , and a nearly complete cessation of PSII efficiency occurred at 1000 µmol photons m −2 s −1 . No reciprocity of irradiance and duration of illumination for PSII function was found. The measured time-dependent decrease in FV/FM was remarkably similar for the naturally coupled, but artificially separated, light and temperature factors. Therefore, the damage of high light on desiccated L. pulmonaria seemed to be an additive effect of high irradiance and high temperatures. Air-dry thalli were highly heat susceptible, being affected already at temperatures around 40 °C. Logging operations in forests are likely to raise the solar radiation at remaining lichen sites to destructive levels. (author)

  9. High temperature materials

    International Nuclear Information System (INIS)

    2003-01-01

    The aim of this workshop is to share the needs of high temperature and nuclear fuel materials for future nuclear systems, to take stock of the status of researches in this domain and to propose some cooperation works between the different research organisations. The future nuclear systems are the very high temperature (850 to 1200 deg. C) gas cooled reactors (GCR) and the molten salt reactors (MSR). These systems include not only the reactor but also the fabrication and reprocessing of the spent fuel. This document brings together the transparencies of 13 communications among the 25 given at the workshop: 1) characteristics and needs of future systems: specifications, materials and fuel needs for fast spectrum GCR and very high temperature GCR; 2) high temperature materials out of neutron flux: thermal barriers: materials, resistance, lifetimes; nickel-base metal alloys: status of knowledge, mechanical behaviour, possible applications; corrosion linked with the gas coolant: knowledge and problems to be solved; super-alloys for turbines: alloys for blades and discs; corrosion linked with MSR: knowledge and problems to be solved; 3) materials for reactor core structure: nuclear graphite and carbon; fuel assembly structure materials of the GCR with fast neutron spectrum: status of knowledge and ceramics and cermets needs; silicon carbide as fuel confinement material, study of irradiation induced defects; migration of fission products, I and Cs in SiC; 4) materials for hydrogen production: status of the knowledge and needs for the thermochemical cycle; 5) technologies: GCR components and the associated material needs: compact exchangers, pumps, turbines; MSR components: valves, exchangers, pumps. (J.S.)

  10. Nitrogen compounds behavior under irradiation environment

    International Nuclear Information System (INIS)

    Ichikawa, Nagayoshi; Takagi, Junichi; Yotsuyanagi, Tadasu

    1991-01-01

    Laboratory experiments were performed to evaluate nitrogen compounds behavior in liquid phase under irradiation environments. Nitrogen compounds take a chemical form of ammonium ion under reducing condition by gamma irradiation, whereas ammonium ions are rather stable even under oxidizing conditions. Key reactions were pointed out and their reaction rate constants and activation energies were estimated through computer code simulation. A reaction scheme for nitrogen compounds including protonate reaction was proposed. (author)

  11. Evaluation of thermal displacement behavior of high temperature piping system in power-up test of HTTR. No. 1 results up to 20 MW operation

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Kojima, Takao; Sumita, Junya; Tachibana, Yukio

    2002-03-01

    Temperature of the primary cooling system of the High Temperature Engineering Test Reactor, HTTR, becomes very high because the coolant temperature at the reactor outlet reaches 950degC, and 400degC at inlet of the reactor. Therefore, it is important to confirm the thermal displacement behavior of the high temperature piping system in the primary cooling system from the viewpoint of the structural integrity. Moreover, newly designed 3-dimensional floating support system is adopted to the cooling system, it is meaningful to verify the thermal displacement behavior of the piping system applied the 3-dimensional floating support system. In the power-up test (up to 20 MW operation), thermal displacement behavior of the high temperature piping system was measured. This paper describes the experimental and analytical results of thermal displacement characteristics of the high temperature piping system. The results showed that the resistance force induced from the supporting system effects to the thermal displacement behavior of cooling system, and the analytical results have a good agreement with the experimental results by optimizing the resistant force of the floating support system. Additionally, structural integrity at the 30 MW operation was confirmed by the analysis. (author)

  12. A study on surface properties and high temperature oxidation behavior of ion nitrided FC-25 gray cast iron

    International Nuclear Information System (INIS)

    Hur, In Chang; Son, Kun Su; Yoon, Jae Hong; Cho, Tong Yul; Park, Bong Gyu; Kim, Hyun Soo; Kim, In Soo

    2005-01-01

    Surface properties and high temperature oxidation behavior were investigated for FC-25 Gray Cast Iron(GCI) and the ion intrided GCI(N-GCI). The GCI was pre-cleaned to improve hardness to the optimum pre-sputtering parameters with an Ar/H 2 ratio of 1/2, working pressure of 3 torr, working temperature of 550 .deg. C and working time of 1hour. The optimum nitriding conditions for the maximum hardness of 560∼575 Hv were an N 2 /H 2 ratio of 3/1, working pressure of 3 torr, and working temperature of 575 deg. C. The thickness of graphite in the GCI was increased by increasing the working temperature from 525 .deg. C to 595 .deg. C for the nitriding time of 6∼18hrs. XRD patterns showed FeO and Fe 2 O 3 peaks for both the oxidized N-GCI and GCI at temperature of 600 .deg. C and 800 .deg. C under atmospheric environment for both 24 and 60hours. At 800 .deg. C, above the Fe 4 N decomposition temperature of 680 .deg. C, the oxidation rate of N-GCI was greater than that of the GCI. The most abundant nitride, Fe 4 N, was decomposed and the nitrogen gas given off by the decomposition made the protective film porous by degassing through the film. But at 600 .deg. C, below the decomposition temperature, the degree of oxidation of N-GCI was lower than that of the GCI because the nitride film worked as protective barrier for oxidation. Finite element modeling of elastic contact wear problems was performed to demonstrate the feasibility of applying the finite element method to fretting wear problems. The elastic beam problem, with existing solutions, is treated as a numerical example. By introducing a control parameter s, which scaled up the wear constant and scaled down the cycle numbers, the algorithm was shown to greatly reduce the time required for the analysis. The work rate model was adopted in the wear model. In the three-dimensional finite element analysis, a quarterly symmetric model was used to simulate cross tubes contacting at right angles. The wear constant of

  13. Annealing experiments on and high-temperature behavior of the superconductor yttrium barium copper oxide (YBa2Cu3Ox)

    NARCIS (Netherlands)

    Brabers, V.A.M.; Jonge, de W.J.M.; Bosch, L.A.; Steen, van der C.; de Groote, A.M.W.; Verheyen, A.A.; Vennix, C.W.H.M.

    1988-01-01

    The high temperature behaviour (300–1100 K) of the superconductor YBa2Cu3Ox has been studied by annealing experiments, thermal dilatation, thermogravimetry and measurements of the electrical resistance and thermoelectric power. For the fast oxidation process of this compound, reaction enthalpies

  14. High temperature oxidation behavior of gamma-nickel+gamma'-nickel aluminum alloys and coatings modified with platinum and reactive elements

    Science.gov (United States)

    Mu, Nan

    Materials for high-pressure turbine blades must be able to operate in the high-temperature gases (above 1000°C) emerging from the combustion chamber. Accordingly, the development of nickel-based superalloys has been constantly motivated by the need to have improved engine efficiency, reliability and service lifetime under the harsh conditions imposed by the turbine environment. However, the melting point of nickel (1455°C) provides a natural ceiling for the temperature capability of nickel-based superalloys. Thus, surface-engineered turbine components with modified diffusion coatings and overlay coatings are used. Theses coatings are capable of forming a compact and adherent oxide scale, which greatly impedes the further transport of reactants between the high-temperature gases and the underlying metal and thus reducing attack by the atmosphere. Typically, these coatings contain beta-NiAl as a principal constituent phase in order to have sufficient aluminum content to form an Al2O3 scale at elevated temperatures. The drawbacks to the currently-used beta-based coatings, such as phase instabilities, associated stresses induced by such phase instabilities, and extensive coating/substrate interdiffusion, are major motivations in this study to seek next-generation coatings. The high-temperature oxidation resistance of novel Pt+Hf-modified gamma-Ni+gamma'-Ni 3Al-based alloys and coatings were investigated in this study. Both early-stage and 4-days isothermal oxidation behavior of single-phase gamma-Ni and gamma'-Ni3Al alloys were assessed by examining the weight changes, oxide-scale structures, and elemental concentration profiles through the scales and subsurface alloy regions. It was found that Pt promotes Al 2O3 formation by suppressing the NiO growth on both gamma-Ni and gamma'-Ni3Al single-phase alloys. This effect increases with increasing Pt content. Moreover, Pt exhibits this effect even at lower temperatures (˜970°C) in the very early stage of oxidation. It

  15. Thermodynamic studies on the ferroelectric phase transition in neutron irradiated (LixK1-x)2SO4 crystals at high temperature

    International Nuclear Information System (INIS)

    Kassem, M.E.; El-Khatib, A.M.; Ammar, E.A.; Denton, M.M.

    1989-05-01

    Thermodynamic studies of (Li x K 1-x ) 2 SO 4 , LKS, mixed crystals have been made in the concentration range (x=0.1,0.2,...,x=0.5). The thermal behavior has been investigated by differential thermal analysis, DTA, and differential scanning calorimeter, DSC, in the vicinity of high temperature phases. Also, the effect of the mixed neutron field of fast and thermal neutrons (10% of the reactor neutron pile is fast neutrons) on the thermal properties of mixed crystals was studied. The results showed a change in the transition temperature Tc, as well as the value of specific heat Cp at transition temperature, due to the change of stoichiometric ratio and radiation doses. The change of enthalpy and entropy of mixed crystals have been estimated numerically. The obtained small values of ΔS/R is characteristic of incommensurate phase transition as previously confirmed by the results of neutron diffraction technique. (author). 16 refs, 5 figs, 1 tab

  16. Investigation on cause of outage of Wide Range Monitor (WRM) in High Temperature engineering Test Reactor (HTTR). Post Irradiation Examination (PIE) toward investigation of the cause

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Takada, Shoji; Ishimi, Akihiro; Katsuyama, Kozo; Motegi, Toshihiro

    2012-08-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the outage by specifying the damaged part causing the event in the WRM. The one is a post irradiation examination using the X-ray computed tomography scanner in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. The other is an experiment using a mock-up simulating the WRM fabricated by the fabricator. The characteristic impedance of the damaged WRM was measured by Time Domain Reflectometry, which was compared with that of the mock-up, which could narrow down the damaged part in the WRM. This report summarized the results of the PIE and the experimental investigation using the mock-up to reveal the cause of outage of WRM. (author)

  17. The ambient and high temperature deformation behavior of Al–Si–Cu–Mg alloy with minor Ti, Zr, Ni additions

    International Nuclear Information System (INIS)

    Hernandez-Sandoval, J.; Garza-Elizondo, G.H.; Samuel, A.M.; Valtiierra, S.; Samuel, F.H.

    2014-01-01

    Highlights: • Characterization on the precipitation of Ni- and Zr-based intermetallics. • High temperature tensile properties of 354 alloy containing Zr and Ni below 0.5%. • Quality index charts as a function of heat treatment. • Yield strength and ductility color contours as a function of aging temperature and aging time. - Abstract: The principal aim of the present work was to investigate the effects of minor additions of nickel and zirconium on the strength of cast aluminum alloy 354 at ambient and high temperatures. Tensile properties of the as-cast and heat-treated alloys were determined at room temperature and at high temperatures (190 °C, 250 °C, 350 °C). The results show that Zr reacts only with Ti, Si and Al. From the quality index charts constructed for these alloys, the quality index attains minimum and maximum values of 259 MPa and 459 MPa, in the as-cast and solution-treated conditions; also, maximum and minimum values of yield strength are observed at 345 MPa and 80 MPa, respectively, within the series of aging treatments applied. A decrease in tensile properties of ∼10% with the addition of 0.4 wt.% nickel is attributed to a nickel–copper reaction. The reduction in mechanical properties due to addition of different elements is attributed principally to the increase in the percentage of intermetallic phase particles formed during solidification; such particles act as stress concentrators, decreasing the alloy ductility. Tensile test results at ambient temperatures show a slight increase (∼10%) in alloys with Zr and Zr/Ni additions, particularly at aging temperatures above 240 °C. Additions of Zr and Zr + Ni increase the high temperature tensile properties, in particular for the alloy containing 0.2 wt.% Zr + 0.2 wt.% Ni, which exhibits an increase of more than 30% in the tensile properties at 300 °C compared with the base 354 alloy

  18. Modeling the effect of water vapor on the interfacial behavior of high-temperature air in contact with Fe20Cr surfaces

    International Nuclear Information System (INIS)

    Chialvo, Ariel A.; Brady, Michael P.; Keiser, James R.; Cole, David R.

    2011-01-01

    Highlights: → Atomistic view of the contrasting interfacial behavior between high-temperature dry- and wet-air in contact with stainless steels. → H 2 O preferentially adsorbs and displaces oxygen at the metal-fluid interface. → Findings are consistent with Ehlers et al.'s proposed competitive adsorption mechanism for the interpretation of the breakaway oxidation. → Significant impact of the inhomogeneous density distribution between the interfacial- and bulk-environments on the fluid transport. -- This work uses molecular dynamics simulation to provide an atomistic view of the contrasting interfacial behavior between high-temperature dry air and wet (10-40 vol.% water) air in contact with stainless steels. A key finding was that H 2 O preferentially adsorbs and displaces oxygen at the metal-fluid interface. We also discuss how these findings are consistent with Ehlers et al. proposed competitive adsorption mechanism for the interpretation of the breakaway oxidation, and highlight their impact on other properties.

  19. Radionuclides release from re-irradiated fuel under high temperature and pressure conditions. Gamma-ray measurements of VEGA-5 test

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Akihide; Kudo, Tamotsu; Nakamura, Takehiko; Kanazawa, Toru; Kiuchi, Toshio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program is being performed at JAERI to clarify mechanisms of radionuclides release from irradiated fuel during severe accidents and to improve source term predictability. The fifth VEGA-5 test was conducted in January 2002 to confirm the reproducibility of decrease in cesium release under elevated pressure that was observed in the VEGA-2 test and to investigate the release behavior of short-life radionuclides. The PWR fuel of 47 GWd/tU after about 8.2 years of cooling was re-irradiated at Nuclear Safety Research Reactor (NSRR) for 8 hours before the heat-up test. After that, the two pellets of 10.9 g without cladding were heated up to about 2,900 K at 1.0 MPa under the inert He condition. The experiment reconfirmed the decrease in cesium release rate under the elevated pressure. The release data on short-life radionuclides such as Ru-103, Ba-140 and Xe-133 that have never been observed in the previous VEGA tests without re-irradiation was obtained using the {gamma} ray measurement. (author)

  20. Grain growth behavior and high-temperature high-strain-rate tensile ductility of iridium alloy DOP-26

    International Nuclear Information System (INIS)

    McKamey, C.G.; Gubbi, A.N.; Lin, Y.; Cohron, J.W.; Lee, E.H.; George, E.P.

    1998-04-01

    This report summarizes results of studies conducted to date under the Iridium Alloy Characterization and Development subtask of the Radioisotope Power System Materials Production and Technology Program to characterize the properties of the new-process iridium-based DOP-26 alloy used for the Cassini space mission. This alloy was developed at Oak Ridge National Laboratory (ORNL) in the early 1980's and is currently used by NASA for cladding and post-impact containment of the radioactive fuel in radioisotope thermoelectric generator (RTG) heat sources which provide electric power for interplanetary spacecraft. Included within this report are data generated on grain growth in vacuum or low-pressure oxygen environments; a comparison of grain growth in vacuum of the clad vent set cup material with sheet material; effect of grain size, test temperature, and oxygen exposure on high-temperature high-strain-rate tensile ductility; and grain growth in vacuum and high-temperature high-strain-rate tensile ductility of welded DOP-26. The data for the new-process material is compared to available old-process data

  1. Modification of tribology and high-temperature behavior of Ti-48Al-2Cr-2Nb intermetallic alloy by laser cladding

    International Nuclear Information System (INIS)

    Liu Xiubo; Wang Huaming

    2006-01-01

    In order to improve the tribology and high-temperature oxidation properties of the Ti-48Al-2Cr-2Nb intermetallic alloy simultaneously, mixed NiCr-Cr 3 C 2 precursor powders had been investigated for laser cladding treatment to modify wear and high-temperature oxidation resistance of the material. The alloy samples were pre-placed with NiCr-80, 50 and 20%Cr 3 C 2 (wt.%), respectively, and laser treated at the same parameters, i.e., laser output power 2.8 kW, beam scanning speed 2.0 mm/s, beam dimension 1 mm x 18 mm. The treated samples underwent tests of microhardness, wear and high-temperature oxidation. The results showed that laser cladding with different constitution of mixed precursor NiCr-Cr 3 C 2 powders improved surface hardness in all cases. Laser cladding with NiCr-50%Cr 3 C 2 resulted in the best modification of tribology and high-temperature oxidation behavior. X-ray diffraction (XRD), optical microscope (OM), scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) analyses indicated that the formation of reinforced Cr 7 C 3 , TiC and both continuous and dense Al 2 O 3 , Cr 2 O 3 oxide scales were supposed to be responsible for the modification of the relevant properties. As a result, the present work had laid beneficial surface engineering foundation for TiAl alloy applied as future light weight and high-temperature structural candidate materials

  2. Development and application of an empirical formula for the high temperature behavior of ferroelectric ceramics switched by electric field at room temperature

    Directory of Open Access Journals (Sweden)

    Dae Won Ji

    2017-05-01

    Full Text Available The strain changes during temperature rise of a poled lead titanate zirconate rectangular parallelepiped switched by electric field at room temperature are obtained by integrating thermal expansion coefficients that are measured using an invar-specimen. By estimating and analyzing pyroelectric and thermal expansion coefficients, first-order differential equations are constructed for polarization and strain changes during temperature increase. The solutions to the differential equations are found and used to calculate the high temperature behavior of the materials. It is shown that the predictions are well compared with measured responses. Finally, the developed formulae are applied to calculate strain butterfly loops from a polarization hysteresis loop at a high temperature.

  3. Deformation behavior of irradiated Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Himbeault, D.D.; Chow, C.K.; Puls, M.P.

    1994-01-01

    A study of the deformation behavior of irradiated highly textured Zr-2.5Nb pressure tube material in the temperature range of 30 degree C to 300 degree C was undertaken to understand better the mechanism for the deterioration of the fracture toughness with neutron irradiation. Strain localization behavior, believed to be a main contributor to reduced toughness, was observed in irradiated transverse tensile specimens at temperature greater than 100 degree C. The strain localization behavior was found to occur by the cooperative twinning of the highly textured grains of the material, resulting in a local softening of the material, where the flow than localizes. It is believed that the effect of the irradiation is to favor twinning at the expense of slip in the early stages of deformation. This effect becomes more pronounced at higher temperature, thus leading to the high-temperature strain localization behavior of the material. A limited amount of dislocation channeling was also observed; however, it is not considered to have a major role in the strain localization behavior of the material. Contrary to previous reports on irradiated zirconium alloys, static strain aging is observed in the irradiated material in the temperature range of 150 degree C to 300 degree C

  4. Supersymmetry at high temperatures

    International Nuclear Information System (INIS)

    Das, A.; Kaku, M.

    1978-01-01

    We investigate the properties of Green's functions in a spontaneously broken supersymmetric model at high temperatures. We show that, even at high temperatures, we do not get restoration of supersymmetry, at least in the one-loop approximation

  5. High temperature structural silicides

    International Nuclear Information System (INIS)

    Petrovic, J.J.

    1997-01-01

    Structural silicides have important high temperature applications in oxidizing and aggressive environments. Most prominent are MoSi 2 -based materials, which are borderline ceramic-intermetallic compounds. MoSi 2 single crystals exhibit macroscopic compressive ductility at temperatures below room temperature in some orientations. Polycrystalline MoSi 2 possesses elevated temperature creep behavior which is highly sensitive to grain size. MoSi 2 -Si 3 N 4 composites show an important combination of oxidation resistance, creep resistance, and low temperature fracture toughness. Current potential applications of MoSi 2 -based materials include furnace heating elements, molten metal lances, industrial gas burners, aerospace turbine engine components, diesel engine glow plugs, and materials for glass processing

  6. Fatigue crack growth behavior of pressure vessel steels and submerged arc weldments in a high-temperature pressurized water environment

    International Nuclear Information System (INIS)

    Liaw, P.K.; Logsdon, W.A.; Begley, J.A.

    1989-01-01

    The fatigue crack growth rate (FCGR) properties of SA508 Cl 2a and SA533 Gr A Cl 2 pressure vessel steels and the corresponding automatic submerged arc weldments were developed in a high-temperature pressurized water (HPW) environment at 288 degrees C (550 degrees F) and 7.2 MPa (1044 psi) at load ratios of 0.20 and 0.50. The properties were generally conservative compared to American Society of Mechanical Engineers Section XI water environment reference curve. The growth rate of fatigue cracks in the base materials, however, was faster in the HPW environment than in a 288 degrees C (550 degrees F) base line air environment. The growth rate of fatigue cracks in the two submerged arc weldments was also accelerated in the HPW environment but to a lesser degree than that demonstrated by the base materials. In the air environment, fatigue striations were observed, independent of material and load ratio, while in the HPW environment, some intergranular facets were present. The greater environmental effect on crack growth rates displayed by the base materials compared the weldments attributed to a different sulfide composition and morphology

  7. Behavior of high Tc-superconductors and irradiated defects under reactor irradiation

    International Nuclear Information System (INIS)

    Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Yoshida, Hiroyuki.

    1991-01-01

    It has been well known that the lattice defects of various types are introduced in ceramics without exception, and exert large effect to the function of these materials. Among oxides, the electronic materials positively using oxygen defect control have been already put in practical use. Also in the oxide high temperature superconductors which are Perovskite type composite oxides, the superconductive characteristics are affected largely by the concentration of the oxygen composing them. This is regarded as an important factor for causing superconductivity, related with the oxygen cavities arising at this time and the carriers bearing superconductivity. In this study, the irradiation effect with relatively low dose, the measurement under irradiation, the effect of irradiation temperature, and the effect of radiation quality were evaluated by the irradiation of YBCO, EBCO and LBCO. The experimental method, and the irradiation effect at low temperature and normal temperature, the effect of Co-60 gamma ray irradiation instead of reactor irradiation are reported. (K.I.)

  8. ''Cs-tetra-ferri-annite:'' High-pressure and high-temperature behavior of a potential nuclear waste disposal phase

    International Nuclear Information System (INIS)

    Comodi, P.; Zanazzi, P.F.

    1999-01-01

    Structure deformations induced by pressure and temperature in synthetic Cs-tetra-ferri-annite 1M [Cs 1.78 (Fe 2+ 5.93 Fe 3+ 0.07 )(Si 6.15 Fe 3+ 1.80 Al 0.05 )O 20 (OH) 4 ], space group C2/m, were analyzed to investigate the capability of the mica structure to store the radiogenic isotopes 135 Cs and 137 Cs. Cs-tetra-ferri-annite is not a mineral name, but for the sake of brevity is used here to designate a synthetic analog of the mineral tetra-ferri-annite. The bulk modulus and its pressure derivative determined by fitting the unit-cell volumes between 0 a/nd 47 kbar to a third-order Birch-Murnaghan equation of state are K 0 = 257(8) kbar and K' 0 = 21(1), respectively. Between 23 C and 582 C, the a and b lattice parameters remain essentially unchanged, but the thermal expansion coefficient of the c axis is α c = 3.12(9) x 10 -5 degree C -1 . High pressure (P) and high temperature (T) produce limited internal strain in the structure. The tetrahedral rotation angle, α, is very small and does not change significantly throughout the P and T range investigated. Above 450 C in air, Cs-tetra-ferri-annite underwent an oxidation of octahedral iron in the M2cis site, balanced by the loss of H and shown by a decrease of the unit-cell volume. Independent isobaric data on thermal expansion and isothermal compressibility data define the geometric equation of state for Cs-tetra-ferri-annite. On the whole, the data confirm that the structure of Cs-tetra-ferri-annite may be a suitable candidate for the storage of large ions, such as Cs in the interlayer and should be considered as a potential Synroc component

  9. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minato, Kazuo [Japan Atomic Energy Research Institute (Japan); Sawa, Kazuhiro [Japan Atomic Energy Research Institute (Japan); Koya, Toshio [Japan Atomic Energy Research Institute (Japan); Tomita, Takeshi [Japan Atomic Energy Research Institute (Japan); Ishikawa, Akiyoshi [Japan Atomic Energy Research Institute (Japan); Baldwin, Charles A; Gabbard, William Alexander [Oak Ridge National Laboratory (United States); Malone, Charlie M [Oak Ridge National Laboratory (United States)

    2000-07-15

    Postirradiation heating tests of TRISO-coated UO{sub 2} particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of {sup 85}Kr, {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, and {sup 154}Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations.

  10. Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Minato, Kazuo; Sawa, Kazuhiro; Koya, Toshio; Tomita, Takeshi; Ishikawa, Akiyoshi; Baldwin, Charles A.; Gabbard, William Alexander; Malone, Charlie M.

    2000-01-01

    Postirradiation heating tests of TRISO-coated UO 2 particles at 1700 and 1800degC were performed to understand fission product release behavior at accident temperatures. The inventory measurements of the individual particles were carried out before and after the heating tests with gamma-ray spectrometry to study the behavior of the individual particles. The time-dependent release behavior of 85 Kr, 110m Ag, 134 Cs, 137 Cs, and 154 Eu were obtained with on-line measurements of fission gas release and intermittent measurements of metallic fission product release during the heating tests. The inventory measurements of the individual particles revealed that fission product release behavior of the individual particles was not uniform, and large particle-to-particle variations in the release behavior of 110m Ag, 134 Cs, 137 Cs, and 154 Eu were found. X-ray microradiography and ceramography showed that the variations could not be explained by only the presence or absence of cracks in the SiC coating layer. The SiC degradation may have been related to the variations

  11. Spalling behavior and residual resistance of fibre reinforced Ultra-High performance concrete after exposure to high temperatures

    Directory of Open Access Journals (Sweden)

    Xiong, Ming-Xiang

    2015-12-01

    Full Text Available Experimental results of spalling and residual mechanical properties of ultra-high performance concrete after exposure to high temperatures are presented in this paper. The compressive strength of the ultra-high performance concrete ranged from 160 MPa~185 MPa. This study aimed to discover the effective way to prevent spalling for the ultra-high performance concrete and gauge its mechanical properties after it was subjected to fire. The effects of fiber type, fiber dosage, heating rate and curing condition were investigated. Test results showed that the compressive strength and elastic modulus of the ultra-high performance concrete declined slower than those of normal strength concrete after elevated temperatures. Polypropylene fiber rather than steel fiber was found effective to prevent spalling but affected workability. The effective fiber type and dosage were recommended to prevent spalling and ensure sufficient workability for casting and pumping of the ultra-high performance concrete.En este trabajo se presentan los resultados más relevantes del trabajo experimental realizado para valorar la laminación y las propiedades mecánicas residuales de hormigón de ultra-altas prestaciones tras su exposición a altas temperaturas. La resistencia a la compresión del hormigón de ultra-altas prestaciones osciló entre 160 MPa~185 MPa. El objetivo de este estudio fue descubrir una manera eficaz de prevenir desprendimientos y/o laminaciones en este hormigón y medir sus propiedades mecánicas después de ser sometido al fuego. Las variables estudiadas fueron la presencia y dosificación de fibras, velocidad de calentamiento y condiciones de curado. Los resultados mostraron, tras la exposición a altas temperaturas, que la resistencia a compresión y el módulo de elasticidad del hormigón de ultra-altas prestaciones disminuían más lento que las de un hormigón con resistencia normal. La fibra de polipropileno resultó más eficaz para prevenir

  12. High temperature oxidation behavior of aluminide on a Ni-based single crystal superalloy in different surface orientations

    Institute of Scientific and Technical Information of China (English)

    Fahamsyah H.Latief; Koji Kakehi; El-Sayed M.Sherif

    2014-01-01

    An investigation on oxidation behavior of coated Ni-based single crystal superalloy in different surface orientations has been carried out at 1100 1C. It has been found that the {100} surface shows a better oxidation resistance than the {110} one, which is attributed that the {110}surface had a slightly higher oxidation rate when compared to the {100} surface. The experimental results also indicated that the anisotropic oxidation behavior took place even with a very small difference in the oxidation rates that was found between the two surfaces. The differences of the topologically close packed phase amount and its penetration depth between the two surfaces, including the ratio of α-Al2O3 after 500 h oxidation, were responsible for the oxidation anisotropy.

  13. THE BEHAVIOR OF SOLUBLE METALS ELUTED FROM Ni/Fe-BASED ALLOY REACTORS AFTER HIGH-TEMPERATURE AND HIGH-PRESSURE WATER PROCESS

    Directory of Open Access Journals (Sweden)

    M. Faisal

    2012-05-01

    Full Text Available The behavior of heavy metals eluted from the wall of Ni/Fe-based alloy reactors after high-temperature and high-pressure water reaction were studied at temperatures ranging from 250 to 400oC. For this purpose, water and cysteic acid were heated in two reactor materials which are SUS 316 and Inconel 625. Under the tested conditions, the erratic behaviors of soluble metals eluted from the wall of Ni/Fe-based alloy in high temperature water were observed. Results showed that metals could be eluted even at a short contact time. The presence of air also promotes elution at sub-critical conditions. At sub-critical conditions, a significant amount of Cr was extracted from SUS 316, while only traces of Ni, Fe, Mo and Mn were eluted. In contrast, Ni was removed in significant amounts compared to Cr when Inconel 625 was tested. It was observed that eluted metals tend to increased under acidic conditions and most of those metals were over the limit of WHO guideline for drinking water. The results are significant both on the viewpoint of environmental regulation on disposal of wastes containing heavy metals, toxicity of resulting product and catalytic effect on a particular reaction.

  14. Law of mixture used to model the flow behavior of a duplex stainless steel at high temperatures

    International Nuclear Information System (INIS)

    Momeni, A.; Dehghani, K.; Poletti, M.C.

    2013-01-01

    In this investigation the flow curves of a duplex stainless steel were drawn by performing hot compression tests over a wide temperature range of 950–1200 °C and strain rates of 0.001–100 s −1 . The flow curves of ferrite and austenite phases in the duplex structure were depicted by conducting similar hot compression tests on two steels that were cast and prepared with the same chemical compositions. The flow curves of the austenitic steel were found typical of dynamic recrystallization. They were successfully modeled by using the experimental exponential equation proposed by Cingara and McQueen. The flow curves of the ferritic steel were typical of dynamic recovery. They were modeled by the dislocation density evolution function proposed by Estrin and Meckning. Comparing the flow curves of three studied steels, it was found that the flow curves of the duplex steel were very similar and close to those of the ferrite steel. It was understood that in a duplex structure of ferrite and austenite the flow behavior is mostly controlled by the softer phase, i.e. ferrite. The law of mixture was modified to consider the strain partitioning between ferrite and austenite. The distribution coefficients of ferrite and austenite were described and determined at different deformation conditions. The results of modeling satisfactorily predicted the experimental curves. It was shown that the influence of austenite on the flow behavior of the duplex structure is almost low. However, it increases as strain rate or temperature rises. - Highlights: ► Flow curves of austenite and ferrite in the duplex steel were modeled separately. ► The flow behavior of the duplex steel is mostly controlled by ferrite. ► The effect of austenite on flow curve increases with temperature and strain rate. ► The flow curve of the duplex steel is modeled by the modified law of mixture

  15. Numerical simulation of vapor film collapse behavior on high-temperature droplet surface with three-dimensional lattice gas cellular automata

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Abe, Yutaka; Matsukuma, Yosuke

    2008-01-01

    It is pointed out that a vapor film on a premixed high-temperature droplet surface is needed to be collapsed to trigger vapor explosion. Thus, it is important to clarify the micromechanism of vapor film collapse behavior for the occurrence of vapor explosion. In a previous study, it is suggested experimentally that vapor film collapse behavior is dominated by phase change phenomena rather than by the surrounding fluid motion. In the present study, vapor film collapse behavior is investigated to clarify the dominant factor of vapor film collapse behavior with lattice gas automata of three-dimensional immiscible lattice gas model (3-D ILG model). First, in order to represent the boiling and phase change phenomena, the thermal model of a heat wall model and a phase change model is newly constructed. Next, the numerical simulation of vapor film collapse behavior is performed with and without the phase change effect. As a result, the computational result with the phase change effect is observed to be almost same as the experimental result. It can be considered that vapor film collapse behavior is dominated by phase change phenomena. (author)

  16. High Temperature Deformation Behavior and Microstructure Evolution of Ti-4Al-4Fe-0.25Si Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Woo; Lee, Yongmoon; Lee, Chong Soo [Pohang University of Science and Technology, Pohang (Korea, Republic of); Yeom, Jong-Taek [Korea Institute of Materials Science, Changwon (Korea, Republic of); Lee, Gi Yeong [KPCM Incorporated, Gyeongsan (Korea, Republic of)

    2016-05-15

    Hot deformation behavior of Ti-4Al-4Fe-0.25Si alloy with martensite microstructure was investigated by compression tests at temperatures of 1023 – 1173 K (α+β phase region) and strain rates of 10{sup -3} – 1 s{sup -1}. By analyzing the deformation behavior, plastic deformation instability parameters including strain rate sensitivity, deformation temperature sensitivity, efficiency of power dissipation, and Ziegler’s instability were evaluated as a function of deformation temperature and strain rate, and they were further examined by drawing deformation processing maps. The microstructure evolution was also studied to determine the deformation conditions under which equiaxed α phase was formed in the microstructure without remnants or kinked α phase platelets and shear bands, these last two of which cause severe cracks during post-forming process. Based on the combined results of the processing maps and the microstructure analysis, the optimum α+β forging conditions for Ti-4Al-4Fe-0.25Si alloy were determined.

  17. Investigation of the oxidation behavior of dispersion stabilized alloys when exposed to a dynamic high temperature environment

    Science.gov (United States)

    Tenney, D. R.

    1974-01-01

    The oxidation behavior of TD-NiCr and TD-NiCrAlY alloys have been studied at 2000 and 2200 F in static and high speed flowing air environments. The TD-NiCrAlY alloys preoxidized to produce an Al2O3 scale on the surface showed good oxidation resistance in both types of environments. The TD-NiCr alloy which had a Cr2O3 oxide scale after preoxidation was found to oxidize more than an order of magnitude faster under the dynamic test conditions than at comparable static test conditions. Although Cr2O3 normally provides good oxidation protection, it was rapidly lost due to formation of volatile CrO3 when exposed to the high speed air stream. The preferred oxide arrangement for the dynamic test consisted of an external layer of NiO with a porous mushroom type morphology, an intermediate duplex layer of NiO and Cr2O3, and a continuous inner layer of Cr2O3 in contact with the alloy substrate. An oxidation model has been developed to explain the observed microstructure and overall oxidation behavior of all alloys.

  18. High-temperature behavior of dicesium molybdate Cs{sub 2}MoO{sub 4}: Implications for fast neutron reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wallez, Gilles, E-mail: gilles.wallez@upmc.fr [Institut de Recherche de Chimie Paris, CNRS—Chimie ParisTech, 11 rue Pierre et Marie Curie, 75005 Paris (France); Université Pierre et Marie Curie, 4 place Jussieu, 75005 Paris (France); Raison, Philippe E., E-mail: philippe.raison@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, 76125 Karlsruhe (Germany); Smith, Anna L., E-mail: anna.smith@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, 76125 Karlsruhe (Germany); Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Clavier, Nicolas, E-mail: nicolas.clavier@cea.fr [ICSM–UMR5257 CNRS/CEA/UM2/ENSCM, Site de Marcoule, Bât 426, BP 17171, 30207 Bagnols/Cèze (France); Dacheux, Nicolas, E-mail: nicolas.dacheux@cea.fr [ICSM–UMR5257 CNRS/CEA/UM2/ENSCM, Site de Marcoule, Bât 426, BP 17171, 30207 Bagnols/Cèze (France)

    2014-07-01

    Dicesium molybdate (Cs{sub 2}MoO{sub 4})'s thermal expansion and crystal structure have been investigated herein by high temperature X ray diffraction in conjunction with Raman spectroscopy. This first crystal-chemical insight at high temperature is aimed at predicting the thermostructural and thermomechanical behavior of this oxide formed by the accumulation of Cs and Mo fission products at the periphery of nuclear fuel rods in sodium-cooled fast reactors. Within the temperature range of the fuel's rim, Cs{sub 2}MoO{sub 4} becomes hexagonal P6{sub 3}/mmc, with disordered MoO{sub 4} tetrahedra and 2D distribution of Cs–O bonds that makes thermal axial expansion both large (50≤α{sub l}≤70 10{sup −6} °C{sup −1}, 500–800 °C) and highly anisotropic (α{sub c}−α{sub a}=67×10{sup −6} °C{sup −1}, hexagonal form). The difference with the fuel's expansion coefficient is of potential concern with respect to the cohesion of the Cs{sub 2}MoO{sub 4} surface film and the possible release of cesium radionuclides in accidental situations. - Graphical abstract: The weakness of the Cs–O bonds and the disordering of the MoO{sub 4} tetrahedra array in the high-temperature form are responsible for the huge thermal expansion of Cs{sub 2}MoO{sub 4} along the c-axis. - Highlights: • Thermomechanical behavior of Cs{sub 2}MoO{sub 4} fission products compound is studied. • High-temperature form of Cs{sub 2}MoO{sub 4} is characterized by XRD and Raman. • Thermal expansion appears very high and anisotropic. • Cohesion between Cs{sub 2}MoO{sub 4} and nuclear fuel seems questionable, and Cs release is expected.

  19. Some trends in constitutive equation model development for high-temperature behavior of fast-reactor structural alloys

    International Nuclear Information System (INIS)

    Pugh, C.E.; Robinson, D.N.

    1977-01-01

    The paper addresses some important features of the inelastic behavior of 2 1 / 4 Cr--1Mo steel and indicates a mathematical framework that is capable of representing these types of response. While the constitutive model discussed embraces capabilities beyond those of equations presently used in design analyses; their implementation into practicable analysis methods (such as finite-element programs) is more demanding. For example, in the case of slow time-dependent deformations, the equations governing accumulation of the inelastic strain components and the evolution of the tensorial state variable α are intimately coupled. A part of recommending any such model for use in design must be a quantitative assessment of the economic feasibility of implementation

  20. Stress-strain behavior under static loading in Gd123 high-temperature superconductors at 77 K

    Science.gov (United States)

    Fujimoto, Hiroyuki; Murakami, Akira; Teshima, Hidekazu; Morita, Mitsuru

    2013-10-01

    Mechanical properties of melt-growth GdBa2Cu3Ox (Gd123) superconducting samples with 10 wt.% Ag2O and 0.5 wt.% Pt were evaluated at 77 K through flexural tests for specimens cut from the samples in order to estimate the mechanical properties of the Gd123 material without metal substrates, buffer layers or stabilization layers. We discuss the mechanical properties; the Young's modulus and flexural strength with stress-strain behavior at 77 K. The results show that the flexural strength and fracture strain of Gd123 at 77 K are approximately 100 MPa and 0.1%, respectively, and that the origin of the fracture is defects such as pores, impurities and non-superconducting compounds. We also show that the Young's modulus of Gd123 is estimated to be 160-165 GPa.

  1. Deformation behavior of laser welds in high temperature oxidation resistant Fe–Cr–Al alloys for fuel cladding applications

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G., E-mail: fieldkg@ornl.gov; Gussev, Maxim N., E-mail: gussevmn@ornl.gov; Yamamoto, Yukinori, E-mail: yamamotoy@ornl.gov; Snead, Lance L., E-mail: sneadll@ornl.gov

    2014-11-15

    Ferritic-structured Fe–Cr–Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe–(13–17.5)Cr–(3–4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  2. Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

    International Nuclear Information System (INIS)

    Hagen, S.

    1984-01-01

    Out-of-pile experiments have been performed to investigate the escalation in temperature of Zircaloy-clad fuel rods during heatup in steam due to the exothermal Zircaloy steam reaction. In these tests single Zircaloy/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring rods--were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 2200 0 C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten Zircaloy. The escalation process was influenced by the temperature behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

  3. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

    Science.gov (United States)

    Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  4. Microstructure and High Temperature Plastic Deformation Behavior of Al-12Si Based Alloy Fabricated by an Electromagnetic Casting and Stirring Process

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Kyung-Soo; Roh, Heung-Ryeol; Kim, Mok-Soon [Inha University, Incheon (Korea, Republic of); Kim, Jong-Ho; Park, Joon-Pyo [Research Institute of Industrial Science and Technology, Pohang (Korea, Republic of)

    2017-06-15

    An as-received EMC/S (electromagnetic casting and stirring)-processed Al-12Si based alloy billet was homogenized to examine its microstructure and high temperature plastic deformation behavior, using compressive tests over the temperature range from 623 to 743 K and a strain rate range from 1.0×10{sup -3} to 1.0×10{sup 0}s{sup -1}. The results were compared with samples processed by the direct chill casting (DC) method. The fraction of equiaxed structure for the as-received EMC/S billet(41%) was much higher than that of the as-received DC billet(6 %). All true stress – true strain curves acquired from the compressive tests exhibited a peak stress at the initial stage of plastic deformation. Flow stress showed a steady state region after the appearance of peak stress with increasing strain. The peak stress decreased with increasing temperature at a given strain rate and a decreasing strain rate at a given temperature. A constitutive equation was made for each alloy, which could be used to predict the peak stress. A recrystallized grain structure was observed in all the deformed specimens, indicating that dynamic recrystallization is the predominant mechanism during high temperature plastic deformation of both the homogenized EMC/S and DC-processed Al-12Si based alloys.

  5. The behavior of interstitials in irradiated graphite

    International Nuclear Information System (INIS)

    Pedraza, D.F.

    1991-01-01

    A computer model is developed to simulate the behavior of self-interstitials with particular attention to clustering. Owing to the layer structure of graphite, atomistic simulations can be performed using a large parallelepipedic supercell containing a few layers. In particular, interstitial clustering is studied here using a supercell that contains two basal planes only. Frenkel pairs are randomly produced. Interstitials are placed at sites between the crystal planes while vacancies are distributed in the two crystal planes. The size of the computational cell is 20000 atoms and periodic boundary conditions are used in two dimensions. Vacancies are assumed immobile whereas interstitials are given a certain mobility. Two point defect sinks are considered, direct recombination of Frenkel pairs and interstitial clusters. The clusters are assumed to be mobile up to a certain size where they are presumed to become loop nuclei. Clusters can shrink by emission of singly bonded interstitials or by recombination of a peripheral interstitial with a neighboring vacancy. The conditions under which interstitial clustering occurs are reported. It is shown that when clustering occurs the cluster size population gradually shifts towards the largest size cluster. The implications of the present results for irradiation growth and irradiation-induced amorphization are discussed

  6. The behaviour of transport from the fission products caesium and strontium in coated particles for high temperature reactors under irradiation conditions

    International Nuclear Information System (INIS)

    Zoller, P.

    1976-07-01

    At first survey is given about existing knowledge of the behaviour of caesium and strontium fission product transport in coated particles. In order to describe the complicated fission product transport mechanisms under irradiation conditions a suitable calculating model (SLIPPER) is taken over and modified to the special problems of an irradiation experiment. Fundamentally, the fission product transport is represented by the two contributions of diffusion and recoil, at which the diffusion is described by effective diffusion coefficients. In difference of that the possibility of a two-phase-diffusion is examined for the Cs diffusion in the fuel kernel. The model application on measuring results from irradiation experiments of KFA-Juelich and Mol-Belgien allowed the explanation from the characteristic of fission product transport in coated particles under irradiation conditions and produced effective diffusion coefficients for the fission products Cs and Sr. (orig.) [de

  7. Investigation on compression behavior of TZM and La{sub 2}O{sub 3} doped TZM Alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Ping, E-mail: huping1985@126.com [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Zhou, Yuhang; Chang, Tian; Yu, Zhitao; Wang, Kuaishe; Yang, Fan; Hu, Boliang [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Cao, Weicheng [Jinduicheng Molybdenum Co., Ltd, Xi’an 710077 (China); Yu, Hailiang [School of Mechanical, Materials, Mechatronic Engineering, University of Wollongong, Wollongong, NSW 2500 (Australia)

    2017-02-27

    Mechanical properties of Titanium-zirconium-molybdenum (TZM) and La{sub 2}O{sub 3} doped TZM alloys under compression were tested at 1000 °C and 1200 °C. Microstructure of TZM and La{sub 2}O{sub 3} doped TZM alloys after compressing was characterized by scanning electron microscopy. The effects on La{sub 2}O{sub 3} doping on the high temperature deformation behavior and microstructure evolution of the TZM alloy were analyzed. Results show that La{sub 2}O{sub 3} doping can refine the grain size of TZM alloy. La{sub 2}O{sub 3} doping changes fracture model of TZM alloy. TZM alloy exhibits mainly intergranular fracture, while the La{sub 2}O{sub 3} doped TZM alloy exhibits both intergranular and transgranular fracture mode.

  8. Development of the novel ferrous-based stainless steel for biomedical applications, part I: high-temperature microstructure, mechanical properties and damping behavior.

    Science.gov (United States)

    Wu, Ching-Zong; Chen, Shih-Chung; Shih, Yung-Hsun; Hung, Jing-Ming; Lin, Chia-Cheng; Lin, Li-Hsiang; Ou, Keng-Liang

    2011-10-01

    This research investigated the high-temperature microstructure, mechanical properties, and damping behavior of Fe-9 Al-30 Mn-1C-5 Co (wt.%) alloy by means of electron microscopy, experimental model analysis, and hardness and tensile testing. Subsequent microstructural transformation occurred when the alloy under consideration was subjected to heat treatment in the temperature range of 1000-1150 °C: γ → (γ+κ). The κ-phase carbides had an ordered L'1(2)-type structure with lattice parameter a = 0.385 nm. The maximum yield strength (σ(y)), hardness, elongation, and damping coefficient of this alloy are 645 MPa, Hv 292, ~54%, and 178.5 × 10(-4), respectively. These features could be useful in further understanding the relationship between the biocompatibility and the wear and corrosion resistance of the alloy, so as to allow the development of a promising biomedical material. Copyright © 2011 Elsevier Ltd. All rights reserved.

  9. High temperature battery. Hochtemperaturbatterie

    Energy Technology Data Exchange (ETDEWEB)

    Bulling, M.

    1992-06-04

    To prevent heat losses of a high temperature battery, it is proposed to make the incoming current leads in the area of their penetration through the double-walled insulating housing as thermal throttle, particularly spiral ones.

  10. Thermomechanics of composite structures under high temperatures

    CERN Document Server

    Dimitrienko, Yu I

    2016-01-01

    This pioneering book presents new models for the thermomechanical behavior of composite materials and structures taking into account internal physico-chemical transformations such as thermodecomposition, sublimation and melting at high temperatures (up to 3000 K). It is of great importance for the design of new thermostable materials and for the investigation of reliability and fire safety of composite structures. It also supports the investigation of interaction of composites with laser irradiation and the design of heat-shield systems. Structural methods are presented for calculating the effective mechanical and thermal properties of matrices, fibres and unidirectional, reinforced by dispersed particles and textile composites, in terms of properties of their constituent phases. Useful calculation methods are developed for characteristics such as the rate of thermomechanical erosion of composites under high-speed flow and the heat deformation of composites with account of chemical shrinkage. The author expan...

  11. High temperature refrigerator

    International Nuclear Information System (INIS)

    Steyert, W.A. Jr.

    1978-01-01

    A high temperature magnetic refrigerator is described which uses a Stirling-like cycle in which rotating magnetic working material is heated in zero field and adiabatically magnetized, cooled in high field, then adiabatically demagnetized. During this cycle the working material is in heat exchange with a pumped fluid which absorbs heat from a low temperature heat source and deposits heat in a high temperature reservoir. The magnetic refrigeration cycle operates at an efficiency 70% of Carnot

  12. High-temperature dynamic behavior in bulk liquid water: A molecular dynamics simulation study using the OPC and TIP4P-Ew potentials

    Science.gov (United States)

    Gabrieli, Andrea; Sant, Marco; Izadi, Saeed; Shabane, Parviz Seifpanahi; Onufriev, Alexey V.; Suffritti, Giuseppe B.

    2018-02-01

    Classical molecular dynamics simulations were performed to study the high-temperature (above 300 K) dynamic behavior of bulk water, specifically the behavior of the diffusion coefficient, hydrogen bond, and nearest-neighbor lifetimes. Two water potentials were compared: the recently proposed "globally optimal" point charge (OPC) model and the well-known TIP4P-Ew model. By considering the Arrhenius plots of the computed inverse diffusion coefficient and rotational relaxation constants, a crossover from Vogel-Fulcher-Tammann behavior to a linear trend with increasing temperature was detected at T* ≈ 309 and T* ≈ 285 K for the OPC and TIP4P-Ew models, respectively. Experimentally, the crossover point was previously observed at T* ± 315-5 K. We also verified that for the coefficient of thermal expansion α P ( T, P), the isobaric α P ( T) curves cross at about the same T* as in the experiment. The lifetimes of water hydrogen bonds and of the nearest neighbors were evaluated and were found to cross near T*, where the lifetimes are about 1 ps. For T T*, water behaves more like a simple liquid. The fact that T* falls within the biologically relevant temperature range is a strong motivation for further analysis of the phenomenon and its possible consequences for biomolecular systems.

  13. Irradiation behaviors of coated fuel particles, (3)

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kashimura, Satoru; Iwamoto, Kazumi; Ikawa, Katsuichi

    1980-07-01

    This report is concerning to the irradiation experiments of the coated fuel particles, which were performed by 72F-6A and 72F-7A capsules in JMTR. The coated particles referred to the preliminary design of VHTR were prepared for the experiments in 1972 and 1973. 72F-6A capsule was irradiated at G-10 hole of JMTR fuel zone for 2 reactor cycles, and 72F-7A capsule had been planned to be irradiated at the same irradiation hole before 72F-6A. However, due to slight leak of the gaseous fission products into the vacuum system controlling irradiation temperature, irradiation of 72F-7A capsule was ceased after 85 hrs since the beginning. In the post irradiation examination, inspection to surface appearance, ceramography, X-ray microradiography and acid leaching for the irradiated particle samples were made, and crushing strength of the two particle samples was measured. (author)

  14. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  15. High Temperature Oxidation Behavior of gamma-Ni+gamma'-Ni3Al Alloys and Coatings Modified with Pt and Reactive Elements

    Energy Technology Data Exchange (ETDEWEB)

    Mu, Nan [Iowa State Univ., Ames, IA (United States)

    2007-12-01

    Materials for high-pressure turbine blades must be able to operate in the high-temperature gases (above 1000 C) emerging from the combustion chamber. Accordingly, the development of nickel-based superalloys has been constantly motivated by the need to have improved engine efficiency, reliability and service lifetime under the harsh conditions imposed by the turbine environment. However, the melting point of nickel (1455 C) provides a natural ceiling for the temperature capability of nickel-based superalloys. Thus, surface-engineered turbine components with modified diffusion coatings and overlay coatings are used. Theses coatings are capable of forming a compact and adherent oxide scale, which greatly impedes the further transport of reactants between the high-temperature gases and the underlying metal and thus reducing attack by the atmosphere. Typically, these coatings contain β-NiAl as a principal constituent phase in order to have sufficient aluminum content to form an Al2O3 scale at elevated temperatures. The drawbacks to the currently-used {beta}-based coatings, such as phase instabilities, associated stresses induced by such phase instabilities, and extensive coating/substrate interdiffusion, are major motivations in this study to seek next-generation coatings. The high-temperature oxidation resistance of novel Pt + Hf-modified γ-Ni + γ-Ni3Al-based alloys and coatings were investigated in this study. Both early-stage and 4-days isothermal oxidation behavior of single-phase γ-Ni and γ'-Ni3Al alloys were assessed by examining the weight changes, oxide-scale structures, and elemental concentration profiles through the scales and subsurface alloy regions. It was found that Pt promotes Al2O3 formation by suppressing the NiO growth on both γ-Ni and γ'Ni3Al single-phase alloys. This effect increases with increasing Pt content. Moreover, Pt exhibits this effect even at

  16. The effect of molten salt on high temperature behavior of stainless steel and titanium alloy with the presence of water vapor

    Science.gov (United States)

    Baharum, Azila; Othman, Norinsan Kamil; Salleh, Emee Marina

    2018-04-01

    The high temperature oxidation experiment was conducted to study the behavior of titanium alloy Ti6A14V and stainless steel 316 in Na2SO4-50%NaCl + Ar-20%O2 (molten salt) and Na2SO4-50%NaCl + Ar-20%O2 + 12% H2O (molten salt + water vapor) environment at 900°C for 30 hours using horizontal tube furnace. The sample then was investigated using weight change measurement analysis and X-ray diffraction (XRD) analysis to study the weight gained and the phase oxidation that occurred. The weight gained of the titanium alloy was higher in molten salt environment compared to stainless steel due to the rapid growth in the oxide scale but showed almost no change of weight gained upon addition of water vapor. This is due to the alloy was fully oxidized. Stainless steel showed more protection and better effect in molten salt environment compared to mixed environment showed by slower weight gain and lower oxidation rate. Meanwhile, the phase oxidation test of the samples showed that the titanium alloy consist of multi oxide layer of rutile (TiO2) and Al2O3 on the surface of the exposed sample. While stainless steel show the formation of both protective Cr-rich oxide and non-protective Fe-rich oxide layer. This can be concluded that stainless steel is better compared to Ti alloy due to slow growing of chromia oxide. Therefore it is proven that stainless steel has better self-protection upon high temperature exposure.

  17. High-Temperature Tensile and Tribological Behavior of Hybrid (ZrB2+Al3Zr)/AA5052 In Situ Composite

    Science.gov (United States)

    Gautam, G.; Kumar, N.; Mohan, A.; Gautam, R. K.; Mohan, S.

    2016-09-01

    During service life, components such as piston, cylinder blocks, brakes, and discs/drums, have to work under high-temperature conditions. In order to have appropriate material for such applications high-temperature studies are important. Hybrid (ZrB2+Al3Zr)/AA5052 in situ composite has been investigated from ambient to 523 K (250 °C) at an interval of 50 deg. (ZrB2+Al3Zr)/AA5052 in situ composite has been fabricated by the direct melt reaction of AA5052 alloy with zirconium and boron salts. Microstructure studies show refinement in the grain size of base alloy on in situ formation of reinforcement particles. Al3Zr particles are observed in rectangular and polyhedron shapes. It is observed from the tensile studies that ultimate tensile strength, yield strength, and percentage elongation decrease with increase in test temperature. Similar kind of behavior is also observed for flow curve properties. The tensile results have also been correlated with fractography. Wear and friction results indicate that the wear rate increases with increase in normal load, whereas coefficient of friction shows decreasing trend. With increasing test temperature, wear rate exhibits a typical phenomenon. After an initial increase, wear rate follows a decreasing trend up to 423 K (150 °C), and finally a rapid increase is observed, whereas coefficient of friction increases continuously with increase in test temperature. The mechanisms responsible for the variation of wear and friction with different temperatures have been discussed in detail with the help of worn surfaces studies under scanning electron microscope (SEM) & 3D-profilometer and debris analysis by XRD.

  18. Burst annealing of high temperature GaAs solar cells

    Science.gov (United States)

    Brothers, P. R.; Horne, W. E.

    1991-01-01

    One of the major limitations of solar cells in space power systems is their vulnerability to radiation damage. One solution to this problem is to periodically heat the cells to anneal the radiation damage. Annealing was demonstrated with silicon cells. The obstacle to annealing of GaAs cells was their susceptibility to thermal damage at the temperatures required to completely anneal the radiation damage. GaAs cells with high temperature contacts and encapsulation were developed. The cells tested are designed for concentrator use at 30 suns AMO. The circular active area is 2.5 mm in diameter for an area of 0.05 sq cm. Typical one sun AMO efficiency of these cells is over 18 percent. The cells were demonstrated to be resistant to damage after thermal excursions in excess of 600 C. This high temperature tolerance should allow these cells to survive the annealing of radiation damage. A limited set of experiments were devised to investigate the feasibility of annealing these high temperature cells. The effect of repeated cycles of electron and proton irradiation was tested. The damage mechanisms were analyzed. Limitations in annealing recovery suggested improvements in cell design for more complete recovery. These preliminary experiments also indicate the need for further study to isolate damage mechanisms. The primary objective of the experiments was to demonstrate and quantify the annealing behavior of high temperature GaAs cells. Secondary objectives were to measure the radiation degradation and to determine the effect of repeated irradiation and anneal cycles.

  19. Burst annealing of high temperature GaAs solar cells

    International Nuclear Information System (INIS)

    Brothers, P.R.; Horne, W.E.

    1991-01-01

    One of the major limitations of solar cells in space power systems is their vulnerability to radiation damage. One solution to this problem is to periodically heat the cells to anneal the radiation damage. Annealing was demonstrated with silicon cells. The obstacle to annealing of GaAs cells was their susceptibility to thermal damage at the temperatures required to completely anneal the radiation damage. GaAs cells with high temperature contacts and encapsulation were developed. The cells tested are designed for concentrator use at 30 suns AMO. The circular active area is 2.5 mm in diameter for an area of 0.05 sq cm. Typical one sun AMO efficiency of these cells is over 18 percent. The cells were demonstrated to be resistant to damage after thermal excursions in excess of 600 degree C. This high temperature tolerance should allow these cells to survive the annealing of radiation damage. A limited set of experiments were devised to investigate the feasibility of annealing these high temperature cells. The effect of repeated cycles of electron and proton irradiation was tested. The damage mechanisms were analyzed. Limitations in annealing recovery suggested improvements in cell design for more complete recovery. These preliminary experiments also indicate the need for further study to isolate damage mechanisms. The primary objective of the experiments was to demonstrate and quantify the annealing behavior of high temperature GaAs cells. Secondary objectives were to measure the radiation degradation and to determine the effect of repeated irradiation and anneal cycles

  20. Some behavioral aspects of adult rats irradiated prenatally

    International Nuclear Information System (INIS)

    Vekovishcheva, O.Yu.; Blagova, O.E.; Borovitskaya, A.E.; Evtushenko, V.I.; Khanson, K.P.

    1992-01-01

    This is a study of the effects of prenatal irradiation on the behavior of rats. The experiments were performed on 42 eighteen month old rats of both sexes. Eight of the males and thirteen females had been irradiated prenatally. The results of this experiment indicated that in general, the activation of behavior, the appearance of aggression and the increase in chaos along with the presence of behavior poses were typical of the suppressed condition of the prenatal irradiated animal. Also, among prenatally irradiated animals, there was a greater degree of anxiety, a slow rate of adjustment to unfamiliar situations and unfriendly relationships between animals of the same sex. These results were compared with the results of behavioral experiments on irradiated adult rats

  1. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

  2. Research on high-temperature compression and creep behavior of porous Cu–Ni–Cr alloy for molten carbonate fuel cell anodes

    Directory of Open Access Journals (Sweden)

    Li W.

    2015-06-01

    Full Text Available The effect of porosity on high temperature compression and creep behavior of porous Cu alloy for the new molten carbonate fuel cell anodes was examined. Optical microscopy and scanning electron microscopy were used to investigate and analyze the details of the microstructure and surface deformation. Compression creep tests were utilized to evaluate the mechanical properties of the alloy at 650 °C. The compression strength, elastic modulus, and yield stress all increased with the decrease in porosity. Under the same creep stress, the materials with higher porosity exhibited inferior creep resistance and higher steadystate creep rate. The creep behavior has been classified in terms of two stages. The first stage relates to grain rearrangement which results from the destruction of large pores by the applied load. In the second stage, small pores are collapsed by a subsequent sintering process under the load. The main deformation mechanism consists in that several deformation bands generate sequentially under the perpendicular loading, and in these deformation bands the pores are deformed by flattering and collapsing sequentially. On the other hand, the shape of a pore has a severe influence on the creep resistance of the material, i.e. every increase of pore size corresponds to a decrease in creep resistance.

  3. Modeling the Effects of Cu Content and Deformation Variables on the High-Temperature Flow Behavior of Dilute Al-Fe-Si Alloys Using an Artificial Neural Network.

    Science.gov (United States)

    Shakiba, Mohammad; Parson, Nick; Chen, X-Grant

    2016-06-30

    The hot deformation behavior of Al-0.12Fe-0.1Si alloys with varied amounts of Cu (0.002-0.31 wt %) was investigated by uniaxial compression tests conducted at different temperatures (400 °C-550 °C) and strain rates (0.01-10 s -1 ). The results demonstrated that flow stress decreased with increasing deformation temperature and decreasing strain rate, while flow stress increased with increasing Cu content for all deformation conditions studied due to the solute drag effect. Based on the experimental data, an artificial neural network (ANN) model was developed to study the relationship between chemical composition, deformation variables and high-temperature flow behavior. A three-layer feed-forward back-propagation artificial neural network with 20 neurons in a hidden layer was established in this study. The input parameters were Cu content, temperature, strain rate and strain, while the flow stress was the output. The performance of the proposed model was evaluated using the K-fold cross-validation method. The results showed excellent generalization capability of the developed model. Sensitivity analysis indicated that the strain rate is the most important parameter, while the Cu content exhibited a modest but significant influence on the flow stress.

  4. Flow behavior and microstructures of powder metallurgical CrFeCoNiMo0.2 high entropy alloy during high temperature deformation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jiawen [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Liu, Yong, E-mail: yonliu@csu.edu.cn [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Liu, Bin, E-mail: binliu@csu.edu.cn [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Wang, Yan [School of Aeronautics and Astronautics, Central South University, Changsha 410083 (China); Cao, Yuankui; Li, Tianchen; Zhou, Rui [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China)

    2017-03-24

    Dynamic recrystallization (DRX) refine grains of high entropy alloys (HEAs) and significant improve the mechanical property of HEAs, but the effect of high melting point element molybdenum (Mo) on high temperature deformation behavior has not been fully understood. In the present study, flow behavior and microstructures of powder metallurgical CrFeCoNiMo{sub 0.2} HEA were investigated by hot compression tests performed at temperatures ranging from 700 to 1100 °C with strain rates from 10{sup −3} to 1 s{sup −1}. The Arrhenius constitutive equation with strain-dependent material constants was used for modeling and prediction of flow stress. It was found that at 700 °C, the dynamic recovery is the dominant softening mechanism, whilst with the increase in compression testing temperature, the DRX becomes the dominant mechanism of softening. In the present HEA, the addition of Mo results in the high activation energy (463 kJ mol{sup −1}) and the phase separation during hot deformation. The formation of Mo-rich σ phase particles pins grain boundary migration during DRX, and therefore refines the size of recrystallized grains.

  5. Influences of Cr content and PWHT on microstructure and oxidation behavior of stainless steel weld overlay cladding materials in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Cao, X.Y.; Ding, X.F. [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Lu, Y.H., E-mail: lu_yonghao@mater.ustb.edu.cn [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Zhu, P. [Suzhou Nuclear Power Research Institute Co. Ltd., 1788 Xihuan Road, 215004 Suzhou (China); Shoji, T. [National Center for Materials Service Safety, University of Science and Technology Beijing, 30 Xueyuan Road, 100083 Beijing (China); Fracture and Reliability Research Institute, Tohoku University, 6-6-01 Aramaki Aoba, Aoba-ku, Sendai City 980-8579 (Japan)

    2015-12-15

    Influences of Cr content and post weld heat treatment (PWHT) on microstructure and oxidation behavior of stainless steel cladding materials in high temperature water were investigated. The amounts of metal oxidized and dissolved were estimated to compare the oxidation behaviors of cladding materials with different Cr contents and PWHT. The results indicated that higher Cr content led to formation of more ferrite content, and carbides were found along δ/γ phase interface after PWHT. Higher Cr content enhanced the pitting resistance and compactness of the oxide film to reduce metal amount oxidized and dissolved, which mitigated the weight changes and the formation of Fe-rich oxides. PWHT promoted more and deeper pitting holes along the δ/γ phase interface due to formation of carbides, which resulted in an increase in metal amount oxidized and dissolved, and were also responsible for more Fe-rich oxides and higher weight changes. - Highlights: • The amounts of metal oxidized and metal dissolved were estimated. • Higher Cr content increased ferrite content and PWHT led to formation of carbides. • PWHT promoted more and deeper pitting holes along the δ/γ phase interface. • Lower Cr content and PWHT promoted the metal amounts oxidized and dissolved. • Lower Cr content and PWHT increased weight changes and Fe-rich film formation.

  6. Tritium and helium behavior in irradiated beryllium

    International Nuclear Information System (INIS)

    Billone, M.C.; Lin, C.C.; Baldwin, D.L.

    1990-11-01

    Large quantities of Be (> 100 metric tons) are planned for use in the ITER blanket design to enhance tritium breeding and to act as a thermal barrier between coolant and breeder. Tritium retention/release and He-induced swelling are important issues in blanket design. The data base on tritium and helium behavior in Be is reviewed. New data on tritium retention/release and He bubble growth are presented for Be irradiated to 5 x 10 22 n(E > 1 MeV)/cm 2 at ∼75 degree C and postirradiation-annealed for 700 hours at 500 degree C. A model (diffusion/desorption) is proposed and tested against the data base to determine tritium diffusivity and the desorption rate constant. Similarly a model for He-induced swelling is developed and tested against the data base. The dependence of tritium retention and release on He content and impurities (e.g. BeO) is also explored. 11 refs., 6 figs

  7. High temperature niobium alloys

    International Nuclear Information System (INIS)

    Wojcik, C.C.

    1991-01-01

    Niobium alloys are currently being used in various high temperature applications such as rocket propulsion, turbine engines and lighting systems. This paper presents an overview of the various commercial niobium alloys, including basic manufacturing processes, properties and applications. Current activities for new applications include powder metallurgy, coating development and fabrication of advanced porous structures for lithium cooled heat pipes

  8. High temperature storage loop :

    Energy Technology Data Exchange (ETDEWEB)

    Gill, David Dennis; Kolb, William J.

    2013-07-01

    A three year plan for thermal energy storage (TES) research was created at Sandia National Laboratories in the spring of 2012. This plan included a strategic goal of providing test capability for Sandia and for the nation in which to evaluate high temperature storage (>650ÀC) technology. The plan was to scope, design, and build a flow loop that would be compatible with a multitude of high temperature heat transfer/storage fluids. The High Temperature Storage Loop (HTSL) would be reconfigurable so that it was useful for not only storage testing, but also for high temperature receiver testing and high efficiency power cycle testing as well. In that way, HTSL was part of a much larger strategy for Sandia to provide a research and testing platform that would be integral for the evaluation of individual technologies funded under the SunShot program. DOEs SunShot program seeks to reduce the price of solar technologies to 6/kWhr to be cost competitive with carbon-based fuels. The HTSL project sought to provide evaluation capability for these SunShot supported technologies. This report includes the scoping, design, and budgetary costing aspects of this effort

  9. High Temperature Electrolysis

    DEFF Research Database (Denmark)

    Elder, Rachael; Cumming, Denis; Mogensen, Mogens Bjerg

    2015-01-01

    High temperature electrolysis of carbon dioxide, or co-electrolysis of carbon dioxide and steam, has a great potential for carbon dioxide utilisation. A solid oxide electrolysis cell (SOEC), operating between 500 and 900. °C, is used to reduce carbon dioxide to carbon monoxide. If steam is also i...

  10. Behavior of optical fibers under heavy irradiation

    International Nuclear Information System (INIS)

    Kakuta, T.; Sagawa, T.

    1998-01-01

    Several kinds of optical diagnostics are planned in a fusion reactor. Complicated optical systems such as periscopes are thought to be primary candidates for optical measurements, especially for visible wavelengths. However, optical fibers have several advantages over such optical systems. Also, the optical fibers could be a far better transmission line for signals under a high electromagnetic field. However, they have been considered vulnerable to heavy irradiation. In this study, several kinds of optical fibers were irradiated in the JMTR fission reactor. The optical transmissivity in fibers was measured in situ during fast neutron and gamma irradiation, up to doses of 2 x 10 24 n m -2 and 5 x 10 9 Gy, respectively. The irradiation temperature ranged from 300 to 700 K. For pure ionizing irradiation environments, some methods for improving the radiation resistance of optical fibers were indicated. The results showed that effects of the irradiation associated with fast neutrons would be different from the effects of pure ionizing irradiation. Some fibers were found to withstand the heavy irradiation, especially in an infrared wavelength range. (orig.)

  11. High-temperature deformation behavior and mechanical properties of rapidly solidified Al-Li-Co and Al-Li-Zr alloys

    International Nuclear Information System (INIS)

    Sastry, S.M.L.; Oneal, J.E.

    1984-01-01

    The deformation behavior at 25-300 C of rapidly solidified Al-3Li-0.6Co and Al-3Li-0.3Zr alloys was studied by tensile property measurements and transmission electron microscopic examination of dislocation substructures. In binary Al-3Li and Al-3Li-Co alloys, the modulus normalized yield stress increases with an increase in temperature up to 150 C and then decreases. The yield stress at 25 C of Al-3Li-0.3Zr alloys is 180-200 MPa higher than that of Al-3Li alloys. However, the yield stress of the Zr-containing alloy decreases drastically with increasing temperatures above 75 C. The short-term yield stresses at 100-200 C of the Al-3Li-based alloys are higher than that of the conventional high-temperature Al alloys. The temperature dependences of the flow stresses of the alloys were analyzed in terms of the magnitudes and temperature dependences of the various strengthening contributions in the two alloys. The dislocation substructures at 25-300 C were correlated with mechanical properties. 19 references

  12. High temperature mechanical behavior of tube stackings – Part I: Microstructural and mechanical characterization of Inconel® 600 constitutive material

    Energy Technology Data Exchange (ETDEWEB)

    Marcadon, V., E-mail: Vincent.Marcadon@onera.fr [Onera – The French Aerospace Lab, F-92322 Châtillon (France); Davoine, C.; Lévêque, D.; Rafray, A.; Popoff, F.; Horezan, N.; Boivin, D. [Onera – The French Aerospace Lab, F-92322 Châtillon (France)

    2016-11-20

    This paper is the first part of a set of two papers dedicated to the mechanical behavior of cellular materials at high temperatures. For that purpose, cellular materials made of brazed tube stacking cores have been considered here. This paper addresses the characterization of the elasto-viscoplastic properties of the constitutive material of the tubes, Inconel®600, by means of tensile tests. Various temperatures and strain rates were investigated, from room temperature to 800 °C, in order to study the influence of both the brazing heat treatment and the test temperature on the mechanical properties of Inconel®600. Whereas the heat treatment drastically decreases the strength of the tubes, a significant viscous effect is revealed at 800 °C. Electron backscattered diffraction analyses carried out post-mortem on samples showed that both dynamic recrystallization and recovery occurred during tensile tests performed at 800 °C, especially at lower strain rates. In contrast, a highly deformed and textured microstructure was observed for the tubes loaded at lower temperatures.

  13. High-Temperature Electrical Insulation Behavior of Alumina Films Prepared at Room Temperature by Aerosol Deposition and Influence of Annealing Process and Powder Impurities

    Science.gov (United States)

    Schubert, Michael; Leupold, Nico; Exner, Jörg; Kita, Jaroslaw; Moos, Ralf

    2018-04-01

    Alumina (Al2O3) is a widely used material for highly insulating films due to its very low electrical conductivity, even at high temperatures. Typically, alumina films have to be sintered far above 1200 °C, which precludes the coating of lower melting substrates. The aerosol deposition method (ADM), however, is a promising method to manufacture ceramic films at room temperature directly from the ceramic raw powder. In this work, alumina films were deposited by ADM on a three-electrode setup with guard ring and the electrical conductivity was measured between 400 and 900 °C by direct current measurements according to ASTM D257 or IEC 60093. The effects of film annealing and of zirconia impurities in the powder on the electrical conductivity were investigated. The conductivity values of the ADM films correlate well with literature data and can even be improved by annealing at 900 °C from 4.5 × 10-12 S/cm before annealing up to 5.6 × 10-13 S/cm after annealing (measured at 400 °C). The influence of zirconia impurities is very low as the conductivity is only slightly elevated. The ADM-processed films show a very good insulation behavior represented by an even lower electrical conductivity than conventional alumina substrates as they are commercially available for thick-film technology.

  14. Investigation on the oxidation behavior of AlCrVxN thin films by means of synchrotron radiation and influence on the high temperature friction

    Science.gov (United States)

    Tillmann, Wolfgang; Kokalj, David; Stangier, Dominic; Paulus, Michael; Sternemann, Christian; Tolan, Metin

    2018-01-01

    Friction minimization is an important topic which is pursued in research and industry. In addition to the use of lubricants, friction-reducing oxide phases can be utilized which occur during. These oxides are called Magnéli phases and especially vanadium oxides exhibit good friction reducing properties. Thereby, the lubrication effect can be traced back to oxygen deficiencies. AlCrN thin films are being used as coatings for tools which have to withstand high temperatures. A further improvement of AlCrN thin films concerning their friction properties is possible by incorporation of vanadium. This study analyzes the temperature dependent oxidation behavior of magnetron sputtered AlCrVN thin films with different vanadium contents up to 13.5 at.-% by means of X-ray diffraction and X-ray absorption near-edge spectroscopy. Up to 400 °C the coatings show no oxidation. A higher temperature of 700 °C leads to an oxidation and formation of Magnéli phases of the coatings with vanadium contents above 10.7 at.-%. Friction coefficients, measured by ball-on-disk test are correlated with the oxide formation in order to figure out the effect of vanadium oxides. At 700 °C a decrease of the friction coefficient with increasing vanadium content can be observed, due to the formation of VO2, V2O3 and the Magnéli phase V4O7.

  15. High temperature oxidation-sulfidation behavior of Cr-Al2O3 and Nb-Al2O3 composites densified by spark plasma sintering

    International Nuclear Information System (INIS)

    Saucedo-Acuna, R.A.; Monreal-Romero, H.; Martinez-Villafane, A.; Chacon-Nava, J.G.; Arce-Colunga, U.; Gaona-Tiburcio, C.; De la Torre, S.D.

    2007-01-01

    The high temperature oxidation-sulfidation behavior of Cr-Al 2 O 3 and Nb-Al 2 O 3 composites prepared by mechanical alloying (MA) and spark plasma sintering (SPS) has been studied. These composite powders have a particular metal-ceramic interpenetrating network and excellent mechanical properties. Oxidation-sulfidation tests were carried out at 900 deg. C, in a 2.5%SO 2 + 3.6%O 2 + N 2 (balance) atmosphere for 48 h. The results revealed the influence of the sintering conditions on the specimens corrosion resistance, i.e. the Cr-Al 2 O 3 and Nb-Al 2 O 3 composite sintered at 1310 deg. C/4 min showed better corrosion resistance (lower weight gains) compared with those found for the 1440 deg. C/5 min conditions. For the former composite, a protective Cr 2 O 3 layer immediately forms upon heating, whereas for the later pest disintegration was noted. Thus, under the same sintering conditions the Nb-Al 2 O 3 composites showed the highest weight gains. The oxidation products were investigated by X-ray diffraction, scanning electron microscopy, and transmission electron microscopy

  16. Time-dependent high-temperature low-cycle fatigue behavior of nickel-base heat-resistant alloys for HTGR

    International Nuclear Information System (INIS)

    Tsuji, Hirokazu; Kondo, Tatsuo

    1988-06-01

    A series of strain controlled low-cycle fatigue tests at 900 deg C in the simulated HTGR helium environment were conducted on Hastelloy X and its modified version, Hastelloy XR in order to examine time-dependent high-temperature low-cycle fatigue behavior. In the tests with the symmetric triangular strain waveform, decreasing the strain rate led to notable reductions in the fatigue life. In the tests with the trapezoidal strain waveform with different holding types, the fatigue life was found to be reduced most effectively in tensile hold-time experiments. Based on the observations of the crack morphology the strain holding in the compressive side was suggested to play the role of suppressing the initiation and the growth of internal cracks or cavities, and to cause crack branching. When the frequency modified fatigue life method and/or the prediction of life by use of the ductility were applied, both the data obtained with the symmetric triangular strain waveform and those with the tensile hold-time experiments lay on the straight line plots. The data, however, obtained with the compressive and/or both hold-time experiments could not be handled satisfactorily by those methods. When the cumulative damage rule was applied, it was found that the reliability of HTGR components was ensured by limiting the creep-fatigue damage fraction within the value of 1. (author)

  17. 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

  18. Transformation behavior and shape memory properties of Ti50Ni15Pd25Cu10 high temperature shape memory alloy at various aging temperatures

    International Nuclear Information System (INIS)

    Rehman, Saif ur; Khan, Mushtaq; Nusair Khan, A.; Ali, Liaqat; Zaman, Sabah; Waseem, Muhammad; Ali, Liaqat; Jaffery, Syed Husain Imran

    2014-01-01

    This research presents an insight into the effect of various aging temperatures on the microstructure, hardness, phase transformation behavior and shape memory properties of Ti 50 Ni 15 Pd 25 Cu 10 high temperature shape memory alloy. The aging temperature was varied from 350 °C to 750 °C, whereas the shape memory properties were evaluated at 100–500 MPa. It was observed that the mentioned properties were strongly dependent on the aging temperatures. Based on the results obtained from scanning electron microscopy, X-ray diffractometry, microhardness testing, differential scanning calorimetry and thermomechanical testing, the aging temperatures can be divided into three ranges. At low aging temperatures (350 °C and below), the properties of the alloy remained the same as were found for solution treated sample, however at intermediate aging temperatures (400–600 °C) the properties of the alloy were changed significantly. Due to the formation of precipitates, the hardness was increased, whereas the phase transformation temperatures and work output were decreased considerably. The recovery ratio was found to be improved for intermediate aging temperatures. At high aging temperatures (650 °C and above), the hardness was decreased and the phase transformation temperatures were increased. Phase transformation temperature at the aging temperature of 750 °C was found to be increased significantly as compared to solution treated sample

  19. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-15

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

  20. Evaluation of Ion Irradiation Behavior of ODS Alloys

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  1. Effects of helium on ductile brittle transition behavior of reduced activation ferritic steels after high concentration he implantation at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, A.; Ejiri, M.; Nogami, S.; Ishiga, M.; Abe, K. [Tohoku Univ., Dept. of Quantum Science and Energy Engr, Sendai (Japan); Kasada, R.; Kimura, A. [Kyoto Univ., Institute of Advanced Energy (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

    2007-07-01

    Full text of publication follows: Influence of Helium (He) on fracture behavior of reduced activation ferritic/martensitic steels including Oxide Dispersion Strengthening (ODS) steels and F82H were examined. To study the He effects on fracture behavior of these steels after He bubble formation conditions, higher concentration of He implantation at around 550 C were performed and examined the relationship between microstructure evolution and fracture behavior of the steels. The 1.5CVN mini size Charpy specimens were used to evaluate impact test behavior. Reduced activation ferritic ODS steels, 9Cr-ODS and 12Cr-ODS steels were examine. F82H was also examined as reference material. Helium implantation was performed by a cyclotron of Tohoku University with a beam of 50 MeV {alpha}-particles at temperature around 550 C. A tandem-type energy degrader system was used to implant He into the specimen from the irradiated surface to the range of 50 MeV {alpha}-particles, that was about 380 {mu}m in iron. Implanted He concentration were about 1000 appm. Charpy impact test was performed using a instrumented impact test apparatus in Oarai branch of IMR, Tohoku University. Analyses of absorbed energy change and fracture surface were carried out. Vickers hardness test was also carried out on He implanted area of the 1.5CVN specimen to estimate irradiation hardening. Microstructural observation was performed by TEM. In the case of F82H, DBTT increased by the 1000 appm He implantation condition was about 80 C and grain boundary fracture surface was only observed in the He implanted area of all the ruptured specimens in brittle manner. On the other hand, DBTT shift and fracture mode change of He implanted 9Cr-ODS steel was not observed after He implantation. Microstructural observation showed that He bubble formation on the lath boundaries and grain boundaries were significant in F82H, but the bubble segregation on grain boundary in ODS steel was not apparent. The bubble formation

  2. Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Shiba, K.; Tanigawa, H.; Ando, M. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge National Laboratory, TN (United States); Stoller, R. [ORNL - Oak Ridge National Laboratory, Materials Science and Technology Div., Oak Ridge, AK TN (United States)

    2007-07-01

    Full text of publication follows: Post irradiation tensile tests on reduced activation ferritic/martensitic steel, F82H have been conducted over the past two decades using Japan Materials Testing Reactor (JMTR) of JAEA, and Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, USA, under Japan/US collaboration programs. According to these results, F82H does not demonstrate irradiation hardening above 673 K up to 60 dpa. The current study has been concentrated on hardening behavior at temperature around 573 K. A series of low temperature irradiation experiment has been conducted at the HFIR under the international collaborative research between JAEA/US-DOE. In this collaboration, the irradiation condition is precisely controlled by the well matured capsule designing and instrumentation. This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels compared with the irradiation properties database on F82H. Post irradiation tensile tests have been conducted on the F82H and its modified steels irradiated at 573 K and the dose level was up to 25 dpa. According to these results, irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated 0.2 % proof stress is less than 1 GPa at ambient temperature. The deterioration of total elongation was also saturated by 9 dpa irradiation. The ductility of some modified steels which showed larger total elongation than that of F82H before irradiation become the same level as that of standard F82H steel after irradiation, even though its magnitude of irradiation hardening is smaller than that of F82H. This suggests that the more ductile steel demonstrates the more ductility loss at this temperature, regardless to the hardening level. The difference in ductility loss behavior between various tensile specimens will be discussed as the ductility could depend on the specimen dimension. (authors)

  3. Mechanical properties of concrete for power reactor at high temperatures

    International Nuclear Information System (INIS)

    Kawase, Kiyotaka; Tanaka, Hitoshi; Nakano, Masayuki

    1985-01-01

    The purpose of this study is to investigate the mechanical properties of concrete for power reactor at high temperature. This paper presents the creep behavior of concrete at high temperature and the cause by which a specified aggregate is broken at a specified high temperature. The creep coefficient at high temperature is smaller than that at ordinary temperature. (author)

  4. High-temperature superconductivity

    International Nuclear Information System (INIS)

    Ginzburg, V.L.

    1987-07-01

    After a short account of the history of experimental studies on superconductivity, the microscopic theory of superconductivity, the calculation of the control temperature and its possible maximum value are presented. An explanation of the mechanism of superconductivity in recently discovered superconducting metal oxide ceramics and the perspectives for the realization of new high-temperature superconducting materials are discussed. 56 refs, 2 figs, 3 tabs

  5. Panel report on high temperature ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Nolet, T C [ed.

    1979-01-01

    Fundamental research is reported concerning high temperature ceramics for application in turbines, engines, batteries, gasifiers, MHD, fuel cells, heat exchangers, and hot wall combustors. Ceramics microstructure and behavior are included. (FS)

  6. Spectral and raw quasi in-situ energy dispersive X-ray data captured via a TEM analysis of an ODS austenitic stainless steel sample under 1 MeV Kr2+ high temperature irradiation.

    Science.gov (United States)

    Brooks, Adam J; Yao, Zhongwen

    2017-10-01

    The data presented in this article is related to the research experiment, titled: ' Quasi in-situ energy dispersive X-ray spectroscopy observation of matrix and solute interactions on Y-Ti-O oxide particles in an austenitic stainless steel under 1 MeV Kr 2+ high temperature irradiation' (Brooks et al., 2017) [1]. Quasi in-situ analysis during 1 MeV Kr 2+ 520 °C irradiation allowed the same microstructural area to be observed using a transmission electron microscope (TEM), on an oxide dispersion strengthened (ODS) austenitic stainless steel sample. The data presented contains two sets of energy dispersive X-ray spectroscopy (EDX) data collected before and after irradiation to 1.5 displacements-per-atom (~1.25×10 -3  dpa/s with 7.5×10 14  ions cm -2 ). The vendor software used to process and output the data is the Bruker Esprit v1.9 suite. The data includes the spectral (counts vs. keV energy) of the quasi in-situ scanned region (512×512 pixels at 56k magnification), along with the EDX scanning parameters. The.raw files from the Bruker Esprit v1.9 output are additionally included along with the.rpl data information files. Furthermore included are the two quasi in-situ HAADF images for visual comparison of the regions before and after irradiation. This in-situ experiment is deemed ' quasi' due to the thin foil irradiation taking place at an external TEM facility. We present this data for critical and/or extended analysis from the scientific community, with applications applying to: experimental data correlation, confirmation of results, and as computer based modeling inputs.

  7. High temperature corrosion behavior of different grain size specimens of 2.25 Cr-1 Mo steel in SO2+O2 environment

    International Nuclear Information System (INIS)

    Ghosh, D.; Mitra, S.K.

    2011-01-01

    The investigation is primarily aimed at the high temperature corrosion behavior of different grain sizes of 2.25 Cr-1 Mo steel at SO 2 +O 2 (mixed oxidation and sulfidation). The various grain sizes (18 μm,26 μm, 48 μm, and 72 μm) are obtained by different annealing treatment. Isothermal corrosion studies are carried out in different grain size specimens at 973K for 8 hours. The corrosion growth rate and the reaction kinetics are studied by weight gain method. The external scales of the post corroded specimen are studied in Scanning Electron Microscope (SEM) to examine the corrosion products morphology on the scale. X-ray mapping analysis of the different elements (Fe, O, Cr and S) is carried out by Energy Dispersive Spectroscopy (EDS) attached with SEM. The X-ray Diffraction Analysis (XRD) is also carried out to identify the corrosion products in the external scale. Finally, it is concluded that that the corrosion rate of 2.25 Cr-1 Mo steel strongly depend on grain sizes of the specimens. The corrosion rate increases with the decreases of grain size. The finer grain (18 μm) show higher corrosion rate than the coarse grains (72 μm). The weight gain kinetics follows the parabolic growth rate which further indicates that the corrosion process is diffusion controlled. The scale analysis shows the thicker scale and extensive scale cracking and spallations in case of finer grain size specimen (18 μm), whereas the coarse grain specimen (72 μm) shows compact and adherent layer. The XRD analysis shows that the corrosion products consist of mixtures of iron oxides( Fe 3 O 4 and Fe 2 O 3 ) and iron sulfides (FeS). The details mechanism of the corrosion is discussed to explain the difference in corrosion rate for different grain sizes. (author)

  8. First-principles study on oxidation effects in uranium oxides and high-pressure high-temperature behavior of point defects in uranium dioxide

    Science.gov (United States)

    Geng, Hua Y.; Song, Hong X.; Jin, K.; Xiang, S. K.; Wu, Q.

    2011-11-01

    Formation Gibbs free energy of point defects and oxygen clusters in uranium dioxide at high-pressure high-temperature conditions are calculated from first principles, using the LSDA+U approach for the electronic structure and the Debye model for the lattice vibrations. The phonon contribution on Frenkel pairs is found to be notable, whereas it is negligible for the Schottky defect. Hydrostatic compression changes the formation energies drastically, making defect concentrations depend more sensitively on pressure. Calculations show that, if no oxygen clusters are considered, uranium vacancy becomes predominant in overstoichiometric UO2 with the aid of the contribution from lattice vibrations, while compression favors oxygen defects and suppresses uranium vacancy greatly. At ambient pressure, however, the experimental observation of predominant oxygen defects in this regime can be reproduced only in a form of cuboctahedral clusters, underlining the importance of defect clustering in UO2+x. Making use of the point defect model, an equation of state for nonstoichiometric oxides is established, which is then applied to describe the shock Hugoniot of UO2+x. Furthermore, the oxidization and compression behavior of uranium monoxide, triuranium octoxide, uranium trioxide, and a series of defective UO2 at 0 K are investigated. The evolution of mechanical properties and electronic structures with an increase of the oxidation degree are analyzed, revealing the transition of the ground state of uranium oxides from metallic to Mott insulator and then to charge-transfer insulator due to the interplay of strongly correlated effects of 5f orbitals and the shift of electrons from uranium to oxygen atoms.

  9. Effect of Chromium on Corrosion Behavior of P110 Steels in CO2-H2S Environment with High Pressure and High Temperature

    Science.gov (United States)

    Sun, Jianbo; Sun, Chong; Lin, Xueqiang; Cheng, Xiangkun; Liu, Huifeng

    2016-01-01

    The novel Cr-containing low alloy steels have exhibited good corrosion resistance in CO2 environment, mainly owing to the formation of Cr-enriched corrosion film. In order to evaluate whether it is applicable to the CO2 and H2S coexistence conditions, the corrosion behavior of low-chromium steels in CO2-H2S environment with high pressure and high temperature was investigated using weight loss measurement and surface characterization. The results showed that P110 steel suffered localized corrosion and both 3Cr-P110 and 5Cr-P110 steels exhibited general corrosion. However, the corrosion rate of 5Cr-P110 was the highest among them. The corrosion process of the steels was simultaneously governed by CO2 and H2S. The outer scales on the three steels mainly consisted of FeS1−x crystals, whereas the inner scales on Cr-containing steels comprised of amorphous FeS1−x, Cr(OH)3 and FeCO3, in contrast with the amorphous FeS1−x and FeCO3 mixture film of P110 steel. The more chromium the steel contains, the more chromium compounds the corrosion products contain. The addition of chromium in steels increases the uniformity of the Cr-enriched corrosion scales, eliminates the localized corrosion, but cannot decrease the general corrosion rates. The formation of FeS1−x may interfere with Cr-enriched corrosion scales and lowering the corrosion performance of 3Cr-P110 and 5Cr-P110 steels. PMID:28773328

  10. Study of mechanical properties and high temperature oxidation behavior of a novel cold-spray Ni-20Cr coating on boiler steels

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Narinder [Semiconductor Materials and Device Laboratory, Department of Semiconductor Science, Dongguk University-Seoul, Seoul 100715 (Korea, Republic of); Kumar, Manoj [School of Mechanical, Materials & Energy Engineering, Indian Institute of Technology Ropar, Rupnagar, Punjab (India); Sharma, Sanjeev K.; Kim, Deuk Young [Semiconductor Materials and Device Laboratory, Department of Semiconductor Science, Dongguk University-Seoul, Seoul 100715 (Korea, Republic of); Kumar, S.; Chavan, N.M.; Joshi, S.V. [International Advanced Research Centre for Powder Metallurgy & New Materials (ARCI), Hyderabad 500005 (India); Singh, Narinder [Department of Chemistry, Indian Institute of Technology Ropar, Rupnagar, Punjab (India); Singh, Harpreet, E-mail: harpreetsingh@iitrpr.ac.in [School of Mechanical, Materials & Energy Engineering, Indian Institute of Technology Ropar, Rupnagar, Punjab (India)

    2015-02-15

    Highlights: • A presynthesized Ni-20Cr nanocrystalline powder was successfully deposited on T22 and SA 516 boilers steels using cold spray process. • The coatings are observed to have more than 2-folds microhardness in comparison with the base steels. • The coating was successful in reducing the weight gain of T22 and SA 516 steel by 71% and 94%. - Abstract: In the current investigation, high temperature oxidation behavior of a novel cold-spray Ni-20Cr nanostructured coating was studied. The nanocrystalline Ni-20Cr powder was synthesized by the investigators using ball milling, which was deposited on T22 and SA 516 steels by cold spraying. The crystallite size based upon Scherrer's formula for the developed coatings was found to be in nano-range for both the substrates. The accelerated oxidation testing was performed in a laboratory tube furnace at a temperature 900 °C under thermal cyclic conditions. Each cycle comprised heating for one hour at 900 °C followed by cooling for 20 min in ambient air. The kinetics of oxidation was established using weight change measurements for the bare and the coated steels. The oxidation products were characterized by X-ray Diffraction (XRD), Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) and X-ray mapping techniques. It was found from the results that the coating was successful in reducing the weight gain of SA213-T22 and SA 516-Grade 70 steel by 71% and 94%, respectively. This may be attributed to relatively denser structure, lower porosity and lower oxide content of the coating. Moreover, the developed nano-structured Ni-20Cr powder coating was found to perform better than its counterpart micron-sized Ni-20Cr powder coating, in terms of offering higher oxidation resistance and hardness.

  11. Effect of Chromium on Corrosion Behavior of P110 Steels in CO2-H2S Environment with High Pressure and High Temperature

    Directory of Open Access Journals (Sweden)

    Jianbo Sun

    2016-03-01

    Full Text Available The novel Cr-containing low alloy steels have exhibited good corrosion resistance in CO2 environment, mainly owing to the formation of Cr-enriched corrosion film. In order to evaluate whether it is applicable to the CO2 and H2S coexistence conditions, the corrosion behavior of low-chromium steels in CO2-H2S environment with high pressure and high temperature was investigated using weight loss measurement and surface characterization. The results showed that P110 steel suffered localized corrosion and both 3Cr-P110 and 5Cr-P110 steels exhibited general corrosion. However, the corrosion rate of 5Cr-P110 was the highest among them. The corrosion process of the steels was simultaneously governed by CO2 and H2S. The outer scales on the three steels mainly consisted of FeS1−x crystals, whereas the inner scales on Cr-containing steels comprised of amorphous FeS1−x, Cr(OH3 and FeCO3, in contrast with the amorphous FeS1−x and FeCO3 mixture film of P110 steel. The more chromium the steel contains, the more chromium compounds the corrosion products contain. The addition of chromium in steels increases the uniformity of the Cr-enriched corrosion scales, eliminates the localized corrosion, but cannot decrease the general corrosion rates. The formation of FeS1−x may interfere with Cr-enriched corrosion scales and lowering the corrosion performance of 3Cr-P110 and 5Cr-P110 steels.

  12. High temperature superconductors

    CERN Document Server

    Paranthaman, Parans

    2010-01-01

    This essential reference provides the most comprehensive presentation of the state of the art in the field of high temperature superconductors. This growing field of research and applications is currently being supported by numerous governmental and industrial initiatives in the United States, Asia and Europe to overcome grid energy distribution issues. The technology is particularly intended for densely populated areas. It is now being commercialized for power-delivery devices, such as power transmission lines and cables, motors and generators. Applications in electric utilities include current limiters, long transmission lines and energy-storage devices that will help industries avoid dips in electric power.

  13. High temperature radioisotope capsule

    International Nuclear Information System (INIS)

    Bradshaw, G.B.

    1976-01-01

    A high temperature radioisotope capsule made up of three concentric cylinders, with the isotope fuel located within the innermost cylinder is described. The innermost cylinder has hemispherical ends and is constructed of a tantalum alloy. The intermediate cylinder is made of a molybdenum alloy and is capable of withstanding the pressure generated by the alpha particle decay of the fuel. The outer cylinder is made of a platinum alloy of high resistance to corrosion. A gas separates the innermost cylinder from the intermediate cylinder and the intermediate cylinder from the outer cylinder

  14. High temperature reaction kinetics

    International Nuclear Information System (INIS)

    Jonah, C.D.; Beno, M.F.; Mulac, W.A.; Bartels, D.

    1985-01-01

    During the last year the dependence of the apparent rate of OD + CO on water pressure was measured at 305, 570, 865 and 1223 K. An explanation was found and tested for the H 2 O dependence of the apparent rate of OH(OD) + CO at high temperatures. The isotope effect for OH(D) with CO was determined over the temperature range 330 K to 1225 K. The reason for the water dependence of the rate of OH(OD) + CO near room temperatures has been investigated but no clear explanation has been found. 1 figure

  15. High temperature brazing of reactor materials

    International Nuclear Information System (INIS)

    Orlov, A.V.; Nechaev, V.A.; Rybkin, B.V.; Ponimash, I.D.

    1990-01-01

    Application of high-temperature brazing for joining products of such materials as molybdenum, tungsten, zirconium, beryllium, magnesium, nickel and aluminium alloys, graphite ceramics etc. is described. Brazing materials composition and brazed joints properties are presented. A satisfactory strength of brazed joints is detected under reactor operation temperatures and coolant and irradiation effect

  16. High temperature thermometric phosphors

    Science.gov (United States)

    Allison, Stephen W.; Cates, Michael R.; Boatner, Lynn A.; Gillies, George T.

    1999-03-23

    A high temperature phosphor consists essentially of a material having the general formula LuPO.sub.4 :Dy.sub.(x),Eu.sub.y) wherein: 0.1 wt %.ltoreq.x.ltoreq.20 wt % and 0.1 wt %.ltoreq.y.ltoreq.20 wt %. The high temperature phosphor is in contact with an article whose temperature is to be determined. The article having the phosphor in contact with it is placed in the environment for which the temperature of the article is to be determined. The phosphor is excited by a laser causing the phosphor to fluoresce. The emission from the phosphor is optically focused into a beam-splitting mirror which separates the emission into two separate emissions, the emission caused by the dysprosium dopant and the emission caused by the europium dopent. The separated emissions are optically filtered and the intensities of the emission are detected and measured. The ratio of the intensity of each emission is determined and the temperature of the article is calculated from the ratio of the intensities of the separate emissions.

  17. SCC growth behavior of cast stainless steels in high-temperature water. Influences of corrosion potential, steel type, thermal aging and cold-work

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Terachi, Takumi; Miyamoto, Tomoki; Arioka, Koji

    2011-01-01

    Recent studies on crack growth rate (CGR) measurement in oxygenated high-temperature pure water conditions, such as normal water chemistry (NWC) in BWRs, using compact tension (CT) type specimens have shown that stainless steel weld metal are susceptible to stress corrosion cracking (SCC). On the other hand, the authors reported that no significant SCC growth was observed on stainless steel weld metals in PWR primary water at temperatures from 250degC to 340degC. Cast austenitic stainless steels are widely used in light water reactors, and there is a similarity between welded and cast stainless steels in terms of the microstructure of the ferrite/austenite duplex structure. However, there are a few reports giving CGR data on cast stainless steels in the BWRs and PWRs. The principal purpose of this study was to examine the SCC growth behavior of cast stainless steels in simulated PWR primary water. A second objective was to examine the effects on SCC growth in hydrogenated and oxygenated water environments at 320degC of: (1) corrosion potential; (2) steels type (Mo in alloy); (3) thermal-aging (up to 400degC x 40 kh); and (4) cold-working (10%). The results were as follows: (1) No significant SCC growth was observed on all types of cast stainless steels: aged (400degC x 40 kh) of SCS14A and SCS13A and 10% cold-working, in hydrogenated (low-potential) water at 320degC. (2) Aging at 400degC x 40 kh SCS14A (10%CW) markedly accelerated the SCC growth of cast material in high-potential water at 320degC, but no significant SCC growth was observed in the hydrogenated water, even after long-term thermal aging (400degC x 40 kh). (3) Thus, cast stainless steels have excellent SCC resistance in PWR primary water. (4) On the other hand, significant SCC growth was observed on all types of cast stainless steels: 10%CW SCS14A and SCS13A, in 8 ppm-oxygenated (high-potential) water at 320degC. (5) No large difference in SCC growth was observed between SCS14A (Mo) and SCS13A. (6) No

  18. High-Temperature Tensile Behaviors of Base Metal and Electron Beam-Welded Joints of Ni-20Cr-9Mo-4Nb Superalloy

    Science.gov (United States)

    Gupta, R. K.; Anil Kumar, V.; Sukumaran, Arjun; Kumar, Vinod

    2018-05-01

    Electron beam welding of Ni-20Cr-9Mo-4Nb alloy sheets was carried out, and high-temperature tensile behaviors of base metal and weldments were studied. Tensile properties were evaluated at ambient temperature, at elevated temperatures of 625 °C to 1025 °C, and at strain rates of 0.1 to 0.001 s-1. Microstructure of the weld consisted of columnar dendritic structure and revealed epitaxial mode of solidification. Weld efficiency of 90 pct in terms of strength (UTS) was observed at ambient temperature and up to an elevated temperature of 850 °C. Reduction in strength continued with further increase of test temperature (up to 1025 °C); however, a significant improvement in pct elongation is found up to 775 °C, which was sustained even at higher test temperatures. The tensile behaviors of base metal and weldments were similar at the elevated temperatures at the respective strain rates. Strain hardening exponent `n' of the base metal and weldment was 0.519. Activation energy `Q' of base metal and EB weldments were 420 to 535 kJ mol-1 determined through isothermal tensile tests and 625 to 662 kJ mol-1 through jump-temperature tensile tests. Strain rate sensitivity `m' was low ( 775 °C) is attributed to the presence of recrystallized grains. Up to 700 °C, the deformation is through slip, where strain hardening is predominant and effect of strain rate is minimal. Between 775 °C to 850 °C, strain hardening is counterbalanced by flow softening, where cavitation limits the deformation (predominantly at lower strain rate). Above 925 °C, flow softening is predominant resulting in a significant reduction in strength. Presence of precipitates/accumulated strain at high strain rate results in high strength, but when the precipitates were coarsened at lower strain rates or precipitates were dissolved at a higher temperature, the result was a reduction in strength. Further, the accumulated strain assisted in recrystallization, which also resulted in a reduction in strength.

  19. Eu2O3: properties and irradiation behavior

    International Nuclear Information System (INIS)

    Pasto, A.E.; Martin, M.M.

    1977-08-01

    Europium sesquioxide is an excellent candidate control material for fast reactors. Its properties and behavior have been under extensive investigation at ORNL since 1972. This report is a compilation of the results of these efforts. Processes for synthesizing powders and fabricating dense pellets from them are described. Physical and chemical properties data measured on these pellets, along with their irradiation behavior, are also summarized

  20. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  1. Evaluation of high temperature pressure sensors

    International Nuclear Information System (INIS)

    Choi, In-Mook; Woo, Sam-Yong; Kim, Yong-Kyu

    2011-01-01

    It is becoming more important to measure the pressure in high temperature environments in many industrial fields. However, there is no appropriate evaluation system and compensation method for high temperature pressure sensors since most pressure standards have been established at room temperature. In order to evaluate the high temperature pressure sensors used in harsh environments, such as high temperatures above 250 deg. C, a specialized system has been constructed and evaluated in this study. The pressure standard established at room temperature is connected to a high temperature pressure sensor through a chiller. The sensor can be evaluated in conditions of changing standard pressures at constant temperatures and of changing temperatures at constant pressures. According to the evaluation conditions, two compensation methods are proposed to eliminate deviation due to sensitivity changes and nonlinear behaviors except thermal hysteresis.

  2. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  3. High temperature materials characterization

    Science.gov (United States)

    Workman, Gary L.

    1990-01-01

    A lab facility for measuring elastic moduli up to 1700 C was constructed and delivered. It was shown that the ultrasonic method can be used to determine elastic constants of materials from room temperature to their melting points. The ease in coupling high frequency acoustic energy is still a difficult task. Even now, new coupling materials and higher power ultrasonic pulsers are being suggested. The surface was only scratched in terms of showing the full capabilities of either technique used, especially since there is such a large learning curve in developing proper methodologies to take measurements into the high temperature region. The laser acoustic system does not seem to have sufficient precision at this time to replace the normal buffer rod methodology.

  4. High temperature metallic recuperator

    Science.gov (United States)

    Ward, M. E.; Solmon, N. G.; Smeltzer, C. E.

    1981-06-01

    An industrial 4.5 MM Btu/hr axial counterflow recuperator, fabricated to deliver 1600 F combustion air, was designed to handle rapid cyclic loading, a long life, acceptable costs, and a low maintenance requirement. A cost benefit anlysis of a high temperature waste heat recovery system utilizing the recurperator and components capable of 1600 F combustion air preheat shows that this system would have a payback period of less than two years. Fifteen companies and industrial associations were interviewed and expressed great interest in recuperation in large energy consuming industries. Determination of long term environmental effects on candidate recuperator tubing alloys was completed. Alloys found to be acceptable in the 2200 F flue gas environment of a steel billet reheat furnace, were identified.

  5. Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, Shinichiro, E-mail: yamashita.shinichiro@jaea.go.jp; Yano, Yasuhide; Ohtsuka, Satoshi; Yoshitake, Tsunemitsu; Kaito, Takeji; Koyama, Shin-ichi; Tanaka, Kenya

    2013-11-15

    Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS{sub F}/M, 12Cr–ODS{sub F}) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS{sub F}/M steel whereas no distinct temperature dependence for 12Cr–ODS{sub F} steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10{sup 26} n/m{sup 2} (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

  6. High temperature structural sandwich panels

    Science.gov (United States)

    Papakonstantinou, Christos G.

    High strength composites are being used for making lightweight structural panels that are being employed in aerospace, naval and automotive structures. Recently, there is renewed interest in use of these panels. The major problem of most commercial available sandwich panels is the fire resistance. A recently developed inorganic matrix is investigated for use in cases where fire and high temperature resistance are necessary. The focus of this dissertation is the development of a fireproof composite structural system. Sandwich panels made with polysialate matrices have an excellent potential for use in applications where exposure to high temperatures or fire is a concern. Commercial available sandwich panels will soften and lose nearly all of their compressive strength temperatures lower than 400°C. This dissertation consists of the state of the art, the experimental investigation and the analytical modeling. The state of the art covers the performance of existing high temperature composites, sandwich panels and reinforced concrete beams strengthened with Fiber Reinforced Polymers (FRP). The experimental part consists of four major components: (i) Development of a fireproof syntactic foam with maximum specific strength, (ii) Development of a lightweight syntactic foam based on polystyrene spheres, (iii) Development of the composite system for the skins. The variables are the skin thickness, modulus of elasticity of skin and high temperature resistance, and (iv) Experimental evaluation of the flexural behavior of sandwich panels. Analytical modeling consists of a model for the flexural behavior of lightweight sandwich panels, and a model for deflection calculations of reinforced concrete beams strengthened with FRP subjected to fatigue loading. The experimental and analytical results show that sandwich panels made with polysialate matrices and ceramic spheres do not lose their load bearing capability during severe fire exposure, where temperatures reach several

  7. INTERWELD - European project to determine irradiation induced material changes in the heat affected zones of austenitic stainless steel welds that influence the stress corrosion behaviour in high-temperature water

    International Nuclear Information System (INIS)

    Roth, A.; Schaaf, Bob van der; Castano, M.L.; Ohms, C.; Gavillet, D.; Dyck, S. van

    2003-01-01

    PWR and BWR RPV internals have experienced stress corrosion cracking in service. The objective of the INTERWELD project is to determine the radiation induced material changes that promote stress corrosion cracking in the heat affected zone of austenitic stainless steel welds. To achieve this goal, welds in austenitic stainless steel types AISI 304/347 have been fabricated, respectively. Stress-relief annealing was applied optionally. The pre-characterisation of both the as-welded and stress relieved material conditions comprises the examination of the weld residual stresses by the ring-core-technique and neutron diffraction, the degree of sensitisation by EPR, and the stress corrosion behaviour by SSRT testing in high-temperature water. The weldments will be irratiated to 2 neutron fluence levels and a postirradiation examination will determine micromechanical, microchemical and microstructural changes in the materials. In detail, the evolution of the residual stress levels and the stress corrosion behaviour after irradiation will be determined. Neutron diffraction will be utilized for the first time with respect to neutron irradiated material. In this paper, the current state of the project will be described and discussed. (orig.)

  8. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Vapor film collapse behavior on high temperature particle surface. JAERI's nuclear research promotion program, H10-027-3. Contract research

    International Nuclear Information System (INIS)

    Abe, Yutaka

    2002-03-01

    The experimental researches were conducted to study vapor film collapse behavior on high temperature melted core material coarsely mixed in the coolant under the film boiling condition. The film collapse is very important incipient incident of the trigger process for the vapor explosion in sever accident of nuclear reactor. In the experiment, pressure pulse was applied to the vapor film on a high temperature particle surface simulating melted core material to observed microscopic vapor film collapse behavior with a high-speed video camera of 40,500 fps. The particle surface temperature and pressure around the particle were simultaneously measured. The transition of the vapor film thickness and two-dimensional vapor-liquid interface movement and the velocity were estimated with visual data analysis technique, PIV and digital data analysis technique. Furthermore, heat conduction analysis was performed to estimate the vapor-liquid interfacial temperature with the measured temperature and estimated vapor film thickness. As the results, it was clarified that the vapor-liquid interface changed white from transparent view for all the experimental conditions. It is also clarified that the vapor-liquid interfacial temperature decreased under the saturation temperature when the pressure pulse arrive at the particle. The experimental facts indicates the possibility that the vapor film collapse occurs due to the liquid phase homogeneous moving toward the particle drove by the pressure reduction caused by the phase change inside the vapor film. (author)

  9. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Vapor film collapse behavior on high temperature particle surface. JAERI's nuclear research promotion program, H10-027-3. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Yutaka [Tsukuba Univ., Institute of Engineering Mechanics and Systems, Tsukuba, Ibaraki (Japan)

    2002-03-01

    The experimental researches were conducted to study vapor film collapse behavior on high temperature melted core material coarsely mixed in the coolant under the film boiling condition. The film collapse is very important incipient incident of the trigger process for the vapor explosion in sever accident of nuclear reactor. In the experiment, pressure pulse was applied to the vapor film on a high temperature particle surface simulating melted core material to observed microscopic vapor film collapse behavior with a high-speed video camera of 40,500 fps. The particle surface temperature and pressure around the particle were simultaneously measured. The transition of the vapor film thickness and two-dimensional vapor-liquid interface movement and the velocity were estimated with visual data analysis technique, PIV and digital data analysis technique. Furthermore, heat conduction analysis was performed to estimate the vapor-liquid interfacial temperature with the measured temperature and estimated vapor film thickness. As the results, it was clarified that the vapor-liquid interface changed white from transparent view for all the experimental conditions. It is also clarified that the vapor-liquid interfacial temperature decreased under the saturation temperature when the pressure pulse arrive at the particle. The experimental facts indicates the possibility that the vapor film collapse occurs due to the liquid phase homogeneous moving toward the particle drove by the pressure reduction caused by the phase change inside the vapor film. (author)

  10. Evidence for single-chain magnet behavior in a Mn(III)-Ni(II) chain designed with high spin magnetic units: a route to high temperature metastable magnets.

    Science.gov (United States)

    Clérac, Rodolphe; Miyasaka, Hitoshi; Yamashita, Masahiro; Coulon, Claude

    2002-10-30

    . This result indicates the presence of a metastable state without magnetic long-range order. This material is the first experimental design of a heterometallic chain with ST = 3 magnetic units showing a "single-chain magnet" behavior predicted in 1963 by R. J. Glauber for an Ising one-dimensional system. This work opens new perspectives for one-dimensional systems to obtain high temperature metastable magnets by combining high spin magnetic units, strong interunit interactions, and uniaxial anisotropy.

  11. SCC growth behavior of stainless steel weld metals in high-temperature water. Influence of corrosion potential, weld type, thermal aging, cold-work and temperature

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Terachi, Takumi; Miyamoto, Tomoki; Arioka, Koji

    2009-01-01

    Recent studies on crack growth rate measurement in oxygenated high-temperature pure water conditions, such as normal water chemistry in boiling water reactors, using compact tension type specimens have shown that weld stainless steels are susceptible to stress corrosion cracking. However, to our knowledge, there is no crack growth data of weld stainless steels in pressurized water reactor primary water. The principal purpose of this study was to examine the SCC growth behavior of stainless steel weld metals in simulated PWR primary water. A second objective was to examine the effect of (1) corrosion potential, (2) thermal-aging, (3) Mo in alloy and (4) cold-working on SCC growth in hydrogenated and oxygenated water environments at 320degC. In addition, the temperature dependence of SCC growth in simulated PWR primary water was also studied. The results were as follows: (1) No significant SCC growth was observed on all types of stainless steel weld metals: as-welded, aged (400degC x 10 kh) 308L and 316L, in 2.7 ppm-hydrogenated (low-potential) water at 320degC. (2) 20% cold-working markedly accelerated the SCC growth of weld metals in high-potential water at 320degC, but no significant SCC growth was observed in the hydrogenated water, even after 20% cold-working. (3) No significant SCC growth was observed on stainless steel weld metals in low-potential water at 250degC and 340degC. Thus, stainless steel weld metals have excellent SCC resistance in PWR primary water. On the other hand, (4) significant SCC growth was observed on all types of stainless steel weld metals: as-weld, aged (400degC x 10 kh) and 20% cold-worked 308L and 316L, in 8 ppm-oxygenated (high-potential) water at 320degC. (5) No large difference in SCC growth was observed between 316L (Mo) and 308L. (6) No large effect on SCC growth was observed between before and after aging up to 400degC for 10 kh. (7) 20% cold-working markedly accelerated the SCC growth of stainless steel weld metals. (author)

  12. High temperature materials and mechanisms

    CERN Document Server

    2014-01-01

    The use of high-temperature materials in current and future applications, including silicone materials for handling hot foods and metal alloys for developing high-speed aircraft and spacecraft systems, has generated a growing interest in high-temperature technologies. High Temperature Materials and Mechanisms explores a broad range of issues related to high-temperature materials and mechanisms that operate in harsh conditions. While some applications involve the use of materials at high temperatures, others require materials processed at high temperatures for use at room temperature. High-temperature materials must also be resistant to related causes of damage, such as oxidation and corrosion, which are accelerated with increased temperatures. This book examines high-temperature materials and mechanisms from many angles. It covers the topics of processes, materials characterization methods, and the nondestructive evaluation and health monitoring of high-temperature materials and structures. It describes the ...

  13. The compression behavior of blödite at low and high temperature up to ~10GPa: Implications for the stability of hydrous sulfates on icy planetary bodies

    Energy Technology Data Exchange (ETDEWEB)

    Comodi, Paola; Stagno, Vincenzo; Zucchini, Azzurra; Fei, Yingwei; Prakapenka, Vitali

    2017-03-01

    Recent satellite inferences of hydrous sulfates as recurrent minerals on the surface of icy planetary bodies link with the potential mineral composition of their interior. Blödite, a mixed Mg-Na sulfate, is here taken as representative mineral of icy satellites surface to investigate its crystal structure and stability at conditions of the interior of icy bodies. To this aim we performed in situ synchrotron angle-dispersive X-ray powder diffraction experiments on natural blödite at pressures up to ~10.4 GPa and temperatures from ~118.8 K to ~490.0 K using diamond anvil cell technique to investigate the compression behavior and establish a low-to-high temperature equation of state that can be used as reference when modeling the interior of sulfate-rich icy satellites such as Ganymede. The experimentally determined volume expansivity, α, varies from 7.6 (7) 10-5 K-1 at 0.0001 GPa (from 118.8 to 413.15 K) to 2.6 (3) 10-5 K-1 at 10 GPa (from 313.0 to 453.0 K) with a δα/δP coefficient = -5.6(9)10-6 GPa-1 K-1. The bulk modulus calculated from the least squares fitting of P-V data on the isotherm at 413 K using a second-order Birch - Murnaghan equation of state is 38(5) GPa, which gives the value of δK/δT equal to 0.01(5) GPa K-1. The thermo-baric behavior of blödite appears strongly anisotropic with c lattice parameter being more deformed with respect to a and b. Thermogravimetric analyses performed at ambient pressure showed three endotherms at 413 K, 533 K and 973 K with weight losses of approximately 11%, 11% and 43% caused by partial dehydration, full dehydration and sulfate decomposition respectively. Interestingly, no clear evidence of dehydration was observed up to ~453 K and ~10.4 GPa, suggesting that pressure acts to stabilize the crystalline structure of blödite. The data collected allow to write the following equation of state, V(P, T) = V

  14. High temperature interface superconductivity

    International Nuclear Information System (INIS)

    Gozar, A.; Bozovic, I.

    2016-01-01

    Highlight: • This review article covers the topic of high temperature interface superconductivity. • New materials and techniques used for achieving interface superconductivity are discussed. • We emphasize the role played by the differences in structure and electronic properties at the interface with respect to the bulk of the constituents. - Abstract: High-T_c superconductivity at interfaces has a history of more than a couple of decades. In this review we focus our attention on copper-oxide based heterostructures and multi-layers. We first discuss the technique, atomic layer-by-layer molecular beam epitaxy (ALL-MBE) engineering, that enabled High-T_c Interface Superconductivity (HT-IS), and the challenges associated with the realization of high quality interfaces. Then we turn our attention to the experiments which shed light on the structure and properties of interfacial layers, allowing comparison to those of single-phase films and bulk crystals. Both ‘passive’ hetero-structures as well as surface-induced effects by external gating are discussed. We conclude by comparing HT-IS in cuprates and in other classes of materials, especially Fe-based superconductors, and by examining the grand challenges currently laying ahead for the field.

  15. High-temperature uncertainty

    International Nuclear Information System (INIS)

    Timusk, T.

    2005-01-01

    Recent experiments reveal that the mechanism responsible for the superconducting properties of cuprate materials is even more mysterious than we thought. Two decades ago, Georg Bednorz and Alex Mueller of IBM's research laboratory in Zurich rocked the world of physics when they discovered a material that lost all resistance to electrical current at the record temperature of 36 K. Until then, superconductivity was thought to be a strictly low-temperature phenomenon that required costly refrigeration. Moreover, the IBM discovery - for which Bednorz and Mueller were awarded the 1987 Nobel Prize for Physics - was made in a ceramic copper-oxide material that nobody expected to be particularly special. Proposed applications for these 'cuprates' abounded. High-temperature superconductivity, particularly if it could be extended to room temperature, offered the promise of levitating trains, ultra-efficient power cables, and even supercomputers based on superconducting quantum interference devices. But these applications have been slow to materialize. Moreover, almost 20 years on, the physics behind this strange state of matter remains a mystery. (U.K.)

  16. Modeling of coated fuel particles irradiation behavior

    International Nuclear Information System (INIS)

    Liang Tongxiang; Phelip, M.

    2006-01-01

    In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10 -7 . As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

  17. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  19. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    Science.gov (United States)

    Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

    2017-05-01

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments.

  20. High temperature pipeline design

    Energy Technology Data Exchange (ETDEWEB)

    Greenslade, J.G. [Colt Engineering, Calgary, AB (Canada). Pipelines Dept.; Nixon, J.F. [Nixon Geotech Ltd., Calgary, AB (Canada); Dyck, D.W. [Stress Tech Engineering Inc., Calgary, AB (Canada)

    2004-07-01

    It is impractical to transport bitumen and heavy oil by pipelines at ambient temperature unless diluents are added to reduce the viscosity. A diluted bitumen pipeline is commonly referred to as a dilbit pipeline. The diluent routinely used is natural gas condensate. Since natural gas condensate is limited in supply, it must be recovered and reused at high cost. This paper presented an alternative to the use of diluent to reduce the viscosity of heavy oil or bitumen. The following two basic design issues for a hot bitumen (hotbit) pipeline were presented: (1) modelling the restart problem, and, (2) establishing the maximum practical operating temperature. The transient behaviour during restart of a high temperature pipeline carrying viscous fluids was modelled using the concept of flow capacity. Although the design conditions were hypothetical, they could be encountered in the Athabasca oilsands. It was shown that environmental disturbances occur when the fluid is cooled during shut down because the ground temperature near the pipeline rises. This can change growing conditions, even near deeply buried insulated pipelines. Axial thermal loads also constrain the design and operation of a buried pipeline as higher operating temperatures are considered. As such, strain based design provides the opportunity to design for higher operating temperature than allowable stress based design methods. Expansion loops can partially relieve the thermal stress at a given temperature. As the design temperature increase, there is a point at which above grade pipelines become attractive options, although the materials and welding procedures must be suitable for low temperature service. 3 refs., 1 tab., 10 figs.

  1. Irradiation behavior of metallic fast reactor fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985

  2. Co and Fe-catalysts supported on sepiolite: effects of preparation conditions on their catalytic behaviors in high temperature gas flow treatment of dye.

    Science.gov (United States)

    Lin, Xiangfeng; Fang, Jian; Chen, Menglin; Huang, Zhi; Su, Chengyuan

    2016-08-01

    An efficient adsorbent/catalyst Co and Fe-catalysts loaded on sepiolite (Co-Fe/sepiolite) was successfully prepared for high temperature gas flow catalytic reaction by a simple impregnation method. The impact of preparation conditions (such as pH value of impregnation solution, impregnation time, calcination temperature, and time) on catalytic activity was studied. We found that the catalytic activity of Co-Fe/sepiolite was strongly influenced by all the investigated parameters. The regeneration efficiency (RE) was used to evaluate the catalytic activity. The RE is more noticeable at pH 5.0 of impregnation solution, impregnation time 18 h, calcination temperature 650 °C, and calcination time 3 h. This Co-Fe/sepiolite has great adsorption capacity in absorbing dye. It is used for an adsorbent to adsorb dye from wastewater solution under dynamic adsorption and saturated with dye, then regenerated with high temperature gas flow for adsorption/oxidation cycles. The Co-Fe/sepiolite acts as a catalyst to degrade the dye during regeneration under high temperature gas flow. Hence, the Co-Fe/sepiolite is not only an adsorbent but also a catalyst. The Co-Fe/sepiolite is more stable than sepiolite when applied in the treatment of plant's wastewater. The Co-Fe/sepiolite can be reused in adsorption-regeneration cycle. The results indicate the usability of the proposed combined process, dye adsorption on Co-Fe/sepiolite followed by the catalytic oxidation in high temperature gas flow.

  3. The behavior of ZrO_2/20%Y_2O_3 and Al_2O_3 coatings deposited on aluminum alloys at high temperature regime

    International Nuclear Information System (INIS)

    Pintilei, G.L.; Crismaru, V.I.; Abrudeanu, M.; Munteanu, C.; Baciu, E.R.; Istrate, B.; Basescu, N.

    2015-01-01

    Highlights: • In both the ZrO_2/20%Y_2O_3 and Al_2O_3 coatings the high temperature caused a decrease of pores volume and a lower thickness of the interface between successive splats. • The NiCr bond layer in the sample with a ZrO_2/20%Y_2O_3 suffered a fragmentation due to high temperature exposure and thermal expansion which can lead to coating exfoliation. • The NiCr bond layer in the sample with an Al_2O_3 coating showed an increase of pore volume due to high temperature. - Abstract: Aluminum alloy present numerous advantages like lightness, high specific strength and diversity which recommend them to a high number of applications from different fields. In extreme environments the protection of aluminum alloys is difficult and requires a high number of requirements like high temperature resistance, thermal fatigue resistance, corrosion fatigue resistance and galvanic corrosion resistance. To obtain these characteristics coatings can be applied to the surfaces so they can enhance the mechanical and chemical properties of the parts. In this paper two coatings were considered for deposition on an AA2024 aluminum alloy, ZrO_2/20%Y_2O_3 and Al_2O_3. To obtain a better adherence of the coating to the base material an additional bond layer of NiCr is used. Both the coatings and bond layer were deposited by atmospheric plasma spraying on the samples. The samples were subjected to a temperature of 500 °C and after that slowly cooled to room temperature. The samples were analyzed by electron microscopy and X-ray diffraction to determine the morphological and phase changes that occurred during the temperature exposure. To determine the stress level in the parts due to thermal expansion a finite element analysis was performed in the same conditions as the tests.

  4. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Gu, E-mail: jglee88@ulsan.ac.kr [School of Materials Science and Engineering, University of Ulsan, Ulsan 44610 (Korea, Republic of); Lee, Gyoung-Ja; Park, Jin-Ju [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of); Lee, Min-Ku, E-mail: leeminku@kaeri.re.kr [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of)

    2017-05-15

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

  5. Role of cavity formation in SCC of cold worked carbon steel in high-temperature water. Part 2. Study of crack initiation behavior

    International Nuclear Information System (INIS)

    Yamada, Takuyo; Aoki, Masanori; Miyamoto, Tomoki; Arioka, Koji

    2013-01-01

    To consider the role of cavity formation in stress corrosion cracking (SCC) of cold worked (CW) carbon steel in high-temperature water, SCC and creep growth (part 1) and initiation (part 2) tests were performed. The part 2 crack initiation tests used blunt notched compact tension (CT) type specimens of CW carbon steel exposed under the static load condition in hydrogenated pure water and in air in the range of temperatures between 360 and 450°C. Inter-granular (IG) crack initiation was observed both in water and in air even in static load condition when steel specimens had been cold worked. 1/T type temperature dependencies of initiation times were observed for CW carbon steel, and the crack initiation times in an operating pressurized heavy water reactor, PHWR (Pt Lepreau) seemed to lie on the extrapolated line of the experimental results. Cavities were identified at the grain boundaries near the bottom of a notch (highly stressed location) before cracks initiated both in water and air. The cavities were probably formed by the condensation of vacancies and they affected the bond strength of the grain boundaries. To assess the mechanism of IGSCC initiation in high temperature water, the diffusion of vacancies driven by stress gradients was studied using a specially designed CT specimen. As a model for IGSCC in CW carbon steel in high temperature water, it was concluded that the formation of cavities from the collapse of vacancies offers the best interpretation of the present data. (author)

  6. Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

    International Nuclear Information System (INIS)

    Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

    2017-01-01

    The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

  7. Fluence behavior of polyethylene films irradiated with high energy electrons

    International Nuclear Information System (INIS)

    Pino, Eddy Segura; Silva, Leonardo G. Andrade e

    1999-01-01

    Polymers are viscoelastic materials at all temperatures, so that mechanical loads induce time dependable deformations. The recovery of these deformations, on load release, take some time and it is not always recovered completely. The main objective of this work was to analyse the creep behavior of electron irradiated polyethylene films. From the experimental results, it was sated that polyethylene creeps less with an increase on irradiation dose and also that creep recovery in this material increases with doses but it is not complete. This behavior can be attributed to the crosslinking effect witch stabilize elements of the molecular structure of the polyethylene, thus reducing their mobility and so inhibiting the creep mechanism. The partial creep recovery could be also attributed to the reticulation effect and to the polyethylene plastic behavior. Additional information on the creep behavior was obtained by fitting the experimental data with exponential functions and evaluating the mathematical parameters with a modified Kelvin-Voigt mechanical model. (author)

  8. Irradiation behavior of miniature experimental uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

  9. Irradiation Behavior and Post-Irradiation Examinations of an Acoustic Sensor Using a Piezoelectric Transducer

    International Nuclear Information System (INIS)

    Lambert, T.; Zacharie-Aubrun, I.; Hanifi, K.; Valot, Ch.; Fayette, L.; Rosenkantz, E.; Ferrandis, J.Y.; Tiratay, X.

    2013-06-01

    The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. In the framework of high burn-up fuel experiments under transient operating conditions, an innovative sensor based on acoustic method was developed by CEA and IES (Southern Electronic Institute).This sensor is used to determine the on-line composition of the gases located in fuel rodlet free volume and thus, allows calculating the molar fractions of fission gases and helium. The main principle of the composition determination by acoustic method consists in measuring the time of flight of an acoustic signal emitted and reflected in a specific cavity. A piezoelectric transducer, driven by a pulse generator, generates the acoustic wave in the cavity. The piezoelectric transducer is a PZT ceramic disk, mainly consisting of lead, zirconium and titanium. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. However, during the irradiation test, acoustic signal degradation was observed, mainly due to irradiation effect but also due to the increasing of the gas temperature. Despite this acoustic signal degradation, the time of flight measurements were carried out with good accuracy throughout the test, thanks to the development of a more efficient signal processing. After experiment, neutronic calculations were performed in order to determine neutron fluence at the level of the piezoelectric transducer. In order to have a better understanding of the acoustic sensor behavior under irradiation, Post Irradiation Examination program was done on piezoelectric transducer and on acoustic coupling material too. These examinations were also realized on a non-irradiated acoustic sensor built in the same conditions and with the same materials and the same

  10. High Temperature Studies of La-Monazite

    Science.gov (United States)

    2004-07-01

    Hay, E. Boakeye, M. D. Petry, Y. Berta, K. Von Lehmden, and J. Welch, " 5 A. Meldrum , L. A. Boatner, and R. C. Ewing, "Electron-Irradiation-Induced... Meldrum , L. A. Boatner, and R. C. Ewing, "A Comparison of Radiation Alumina-based Fiber for High Temperature Composite Reinforcement," Ceram. Eng... acid . The processing included procedures that allowed the La/P ratio to be controlled to be very close to the stoichiometric value of unity (within less

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

    Energy Technology Data Exchange (ETDEWEB)

    Lambard, V

    2000-07-01

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

  12. Effects of prenatal ionizing irradiation on neural function and behavior

    International Nuclear Information System (INIS)

    Brizzee, K.R.; Ordy, J.M.; Pennwalt Corp., Rochester, NY

    1986-01-01

    Behavioral studies in the past decade on postnatal effects of prenatal ionizing irradiation (125-R) revealed alterations in circadian locomotor activity and modifications of duration, frequency and sequences of certain behavioral acts in irradiated rats. Other studies have shown that the effect of irradiation (150-R) and enriched environment were both significant in initial, repetitive and total error scores while at 200-R enrichment was less effective. Rats irradiated on the 13th, 14th and 15th day of gestation were born with a hopping gait, paired hind and forelimbs moving in unison. Thoracic cord section led to crossed extention hind-limb reflexes in control rats and simultaneus withdrawal of hind limbs in hopping rats, in response to pinprick. In 90 day old squirrel monkeys the percent of correct response in visual orientation, discrimination and reversal learning in 50- and 100-R offspring (Co 60 irradiation at 89-90 days gestation) were significantly lower than those of controls, and differences in reversal learning persisted undiminished at 2 years of age. Time required for body righting, head-up orientation and climbing at 45 0 incline was significantly greater in irradiated (mainly 100-R) than control animals at from 2 to 28 days of age. Visual acuity levels of 50- and 100-R 30 day old infants were significantly lower than in control infants. Stabilimeter activity in the dark was significantly higher in 50- and 100-R 30 day old infants than in controls. In the squirrel monkey studies effects of Co 60 irradiation (100-R) on postnatal development of the brain and behaviour can be identified at statistically significant levels of confidence independent of offspring nursery rearing effects. (orig.)

  13. Influence of Zn injection on corrosion behavior and oxide film characteristics of 304 stainless steel in borated and lithiated high temperature water

    International Nuclear Information System (INIS)

    Wu, Xinqiang; Liu, Xiahe; Han, En-Hou; Ke, Wei

    2012-09-01

    Water chemistry of the reactor coolant system plays a major role in maintaining safety and reliability of light water reactor nuclear power plants (NPPs). Zn water chemistry into pressurized water reactors (PWRs) in order to reduce the radiation buildup in primary coolant system has been widely applied, and the reduction effect has been experimentally confirmed. Zn injection can also lessen the corrosion phenomena in high temperature pressurized water by changing oxide films formed on components materials. Both the radiation buildup and material corrosion resistance in PWR coolant system are closely dependent on the oxide films formed. However, the influence of Zn injection on the chemical composition and structure of the oxide films on their protective properties is still a matter of considerable debate. The influence of Zn injection on corrosion inhibition and environmental degradation has not been fully clarified yet. Therefore, the understanding of corrosion behaviour, oxide film characteristics and their protective property is of significance to clarify the environmentally assisted material failure problems in NPPs. In the present work, oxide films formed on nuclear-grade 304 SS exposed to borated and lithiated high temperature water environments at 300 deg. C up to 4000 h with or without 10 ppb Zn injection were investigated ex-situ. Without Zn injection, the oxide films mainly consisted of Fe 3 O 4 and FeCr 2 O 4 . With Zn injection, ZnFe 2 O 4 and ZnCr 2 O 4 were detected in the oxide films at the initial stage of immersion and ZnCr 2 O 4 became dominant after long-term immersion. It was believed that the above Zn-Fe and Zn-Cr spinel oxides were formed by substitution reactions between Zn 2+ and Fe 2+ . At the initial stage of immersion, water chemistry significantly affected the formation of the oxide films. Once a stable oxide film formed, it is rather difficult to change its structure through changing water chemistry. The potential-pH diagrams for Zn

  14. Behavior of fluorine 18 in neutron irradiated zeolites

    International Nuclear Information System (INIS)

    Estevez Lopez, D.R.

    1992-01-01

    The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

  15. Effect of combined addition of N and Nb on the high temperature behavior of a 25Cr-20Ni stainless steel

    International Nuclear Information System (INIS)

    Park, In Duck; Nam, Ki Woo

    2002-01-01

    In order to clarify the effect of precipitates on creep strength at high temperature in 25Cr-20Ni stainless steels, threshold stress and void-hardening stress have been measured and compared each others. The value of threshold stress for high temperature, measured by stress abruptly loading test, is about 130 MPa. Threshold stress must be measured by the function of stress loading time in the combination hardening steel of solution hardening and precipitate hardening. Average diameter and inter-partial distance of precipitates from TEM microstructure can determine void-hardening stress. The value of void-hardening stress, evaluated by using Scattergood and Bacon's equation, was from 101 to 130 MPa. The ratio of average void hardening to threshold stress is 1.13 and both values are not equal. The result from analyzing the electron diffraction pattern shows that the dispersed precipitates in SUS310J1TB is NbCrN nitrides. Mechanism of the interaction between dislocations and precipitate particles in SUS310J1TB is Srolovitz mechanism, which is a gravitation-type interaction

  16. Effect of gamma irradiation on the behavioral properties of crotoxin

    Directory of Open Access Journals (Sweden)

    E.G. Moreira

    1997-02-01

    Full Text Available Crotoxin has been detoxified with gamma radiation in order to improve crotalic antiserum production. Nevertheless, present knowledge of the biological characteristics of irradiated crotoxin is insufficient to propose it as an immunizing agent. Crotoxin is known to increase the emotional state of rats and to decrease their exploratory behavior (Moreira EG, Nascimento N, Rosa GJM, Rogero JR and Vassilieff VS (1996 Brazilian Journal of Medical and Biological Research, 29: 629-632. Therefore, we decided 1 to evaluate the effects of crotoxin in the social interaction test, which has been widely used for the evaluation of anxiogenic drugs, and 2 to determine if irradiated crotoxin induces behavioral alterations similar to those of crotoxin in the social interaction, open-field and hole-board tests. Male Wistar rats (180-220 g were used. Crotoxin (100, 250, and 500 µg/kg was injected intraperitoneally 2 h before the social interaction test. Similarly, irradiated crotoxin (2000 Gy gamma radiation from a 60Co source was administered at the doses of 100, 250, and 500 µg/kg for the hole-board test, and at the doses of 1000 and 2500 µg/kg for the open-field and social interaction tests. ANOVA complemented with the Dunnett test was used for statistical analysis (P<0.05. Crotoxin decreased the social interaction time (s at the doses of 100, 250 and 500 µg/kg (means ± SEM from 51.6 ± 4.4 to 32.6 ± 3.7, 28.0 ± 3.6 and 31.6 ± 4.4, respectively. Irradiated crotoxin did not induce behavioral alterations. These results indicate that 1 crotoxin may be an anxiogenic compound, and 2 in contrast to crotoxin, irradiated crotoxin was unable to induce behavioral alterations, which makes it a promising compound for the production of crotalic antiserum

  17. Creep of high temperature composites

    International Nuclear Information System (INIS)

    Sadananda, K.; Feng, C.R.

    1993-01-01

    High temperature creep deformation of composites is examined. Creep of composites depends on the interplay of many factors. One of the basic issues in the design of the creep resistant composites is the ability to predict their creep behavior from the knowledge of the creep behavior of the individual components. In this report, the existing theoretical models based on continuum mechanics principles are reviewed. These models are evaluated using extensive experimental data on molydisilicide-silicon carbide composites obtained by the authors. The analysis shows that the rule of mixture based on isostrain and isostress provides two limiting bounds wherein all other theoretical predictions fall. For molydisilicide composites, the creep is predominantly governed by the creep of the majority phase, i.e. the matrix with fibers deforming elastically. The role of back stresses both on creep rates and activation energies are shown to be minimum. Kinetics of creep in MoSi 2 is shown to be controlled by the process of dislocation glide with climb involving the diffusion of Mo atoms

  18. High-Temperature Piezoelectric Sensing

    Directory of Open Access Journals (Sweden)

    Xiaoning Jiang

    2013-12-01

    Full Text Available Piezoelectric sensing is of increasing interest for high-temperature applications in aerospace, automotive, power plants and material processing due to its low cost, compact sensor size and simple signal conditioning, in comparison with other high-temperature sensing techniques. This paper presented an overview of high-temperature piezoelectric sensing techniques. Firstly, different types of high-temperature piezoelectric single crystals, electrode materials, and their pros and cons are discussed. Secondly, recent work on high-temperature piezoelectric sensors including accelerometer, surface acoustic wave sensor, ultrasound transducer, acoustic emission sensor, gas sensor, and pressure sensor for temperatures up to 1,250 °C were reviewed. Finally, discussions of existing challenges and future work for high-temperature piezoelectric sensing are presented.

  19. Irradiation behavior of uranium oxide - Aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Rest, Jeffrey; Snelgrove, James L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products and as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show that, with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 g U/cm 3 ) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼63% 235 U burnup). (author)

  20. Irradiation behavior of uranium oxide-aluminum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    1996-01-01

    An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO 2 -Al dispersion fuel. The aluminum-fuel interaction models were developed based on U 3 O 8 -Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U 3 O 8 are valid for UO 2 , the LEU UO 2 -Al with a 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x 10 27 fissions m -3 (∼ 63% 235 U burnup)

  1. High Temperature Superconductor Resonator Detectors

    Data.gov (United States)

    National Aeronautics and Space Administration — High Temperature Superconductor (HTS) infrared detectors were studied for years but never matured sufficiently for infusion into instruments. Several recent...

  2. Difference in the Dissolution Behaviors of Tablets Containing Polyvinylpolypyrrolidone (PVPP) Depending on Pharmaceutical Formulation After Storage Under High Temperature and Humid Conditions.

    Science.gov (United States)

    Takekuma, Yoh; Ishizaka, Haruka; Sumi, Masato; Sato, Yuki; Sugawara, Mitsuru

    Storage under high temperature and humid conditions has been reported to decrease the dissolution rate for some kinds of tablets containing polyvinylpolypyrrolidone (PVPP) as a disintegrant. The aim of this study was to elucidate the properties of pharmaceutical formulations with PVPP that cause a decrease in the dissolution rate after storage under high temperature and humid conditions by using model tablets with a simple composition. Model tablets, which consisted of rosuvastatin calcium or 5 simple structure compounds, salicylic acid, 2-aminodiphenylmethane, 2-aminobiphenyl, 2-(p-tolyl)benzoic acid or 4.4'-biphenol as principal agents, cellulose, lactose hydrate, PVPP and magnesium stearate as additives, were made by direct compression. The model tables were wrapped in paraffin papers and stored for 2 weeks at 40°C/75% relative humidity (RH). Dissolution tests were carried out by the paddle method in the Japanese Pharmacopoeia 16th edition. Model tablets with a simple composition were able to reproduce a decreased dissolution rate after storage at 40°C/75% RH. These tablets showed significantly decreased water absorption activities after storage. In the case of tablets without lactose hydrate by replacing with cellulose, a decreased dissolution rate was not observed. Carboxyl and amino groups in the structure of the principal agent were not directly involved in the decreased dissolution. 2-Benzylaniline tablets showed a remarkably decreased dissolution rate and 2-aminobiphenyl and 2-(p-tolyl)benzoic acid tablets showed slightly decreased dissolution rates, though 4,4'-biphenol tablets did not show a decrease dissolution rate. We demonstrated that additives and structure of the principal agent were involved in the decreased in dissolution rate for tablets with PVPP. The results suggested that one of the reasons for a decreased dissolution rate was the inclusion of lactose hydrate in tablets. The results also indicated that compounds as principal agents with low

  3. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil, E-mail: hgkim@kaeri.re.kr; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun

    2015-10-15

    A 3D laser coating technology using Cr powder was developed for Zr-based alloys considering parameters such as: the laser beam power, inert gas flow, cooling of Zr-based alloys, and Cr powder control. This technology was then applied to Zr cladding tube samples to study the effect of Cr coating on the high-temperature oxidation of Zr-based alloys in a steam environment of 1200 °C for 2000s. It was revealed that the oxide layer thickness formed on the Cr-coated tube surface was about 25-times lower than that formed on a Zircaloy-4 tube surface. In addition, both the ring compression and the tensile tests were performed to evaluate the adhesion properties of the Cr-coated sample. Although some cracks were formed on the Cr-coated layer, the Cr-coated layer had not peeled off after the two tests.

  4. Nuclear graphite for high temperature reactors

    International Nuclear Information System (INIS)

    Marsden, B.J.

    2001-01-01

    The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)

  5. Deformation behavior of cell spring of an irradiated spacer grid

    International Nuclear Information System (INIS)

    Jin, Y. G.; Baek, S. J.; Ryu, W. S.; Kim, G. S.; Yoo, B. O.; Kim, D. S.; Ahn, S. B.; Chun, Y. B.; Choo, Y. S.

    2012-01-01

    Mechanical properties of a space grid of a fuel assembly are of great importance for fuel operation reliability in extended fuel burnup and duration of fuel life. The spacer grid with inner and outer straps has cell spring and dimples, which are in contact with the fuel rod. The spacer grids supporting the fuel rods absorb vibration impacts due to the reactor coolant flow and also grid spring force is decreasing under irradiation. This reduction of contact force might cause the grid to rod fretting wear. The fretting failure of the fuel rod is one of the significant issues recently in the nuclear industry from an economical as well as a safety concern. Thus, it is important to understand the characteristics of cell spring behavior for an irradiated spacer grid. In the present study, the stiffness test and dimensional measurement of cell springs were conducted to investigate the deformation behavior of cell springs of an irradiated spacer grid in a hot cell at IMEF (irradiated materials examination facility) of KAERI

  6. Basic properties of fuel determining its behavior under irradiation

    International Nuclear Information System (INIS)

    Konovalov, I.I.

    2000-01-01

    The theoretical model describing a swelling of nuclear fuel at low irradiation temperatures is considered. The critical physical parameters of substances determining behavior of point defects, gas fission atoms, dislocation density, nucleation and growth of gas-contained pores are determined. The correlation between meanings of critical parameters and physical properties of substance is offered. The accounts of swelling of various dense fuels with reference to work in conditions of research reactors are given. (author)

  7. The behavior of ZrO2/20%Y2O3 and Al2O3 coatings deposited on aluminum alloys at high temperature regime

    Science.gov (United States)

    Pintilei, G. L.; Crismaru, V. I.; Abrudeanu, M.; Munteanu, C.; Baciu, E. R.; Istrate, B.; Basescu, N.

    2015-10-01

    Aluminum alloy present numerous advantages like lightness, high specific strength and diversity which recommend them to a high number of applications from different fields. In extreme environments the protection of aluminum alloys is difficult and requires a high number of requirements like high temperature resistance, thermal fatigue resistance, corrosion fatigue resistance and galvanic corrosion resistance. To obtain these characteristics coatings can be applied to the surfaces so they can enhance the mechanical and chemical properties of the parts. In this paper two coatings were considered for deposition on an AA2024 aluminum alloy, ZrO2/20%Y2O3 and Al2O3. To obtain a better adherence of the coating to the base material an additional bond layer of NiCr is used. Both the coatings and bond layer were deposited by atmospheric plasma spraying on the samples. The samples were subjected to a temperature of 500 °C and after that slowly cooled to room temperature. The samples were analyzed by electron microscopy and X-ray diffraction to determine the morphological and phase changes that occurred during the temperature exposure. To determine the stress level in the parts due to thermal expansion a finite element analysis was performed in the same conditions as the tests.

  8. Irradiation behavior of experimental miniature uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

  9. Advanced High Temperature Structural Seals

    Science.gov (United States)

    Newquist, Charles W.; Verzemnieks, Juris; Keller, Peter C.; Rorabaugh, Michael; Shorey, Mark

    2002-10-01

    This program addresses the development of high temperature structural seals for control surfaces for a new generation of small reusable launch vehicles. Successful development will contribute significantly to the mission goal of reducing launch cost for small, 200 to 300 pound payloads. Development of high temperature seals is mission enabling. For instance, ineffective control surface seals can result in high temperature (3100 F) flows in the elevon area exceeding structural material limits. Longer sealing life will allow use for many missions before replacement, contributing to the reduction of hardware, operation and launch costs.

  10. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-01-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  11. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Roychowdhury, S., E-mail: sroy27@gmail.com [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland); Materials Processing & Corrosion Engineering Division, Mod-Lab, D-Block, Bhabha Atomic Research Centre, Mumbai 400085 (India); Seifert, H.-P.; Spätig, P.; Que, Z. [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland)

    2016-09-15

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  12. Microstructure and wear behaviors of laser clad NiCr/Cr3C2-WS2 high temperature self-lubricating wear-resistant composite coating

    Science.gov (United States)

    Yang, Mao-Sheng; Liu, Xiu-Bo; Fan, Ji-Wei; He, Xiang-Ming; Shi, Shi-Hong; Fu, Ge-Yan; Wang, Ming-Di; Chen, Shu-Fa

    2012-02-01

    The high temperature self-lubricating wear-resistant NiCr/Cr3C2-30%WS2 coating and wear-resistant NiCr/Cr3C2 coating were fabricated on 0Cr18Ni9 austenitic stainless steel by laser cladding. Phase constitutions and microstructures were investigated, and the tribological properties were evaluated using a ball-on-disc wear tester under dry sliding condition at room-temperature (17 °C), 300 °C and 600 °C, respectively. Results indicated that the laser clad NiCr/Cr3C2 coating consisted of Cr7C3 primary phase and γ-(Fe,Ni)/Cr7C3 eutectic colony, while the coating added with WS2 was mainly composed of Cr7C3 and (Cr,W)C carbides, with the lubricating WS2 and CrS sulfides as the minor phases. The wear tests showed that the friction coefficients of two coatings both decrease with the increasing temperature, while the both wear rates increase. The friction coefficient of laser clad NiCr/Cr3C2-30%WS2 is lower than the coating without WS2 whatever at room-temperature, 300 °C, 600 °C, but its wear rate is only lower at 300 °C. It is considered that the laser clad NiCr/Cr3C2-30%WS2 composite coating has good combination of anti-wear and friction-reducing capabilities at room-temperature up to 300 °C.

  13. Corrosion behavior of Al-Fe-sputtering-coated steel, high chromium steels, refractory metals and ceramics in high temperature Pb-Bi

    International Nuclear Information System (INIS)

    Abu Khalid, Rivai; Minoru, Takahashi

    2007-01-01

    Corrosion tests of Al-Fe-coated steel, high chromium steels, refractory metals and ceramics were carried out in high temperature Pb-Bi at 700 C degrees. Oxygen concentrations in this experiment were 6.8*10 -7 wt.% for Al-Fe-coated steels and 5*10 -6 wt.% for high chromium steels, refractory metals and ceramics. All specimens were immersed in molten Pb-Bi in a corrosion test pot for 1.000 hours. Coating was done with using the unbalanced magnetron sputtering (UBMS) technique to protect the steel from corrosion. Sputtering targets were Al and SUS-304. Al-Fe alloy was coated on STBA26 samples. The Al-Fe alloy-coated layer could be a good protection layer on the surface of steel. The whole of the Al-Fe-coated layer still remained on the base surface of specimen. No penetration of Pb-Bi into this layer and the matrix of the specimen. For high chromium steels i.e. SUS430 and Recloy10, the oxide layer formed in the early time could not prevent the penetration of Pb-Bi into the base of the steels. Refractory metals of tungsten (W) and molybdenum (Mo) had high corrosion resistance with no penetration of Pb-Bi into their matrix. Penetration of Pb-Bi into the matrix of niobium (Nb) was observed. Ceramic materials were SiC and Ti 3 SiC 2 . The ceramic materials of SiC and Ti 3 SiC 2 had high corrosion resistance with no penetration of Pb-Bi into their matrix. (authors)

  14. Thermal behavior of spatial structures under solar irradiation

    International Nuclear Information System (INIS)

    Liu, Hongbo; Liao, Xiangwei; Chen, Zhihua; Zhang, Qian

    2015-01-01

    The temperature, particularly the non-uniform temperature under solar irradiation, is the main load for large-span steel structures. Due the shortage of in-site temperature test in previous studies, an in-site test was conducted on the large-span steel structures under solar irradiation, which was covered by glass roof and light roof, to gain insight into the temperature distribution of steel members under glass roof or light roof. A numerical method also was presented and verified to forecast the temperature of steel member under glass roof or light roof. Based on the on-site measurement and numerical analyses conducted, the following conclusions were obtained: 1) a remarkable temperature difference exists between the steel member under glass roof and that under light roof, 2) solar irradiation has a significant effect on the temperature distribution and thermal behavior of large-span spatial structures, 3) negative thermal load is the controlling factor for member stress, and the positive thermal load is the controlling factor for nodal displacement. - Highlights: • Temperature was measured for a steel structures under glass roof and light roof. • Temperature simulation method was presented and verified. • The thermal behavior of steel structures under glass or light roof was presented

  15. Anatomic defects and behavioral abnormalities in rats irradiated in utero

    International Nuclear Information System (INIS)

    Kimler, B.F.; Norton, S.

    1987-01-01

    Pregnant rats were irradiated with 1.0 Gy whole-body doses of Cs-137 γ-rays on gestational days 11, 13, 15 and 17. Postnatal growth and preweaning behavior of the offspring were monitored prior to sacrifice or post-partuition day 28. Brain (sensory motor cortex) and pituitary tissues were processed for histological evaluation and morphometric analysis. The gestational days on which irradiation produced significant (rho<0.05) changes relative to controls are enclosed in parentheses, with the day(s) on which irradiation produced the maximum effect being underlined for the various parameters: body weight on post-partuition day 7, pituitary nuclear area, percent acidophils, and percent vacuolization, thickness of cortical layer I, II, III, IV, V, VI, and total cortical thickness; negative geotaxis, reflex suspension, continuous corridor activity, and gait. These data indicate that the critical period of development for radiation-induced alterations in post-natal growth, development, and behavior changes from the pituitary at gestational day 11 to the brain (primitive cortex) at days 13 to 17 with a peak of sensitivity at day 15

  16. SQUID picovoltometry of single crystal Bi2Sr2CaCu2O(8+delta) - Observation of the crossover from high-temperature Arrhenius to low-temperature vortex-glass behavior

    Science.gov (United States)

    Safar, H.; Gammel, P. L.; Bishop, D. J.; Mitzi, D. B.; Kapitulnik, A.

    1992-04-01

    A SQUID voltmeter has been used to measure current-voltage curves in untwinned crystals of Bi2Sr2CaCu2O(8+delta) as a function of temperature and magnetic field. The data show a clear crossover from high-temperature Arrhenius behavior to a critical region associated with the low-temperature three-dimensional vortex-glass phase transition. The critical exponents v(z - 1) = 7 +/- 1 in this system are in accord with theoretical models and previous measurements in YBa2Cu3O7. The width of the critical region collapses below 2 T, reflecting the changing role of dimensionality with field.

  17. High temperature superconductor accelerator magnets

    NARCIS (Netherlands)

    van Nugteren, J.

    2016-01-01

    For future particle accelerators bending dipoles are considered with magnetic fields exceeding 20T. This can only be achieved using high temperature superconductors (HTS). These exhibit different properties from classical low temperature superconductors and still require significant research and

  18. High Temperature Materials Laboratory (HTML)

    Data.gov (United States)

    Federal Laboratory Consortium — The six user centers in the High Temperature Materials Laboratory (HTML), a DOE User Facility, are dedicated to solving materials problems that limit the efficiency...

  19. Computer simulation of defect behavior under fusion irradiation environments

    International Nuclear Information System (INIS)

    Muroga, T.; Ishino, S.

    1983-01-01

    To simulate defect behavior under irradiation, three kinds of cascade-annealing calculations have been carried out in alpha-iron using the codes MARLOWE, DAIQUIRI and their modifications. They are (1) cascade-annealing calculation with different masses of projectile, (2) defect drifting near dislocations after cascade production and (3) cascade-overlap calculation. The defect survival ratio is found to increase as decreasing mass of the projectile both after athermal close-pair recombination and after thermal annealing. It is shown that at moderate temperatures vacancy clustering is enhanced near dislocations. Cascade-overlap is found to decrease the defect survivability. In addition, the role of helium in vacancy clustering has been calculated in aluminium lattices and its effect is found to depend strongly on temperature, interstitials and the mobility of small clusters. These results correspond well to the experimental data and will be helpful for correlating between fusion and simulation irradiations. (orig.)

  20. MOX fuel irradiation behavior in steady state (irradiation test in HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Kohno, S; Kamimura, K [Power Reactor and Nuclear Fuel Development Corp., Naka, Ibaraki (Japan)

    1997-08-01

    Two rigs of plutonium-uranium oxide (MOX) fuel rods have been irradiated in Halden boiling water reactor (HBWR) to investigate high burnup MOX fuel behavior for thermal reactor. The objective of irradiation tests is to investigate fuel behavior as influenced by pellet shape, pellet surface treatment, pellet-cladding gap size and MOX fuel powder preparations process. The two rigs have instrumentations for in-pile measurements of the fuel center-line temperature, plenum pressure, cladding elongation and fuel stack length change. The data, taken through in-operation instrumentation, have been analysed and compared with those from post-irradiation examination. The following observations are made: 1) PNC MOX fuels have achieved high burn-up as 59GWd/tMOX (67GWd/tM) at pellet peak without failure; 2) there was no significant difference in fission gas release fraction between PNC MOX fuels and UO{sub 2} fuels; 3) fission gas release from the co-converted fuel was lower than that from the mechanically blended fuel; 4) gap conductance was evaluated to decrease gradually with burn-up and to get stable in high burn-up region. 5) no evident difference of onset LHR for PCMI in experimental parameters (pellet shape and pellet-cladding gap size) was observed, but it decreased with burn-up. (author). 13 refs, 15 figs, 3 tabs.

  1. Improving High-Temperature Tensile and Low-Cycle Fatigue Behavior of Al-Si-Cu-Mg Alloys Through Micro-additions of Ti, V, and Zr

    Science.gov (United States)

    Shaha, S. K.; Czerwinski, F.; Kasprzak, W.; Friedman, J.; Chen, D. L.

    2015-07-01

    High-temperature tensile and low-cycle fatigue tests were performed to assess the influence of micro-additions of Ti, V, and Zr on the improvement of the Al-7Si-1Cu-0.5Mg (wt pct) alloy in the as-cast condition. Addition of transition metals led to modification of microstructure where in addition to conventional phases present in the Al-7Si-1Cu-0.5Mg base, new thermally stable micro-sized Zr-Ti-V-rich phases Al21.4Si4.1Ti3.5VZr3.9, Al6.7Si1.2TiZr1.8, Al2.8Si3.8V1.6Zr, and Al5.1Si35.4Ti1.6Zr5.7Fe were formed. The tensile tests showed that with increasing test temperature from 298 K to 673 K (25 °C to 400 °C), the yield stress and tensile strength of the present studied alloy decreased from 161 to 84 MPa and from 261 to 102 MPa, respectively. Also, the studied alloy exhibited 18, 12, and 5 pct higher tensile strength than the alloy A356, 354 and existing Al-Si-Cu-Mg alloy modified with additions of Zr, Ti, and Ni, respectively. The fatigue life of the studied alloy was substantially longer than those of the reference alloys A356 and the same Al-7Si-1Cu-0.5Mg base with minor additions of V, Zr, and Ti in the T6 condition. Fractographic analysis after tensile tests revealed that at the lower temperature up to 473 K (200 °C), the cleavage-type brittle fracture for the precipitates and ductile fracture for the matrix were dominant while at higher temperature fully ductile-type fracture with debonding and pull-out of cracked particles was identified. It is believed that the intermetallic precipitates containing Zr, Ti, and V improve the alloy performance at increased temperatures.

  2. On the electrical contact and long-term behavior of compression-type connections with conventional and high-temperature conductor ropes with low sag

    International Nuclear Information System (INIS)

    Hildmann, Christian

    2016-01-01

    In Germany and in Europe it is due to the ''Energiewende'' necessary to transmit more electrical energy with existing overhead transmission lines. One possible technical solution to reach this aim is the use of high temperature low sag conductors (HTLS-conductors). Compared to the common Aluminium Conductor Steel Reinforced (ACSR), HTLS-conductors have higher rated currents and rated temperatures. Thus the electrical connections for HTLS-conductors are stressed to higher temperatures too. These components are most important for the safe and reliable operation of an overhead transmission line. Besides other connection technologies, hexagonal compression connections with ordinary transmission line conductors have proven themselves since decades. From the literature, mostly empirical studies with electrical tests for compression connections are known. The electrical contact behaviour, i.e. the quality of the electrical contact after assembly, of these connections has been investigated insufficiently. This work presents and enhances an electrical model of compression connections, so that the electrical contact behaviour can be determined more accurate. Based on this, principal considerations on the current distribution in the compression connection and its influence on the connection resistance are presented. As a result from the theoretical and the experimental work, recommendations for the design of hexagonal compression connections for transmission line conductors were developed. Furthermore it is known from the functional principle of compression type connections, that the electrical contact behaviour can be influenced from their form fit, force fit and cold welding. In particular the forces in compression connections have been calculated up to now by approximation. The known analytical calculations simplify the geometry and material behaviour and do not consider the correct mechanical load during assembly. For these reasons the joining process

  3. Probing the thermal decomposition behaviors of ultrathin HfO2 films by an in situ high temperature scanning tunneling microscope.

    Science.gov (United States)

    Xue, Kun; Wang, Lei; An, Jin; Xu, Jianbin

    2011-05-13

    The thermal decomposition of ultrathin HfO(2) films (∼0.6-1.2 nm) on Si by ultrahigh vacuum annealing (25-800 °C) is investigated in situ in real time by scanning tunneling microscopy. Two distinct thickness-dependent decomposition behaviors are observed. When the HfO(2) thickness is ∼ 0.6 nm, no discernible morphological changes are found below ∼ 700 °C. Then an abrupt reaction occurs at 750 °C with crystalline hafnium silicide nanostructures formed instantaneously. However, when the thickness is about 1.2 nm, the decomposition proceeds gradually with the creation and growth of two-dimensional voids at 800 °C. The observed thickness-dependent behavior is closely related to the SiO desorption, which is believed to be the rate-limiting step of the decomposition process.

  4. High temperature corrosion of metals

    International Nuclear Information System (INIS)

    Quadakkers, W.J.; Schuster, H.; Ennis, P.J.

    1988-08-01

    This paper covers three main topics: 1. high temperature oxidation of metals and alloys, 2. corrosion in sulfur containing environments and 3. structural changes caused by corrosion. The following 21 subjects are discussed: Influence of implanted yttrium and lanthanum on the oxidation behaviour of beta-NiA1; influence of reactive elements on the adherence and protective properties of alumina scales; problems related to the application of very fine markers in studying the mechanism of thin scale formation; oxidation behaviour of chromia forming Co-Cr-Al alloys with or without reactive element additions; growth and properties of chromia-scales on high-temperature alloys; quantification of the depletion zone in high temperature alloys after oxidation in process gas; effects of HC1 and of N2 in the oxidation of Fe-20Cr; investigation under nuclear safety aspects of Zircaloy-4 oxidation kinetics at high temperatures in air; on the sulfide corrosion of metallic materials; high temperature sulfide corrosion of Mn, Nb and Nb-Si alloys; corrosion behaviour or NiCrAl-based alloys in air and air-SO2 gas mixtures; sulfidation of cobalt at high temperatures; preoxidation for sulfidation protection; fireside corrosion and application of additives in electric utility boilers; transport properties of scales with complex defect structures; observations of whiskers and pyramids during high temperature corrosion of iron in SO2; corrosion and creep of alloy 800H under simulated coal gasification conditions; microstructural changes of HK 40 cast alloy caused by exploitation in tubes in steam reformer installation; microstructural changes during exposure in corrosive environments and their effect on mechanical properties; coatings against carburization; mathematical modeling of carbon diffusion and carbide precipitation in Ni-Cr-based alloys. (MM)

  5. Effect of heat-treatment on microstructure and high-temperature deformation behavior of a low rhenium-containing single crystal nickel-based superalloy

    International Nuclear Information System (INIS)

    Sun, Nairong; Zhang, Lanting; Li, Zhigang; Shan, Aidang

    2014-01-01

    A low rhenium-containing [001] oriented single crystal nickel-based superalloy with different γ′ morphologies induced by various aging treatments was compressed from room temperature to 1000 °C. All the single crystal samples with different γ′ morphologies exhibit anomalous yield behavior. The sample first aged at 1180 °C has the widest anomalous temperature domain and highest yield strengths. The sample first aged at 1000 °C has the highest anomalous peak stress temperature

  6. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

    2005-01-01

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  7. Normal and refractory concretes for LMFBR applications. Volume 1. Review of literature on high-temperature behavior of portland cement and refractory concretes. Final report

    International Nuclear Information System (INIS)

    Bazant, Z.P.; Chern, J.C.; Abrams, M.S.; Gillen, M.P.

    1982-06-01

    The extensive literature on the properties and behavior at elevated temperature of portland cement concrete and various refractory concretes was reviewed to collect in concise form the physical and chemical properties of castable refractory concretes and of conventional portland cement concretes at elevated temperature. This survey, together with an extensive bibliography of source documents, is presented in Volume 1. A comparison was made of these properties, the relative advantages of the various concretes was evaluated for possible liquid metal fast breeder reactor applications, and a selection was made of several materials of interest for such applications. Volume 2 concludes with a summary of additional knowledge needed to support such uses of these materials together with recommendations on research to provide that knowledge

  8. Highly efficient high temperature electrolysis

    DEFF Research Database (Denmark)

    Hauch, Anne; Ebbesen, Sune; Jensen, Søren Højgaard

    2008-01-01

    High temperature electrolysis of water and steam may provide an efficient, cost effective and environmentally friendly production of H-2 Using electricity produced from sustainable, non-fossil energy sources. To achieve cost competitive electrolysis cells that are both high performing i.e. minimum...... internal resistance of the cell, and long-term stable, it is critical to develop electrode materials that are optimal for steam electrolysis. In this article electrolysis cells for electrolysis of water or steam at temperatures above 200 degrees C for production of H-2 are reviewed. High temperature...... electrolysis is favourable from a thermodynamic point of view, because a part of the required energy can be supplied as thermal heat, and the activation barrier is lowered increasing the H-2 production rate. Only two types of cells operating at high temperature (above 200 degrees C) have been described...

  9. Computer code validation by high temperature chemistry

    International Nuclear Information System (INIS)

    Alexander, C.A.; Ogden, J.S.

    1988-01-01

    At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered

  10. High-temperature metallography setup

    International Nuclear Information System (INIS)

    Blumenfeld, M.; Shmarjahu, D.; Elfassy, S.

    1979-06-01

    A high-temperature metallography setup is presented. In this setup the observation of processes such as that of copper recrystallization was made possible, and the structure of metals such as uranium could be revealed. A brief historical review of part of the research works that have been done with the help of high temperature metallographical observation technique since the beginning of this century is included. Detailed description of metallographical specimen preparation technique and theoretical criteria based on the rate of evaporation of materials present on the polished surface of the specimens are given

  11. High temperature corrosion in gasifiers

    Directory of Open Access Journals (Sweden)

    Bakker Wate

    2004-01-01

    Full Text Available Several commercial scale coal gasification combined cycle power plants have been built and successfully operated during the last 5-10 years. Supporting research on materials of construction has been carried out for the last 20 years by EPRI and others. Emphasis was on metallic alloys for heat exchangers and other components in contact with hot corrosive gases at high temperatures. In this paper major high temperature corrosion mechanisms, materials performance in presently operating gasifiers and future research needs will be discussed.

  12. High temperature materials; Materiaux a hautes temperatures

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    The aim of this workshop is to share the needs of high temperature and nuclear fuel materials for future nuclear systems, to take stock of the status of researches in this domain and to propose some cooperation works between the different research organisations. The future nuclear systems are the very high temperature (850 to 1200 deg. C) gas cooled reactors (GCR) and the molten salt reactors (MSR). These systems include not only the reactor but also the fabrication and reprocessing of the spent fuel. This document brings together the transparencies of 13 communications among the 25 given at the workshop: 1) characteristics and needs of future systems: specifications, materials and fuel needs for fast spectrum GCR and very high temperature GCR; 2) high temperature materials out of neutron flux: thermal barriers: materials, resistance, lifetimes; nickel-base metal alloys: status of knowledge, mechanical behaviour, possible applications; corrosion linked with the gas coolant: knowledge and problems to be solved; super-alloys for turbines: alloys for blades and discs; corrosion linked with MSR: knowledge and problems to be solved; 3) materials for reactor core structure: nuclear graphite and carbon; fuel assembly structure materials of the GCR with fast neutron spectrum: status of knowledge and ceramics and cermets needs; silicon carbide as fuel confinement material, study of irradiation induced defects; migration of fission products, I and Cs in SiC; 4) materials for hydrogen production: status of the knowledge and needs for the thermochemical cycle; 5) technologies: GCR components and the associated material needs: compact exchangers, pumps, turbines; MSR components: valves, exchangers, pumps. (J.S.)

  13. Effect of oxidation on the fatigue crack propagation behavior of Z3CN20.09M dyplex stainless steel in high temperature water

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Huan Chun; Yang, Bin [State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, Beijing (China); Chen, Yue Feng; Chen, Xu Dong [Collaborative Innovation Center of Steel Technology, Beijing (China)

    2017-06-15

    The fatigue crack propagation behaviors of Z3CN20.09M duplex stainless steel (DSS) were investigated by studying oxide films of specimens tested in 290°C water and air. The results indicate that a full oxide film that consisted of oxides and hydroxides was formed in 290°C water. By contrast, only a half-baked oxide film consisting of oxides was formed in 290°C air. Both environments are able to deteriorate the elastic modulus and hardness of the oxide films, especially the 290°C water. The fatigue lives of the specimens tested in 290°C air were about twice of those tested in 290°C water at all strain amplitudes. Moreover, the crack propagation rates of the specimen tested in 290°C water were confirmed to be faster than those tested in 290°C air, which was thought to be due to the deteriorative strength of the oxide films induced by the mutual promotion of oxidation and crack propagation at the crack tip. It is noteworthy that the crack propagation can be postponed by the ferrite phase in the DSS, especially when the specimens were tested in 290°C water.

  14. High-temperature plasma physics

    International Nuclear Information System (INIS)

    Furth, H.P.

    1988-03-01

    Both magnetic and inertial confinement research are entering the plasma parameter range of fusion reactor interest. This paper reviews the individual and common technical problems of these two approaches to the generation of thermonuclear plasmas, and describes some related applications of high-temperature plasma physics

  15. High Temperature Superconductor Machine Prototype

    DEFF Research Database (Denmark)

    Mijatovic, Nenad; Jensen, Bogi Bech; Træholt, Chresten

    2011-01-01

    A versatile testing platform for a High Temperature Superconductor (HTS) machine has been constructed. The stationary HTS field winding can carry up to 10 coils and it is operated at a temperature of 77K. The rotating armature is at room temperature. Test results and performance for the HTS field...

  16. High Temperature Transparent Furnace Development

    Science.gov (United States)

    Bates, Stephen C.

    1997-01-01

    This report describes the use of novel techniques for heat containment that could be used to build a high temperature transparent furnace. The primary objective of the work was to experimentally demonstrate transparent furnace operation at 1200 C. Secondary objectives were to understand furnace operation and furnace component specification to enable the design and construction of a low power prototype furnace for delivery to NASA in a follow-up project. The basic approach of the research was to couple high temperature component design with simple concept demonstration experiments that modify a commercially available transparent furnace rated at lower temperature. A detailed energy balance of the operating transparent furnace was performed, calculating heat losses through the furnace components as a result of conduction, radiation, and convection. The transparent furnace shells and furnace components were redesigned to permit furnace operation at at least 1200 C. Techniques were developed that are expected to lead to significantly improved heat containment compared with current transparent furnaces. The design of a thermal profile in a multizone high temperature transparent furnace design was also addressed. Experiments were performed to verify the energy balance analysis, to demonstrate some of the major furnace improvement techniques developed, and to demonstrate the overall feasibility of a high temperature transparent furnace. The important objective of the research was achieved: to demonstrate the feasibility of operating a transparent furnace at 1200 C.

  17. High temperature electronic gain device

    International Nuclear Information System (INIS)

    McCormick, J.B.; Depp, S.W.; Hamilton, D.J.; Kerwin, W.J.

    1979-01-01

    An integrated thermionic device suitable for use in high temperature, high radiation environments is described. Cathode and control electrodes are deposited on a first substrate facing an anode on a second substrate. The substrates are sealed to a refractory wall and evacuated to form an integrated triode vacuum tube

  18. Containment of high temperature plasmas

    International Nuclear Information System (INIS)

    Bass, R.W.; Ferguson, H.R.P.; Fletcher, H. Jr.; Gardner, J.; Harrison, B.K.; Larsen, K.M.

    1973-01-01

    Apparatus is described for confining a high temperature plasma which comprises: 1) envelope means shaped to form a toroidal hollow chamber containing a plasma, 2) magnetic field line generating means for confining the plasma in a smooth toroidal shape without cusps. (R.L.)

  19. Chemistry of high temperature superconductors

    CERN Document Server

    1991-01-01

    This review volume contains the most up-to-date articles on the chemical aspects of high temperature oxide superconductors. These articles are written by some of the leading scientists in the field and includes a comprehensive list of references. This is an essential volume for researchers working in the fields of ceramics, materials science and chemistry.

  20. Properties of high temperature SQUIDS

    International Nuclear Information System (INIS)

    Falco, C.M.; Wu, C.T.

    1978-01-01

    A review is given of the present status of weak links and dc and rf biased SQUIDs made with high temperature superconductors. A method for producing reliable, reproducible devices using Nb 3 Sn is outlined, and comments are made on directions future work should take

  1. High temperature component life assessment

    CERN Document Server

    Webster, G A

    1994-01-01

    The aim of this book is to investigate and explain the rapid advances in the characterization of high temperature crack growth behaviour which have been made in recent years, with reference to industrial applications. Complicated mathematics has been minimized with the emphasis placed instead on finding solutions using simplified procedures without the need for complex numerical analysis.

  2. 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

  3. A Comparative Study on Johnson Cook, Modified Zerilli-Armstrong and Arrhenius-Type Constitutive Models to Predict High-Temperature Flow Behavior of Ti-6Al-4V Alloy in α + β Phase

    Science.gov (United States)

    Cai, Jun; Wang, Kuaishe; Han, Yingying

    2016-03-01

    True stress and true strain values obtained from isothermal compression tests over a wide temperature range from 1,073 to 1,323 K and a strain rate range from 0.001 to 1 s-1 were employed to establish the constitutive equations based on Johnson Cook, modified Zerilli-Armstrong (ZA) and strain-compensated Arrhenius-type models, respectively, to predict the high-temperature flow behavior of Ti-6Al-4V alloy in α + β phase. Furthermore, a comparative study has been made on the capability of the three models to represent the elevated temperature flow behavior of Ti-6Al-4V alloy. Suitability of the three models was evaluated by comparing both the correlation coefficient R and the average absolute relative error (AARE). The results showed that the Johnson Cook model is inadequate to provide good description of flow behavior of Ti-6Al-4V alloy in α + β phase domain, while the predicted values of modified ZA model and the strain-compensated Arrhenius-type model could agree well with the experimental values except under some deformation conditions. Meanwhile, the modified ZA model could track the deformation behavior more accurately than other model throughout the entire temperature and strain rate range.

  4. High Temperature Operational Experiences of Helium Experimental Loop

    International Nuclear Information System (INIS)

    Kim, Chan Soo; Hong, Sung-Deok; Kim, Eung-Seon; Kim, Min Hwan

    2015-01-01

    The development of high temperature components of VHTR is very important because of its higher operation temperature than that of a common light water reactor and high pressure industrial process. The development of high temperature components requires the large helium loop. Many countries have high temperature helium loops or a plan for its construction. Table 1 shows various international state-of-the-art of high temperature and high pressure gas loops. HELP performance test results show that there is no problem in operation of HELP at the very high temperature experimental condition. These experimental results also provide the basic information for very high temperature operation with bench-scale intermediate heat exchanger prototype in HELP. In the future, various heat exchanger tests will give us the experimental data for GAMMA+ validation about transient T/H behavior of the IHX prototype and the optimization of the working fluid in the intermediate loop

  5. Behavior of solid fission products in irradiated fuel

    International Nuclear Information System (INIS)

    Song, Ung Sup; Jung, Yang Hong; Kim, Hee Moon; Yoo, Byun Gok; Kim, Do Sik; Choo, Yong Sun; Hong, Kwon Pyo

    2004-01-01

    Many fission products are generated by fission events in UO 2 fuel under irradiation in nuclear reactor. Concentration of each fission product is changed by conditions of neutron energy spectrum, fissile material, critical thermal power, irradiation period and cooling time. Volatile materials such as Cs and I, the fission products, degrade nuclear fuel rod by the decrease of thermal conductivity in pellet and the stress corrosion cracking in cladding. Metal fission products (white inclusion) make pellet be swelled and decrease volume of pellet by densification. It seems that metal fission products are filled in the pore in pellet and placed between UO 2 lattices as interstitial. In addition, metal oxide state may change structural lattice volume. Considering behavior of fission products mentioned above, concentration of them is important. Fission products could be classified as bellows; solid solution in matrix : Sr, Zr, Nb, Y, La, Ce, Pr, Nd, Pm, Sm - metal precipitates : Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sb, Te - oxide precipitates : Ba, Zr, Nb, Mo, (Rb, Cs, Te) - volatile and gases : Kr, Xe, Br, I, (Rb, Cs, Te)

  6. On transient irradiation behavior of HTGR fuel particles

    International Nuclear Information System (INIS)

    Mortenson, S.C.; Okrent, D.

    1977-01-01

    An examination of HTGR TRISO coated fuel particles was made in which the particles' stress-strain histories were determined during both steady-state and transient operating conditions. The basis for the examination was a modified version of a computer code written by Kaae which assumed spherical symmetry, isotropic thermal expansion, isotropic elastic constants, time-temperature-irradiation invariant materials properties, and steady state operation during particle exposure. Additionally, the Kaae code modelled potential separation of layers at the SiC-inner PyC interface and considered that several entrapped fission products could exist in either the gaseous or solid state, dependent upon particle operating conditions. Using the modified code which modelled transient behavior in a quasi-static fashion, a series of both steady-state and transient operating condition computer simulations was made. For the former set of runs, a candidate set of particle dimensions and a nominal set of materials' properties was assumed. Layer thicknesses were assumed to be normally distributed about the nominal thickenesses and a probability distribution of SiC tensile stresses was generated; sensitivity of the stress distribution to assumed standard deviation of the layer thicknesses was acute. Further, this series of steady-state runs demonstrated that for certain combinations of the assumed PyC-SiC bond interface strength and irradiation-induced creep constant, anomalous predicted stresses may be obtained in the PyC layers. The steady-state runs also suggest that transient behavior would most likely not be significant at fast neutron exposures below about 10 21 NVT due to both low fission gas pressure and likely beneficial interface separation

  7. Ceramics for high temperature applications

    International Nuclear Information System (INIS)

    Mocellin, A.

    1977-01-01

    Problems related to materials, their fabrication, properties, handling, improvements are examined. Silicium nitride and silicium carbide are obtained by vacuum hot-pressing, reaction sintering and chemical vapour deposition. Micrographs are shown. Mechanical properties i.e. room and high temperature strength, creep resistance fracture mechanics and fatigue resistance. Recent developments of pressureless sintered Si C and the Si-Al-O-N quaternary system are mentioned

  8. High-temperature geothermal cableheads

    Science.gov (United States)

    Coquat, J. A.; Eifert, R. W.

    1981-11-01

    Two high temperature, corrosion resistant logging cable heads which use metal seals and a stable fluid to achieve proper electrical terminations and cable sonde interfacings are described. A tensile bar provides a calibrated yield point, and a cone assembly anchors the cable armor to the head. Electrical problems of the sort generally ascribable to the cable sonde interface were absent during demonstration hostile environment loggings in which these cable heads were used.

  9. Summary: High Temperature Downhole Motor

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-10-01

    Directional drilling can be used to enable multi-lateral completions from a single well pad to improve well productivity and decrease environmental impact. Downhole rotation is typically developed with a motor in the Bottom Hole Assembly (BHA) that develops drilling power (speed and torque) necessary to drive rock reduction mechanisms (i.e., the bit) apart from the rotation developed by the surface rig. Historically, wellbore deviation has been introduced by a “bent-sub,” located in the BHA, that introduces a small angular deviation, typically less than 3 degrees, to allow the bit to drill off-axis with orientation of the BHA controlled at the surface. The development of a high temperature downhole motor would allow reliable use of bent subs for geothermal directional drilling. Sandia National Laboratories is pursuing the development of a high temperature motor that will operate on either drilling fluid (water-based mud) or compressed air to enable drilling high temperature, high strength, fractured rock. The project consists of designing a power section based upon geothermal drilling requirements; modeling and analysis of potential solutions; and design, development and testing of prototype hardware to validate the concept. Drilling costs contribute substantially to geothermal electricity production costs. The present development will result in more reliable access to deep, hot geothermal resources and allow preferential wellbore trajectories to be achieved. This will enable development of geothermal wells with multi-lateral completions resulting in improved geothermal resource recovery, decreased environmental impact and enhanced well construction economics.

  10. High temperature fusion reactor design

    International Nuclear Information System (INIS)

    Harkness, S.D.; dePaz, J.F.; Gohar, M.Y.; Stevens, H.C.

    1979-01-01

    Fusion energy may have unique advantages over other systems as a source for high temperature process heat. A conceptual design of a blanket for a 7 m tokamak reactor has been developed that is capable of producing 1100 0 C process heat at a pressure of approximately 10 atmospheres. The design is based on the use of a falling bed of MgO spheres as the high temperature heat transfer system. By preheating the spheres with energy taken from the low temperature tritium breeding part of the blanket, 1086 MW of energy can be generated at 1100 0 C from a system that produces 3000 MW of total energy while sustaining a tritium breeding ratio of 1.07. The tritium breeding is accomplished using Li 2 O modules both in front of (6 cm thick) and behind (50 cm thick) the high temperature ducts. Steam is used as the first wall and front tritium breeding module coolant while helium is used in the rear tritium breeding region. The system produces 600 MW of net electricity for use on the grid

  11. Fatigue behavior of Type 316 stainless steel following neutron irradiation inducing helium

    International Nuclear Information System (INIS)

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

    1980-01-01

    Since a tokamak reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied

  12. High Temperature Fluoride Salt Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Aaron, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cunningham, Richard Burns [Univ. of Tennessee, Knoxville, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peretz, Fred J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yoder, Jr, Graydon L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

  13. Fibras de polipropileno e sua influência no comportamento de concretos expostos a altas temperaturas: revisão Polypropylene fibers and their influence on the behavior of concretes exposed to high temperatures: review

    Directory of Open Access Journals (Sweden)

    A. L. de Castro

    2011-03-01

    before fire; however, in practice the stability of this material is reduced by high temperatures. Unfortunately, under such circumstances, concrete elements present excessive damages or even catastrophic failures. When exposed to high temperatures, cement based materials undergo physicochemical changes that damage their mechanical properties and spoil their resistance to heat transfer. Although the thermal features of a high strength concrete are similar to those of a conventional concrete, this material has a greater sensibility to high temperatures due to its reduced porosity, showing a higher relative loss of the mechanical properties and explosive spalling in the temperature range between 100 ºC and 400 ºC. The spalling can be avoided by adding polypropylene fibers in concrete: when melted and partially absorbed by the cement matrix, the fibers generate a permeable network that allows the outward gas migration, decreasing the pore pressure in the material and, consequently, eliminating the possibility of explosive spalling occurrence. Thus, in the present paper, a review regarding the behavior of concretes exposed to high temperatures, as well as the influence of polypropylene fibers have been addressed for concretes applied in the civil engineering area.

  14. Irradiation performance of HTGR fertile fuel in HFIR target capsules HT-12 through HT-15. Part I. Experiment description and fission product behavior

    International Nuclear Information System (INIS)

    Kania, M.J.; Lindemer, T.B.; Morgan, M.T.; Robbins, J.M.

    1977-02-01

    Sixteen types of Biso-coated designs, on ThO 2 kernels, were irradiated in High Flux Isotope Reactor target capsules HT-12 through HT-15. The report addresses the description of the experiment and extensive postirradiation analyses and experiments to determine fertile-particle burnup, fuel coating failures, and fission product behavior. Several low-temperature isotropic (LTI) pyrocarbon coatings, which ''survived'' according to visual inspection, were shown to have developed permeability during irradiation. These particles were irradiated at temperatures approximately equal to 1250 0 C and to burnups equal to or greater than 8 percent fission per initial heavy-metal atom (FIMA). No evidence of permeability was found in similar particles irradiated at temperatures approximately equal to 1550 0 C and burnups approximately equal to 16 percent FIMA. Failures due to permeability were not detectable by visual inspection but required a more extensive investigation by the 1000 0 C gaseous chlorine leaching technique. Maximum particle surface operating temperatures were found to be approximately 300 0 C in excess of design limits of 900 0 C (low-temperature magazines) and 1250 0 C (high-temperature magazines). The extremes of high temperatures and fast neutron fluences up to 1.6 x 10 22 neutrons/cm 2 produced severe degradation and swelling of the Poco graphite magazines and sample holders

  15. High temperature embrittlement of metals by helium

    International Nuclear Information System (INIS)

    Schroeder, H.

    1983-01-01

    The present knowledge of the influence of helium on the high temperature mechanical properties of metals to be used as structural materials in fast fission and in future fusion reactors is reviewed. A wealth of experimental data has been obtained by many different experimental techniques, on many different alloys, and on different properties. This review is mostly concentrated on the behaviour of austenitic alloys -especially austenitic stainless steels, for which the data base is by far the largest - and gives only a few examples of special bcc alloys. The effect of the helium embrittlement on the different properties - tensile, fatigue and, with special emphasis, creep - is demonstrated by representative results. A comparison between data obtained from in-pile (-beam) experiments and from post-irradiation (-implantation) experiments, respectively, is presented. Theoretical models to describe the observed phenomena are briefly outlined and some suggestions are made for future work to resolve uncertainties and differences between our experimental knowledge and theoretical understanding of high temperature helium embrittlement. (author)

  16. Overall viscoplastic behavior of non-irradiated porous nuclear ceramics

    International Nuclear Information System (INIS)

    Monerie, Yann; Gatt, Jean-Marie

    2006-01-01

    This paper deals with the overall behavior of nonlinear viscous and porous nuclear ceramics. Bi-viscous isotropic porous materials are considered: the matrix is subjected to two power-law viscosities with different exponents related to two stationary temperature-activated creeping mechanisms (scattering-creep and dislocation-creep), and this matrix contains a low porosity volume fraction. The overall behavior of these types of composite materials is obtained with the help of quadratic strain-rate potentials combined with experimental-based coupling function depending on stress and temperature. For each creeping mechanism, the hollow sphere model of [Michel, J.-C., Suquet, P., 1992. The constitutive law of nonlinear viscous and porous materials. Journal of the Mechanics and Physics of Solids 40, 783-812] is used. Mechanical parameters of the resulting model are identified and validated in the particular case of non-irradiated uranium dioxide nuclear ceramics. This model predicts, under pure thermo-mechanical loading, a variation of the material volume and a variation of the porosity volume fraction (the so-called densification or swelling). (authors)

  17. Experimental needs of high temperature concrete

    International Nuclear Information System (INIS)

    Chern, J.C.; Marchertas, A.H.

    1985-01-01

    The needs of experimental data on concrete structures under high temperature, ranging up to about 370 0 C for operating reactor conditions and to about 900 0 C and beyond for hypothetical accident conditions, are described. This information is required to supplement analytical methods which are being implemented into the finite element code TEMP-STRESS to treat reinforced concrete structures. Recommended research ranges from material properties of reinforced/prestressed concrete, direct testing of analytical models used in the computer codes, to investigations of certain aspects of concrete behavior, the phenomenology of which is not well understood. 10 refs

  18. Multichannel euv spectroscopy of high temperature plasmas

    International Nuclear Information System (INIS)

    Fonck, R.J.

    1983-11-01

    Spectroscopy of magnetically confined high temperature plasmas in the visible through x-ray spectral ranges deals primarily with the study of impurity line radiation or continuum radiation. Detailed knowledge of absolute intensities, temporal behavior, and spatial distributions of the emitted radiation is desired. As tokamak facilities become more complex, larger, and less accessible, there has been an increased emphasis on developing new instrumentation to provide such information in a minimum number of discharges. The availability of spatially-imaging detectors for use in the vacuum ultraviolet region (especially the intensified photodiode array) has generated the development of a variety of multichannel spectrometers for applications on tokamak facilities

  19. Behavior of ferritic steels irradiated by fast neutrons

    International Nuclear Information System (INIS)

    Erler, Jean; Maillard, Arlette; Brun, Gilbert; Lehmann, Jeanne; Dupouy, J.-M.

    1979-01-01

    Ferritic steels were irradiated in Rapsodie and Phenix at varying doses. The swelling and irradiation creep characteristics are reported below as are the mechanical characteristics of these materials [fr

  20. FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

    Energy Technology Data Exchange (ETDEWEB)

    Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

  1. Passivation of high temperature superconductors

    Science.gov (United States)

    Vasquez, Richard P. (Inventor)

    1991-01-01

    The surface of high temperature superconductors such as YBa2Cu3O(7-x) are passivated by reacting the native Y, Ba and Cu metal ions with an anion such as sulfate or oxalate to form a surface film that is impervious to water and has a solubility in water of no more than 10(exp -3) M. The passivating treatment is preferably conducted by immersing the surface in dilute aqueous acid solution since more soluble species dissolve into the solution. The treatment does not degrade the superconducting properties of the bulk material.

  2. "Green" High-Temperature Polymers

    Science.gov (United States)

    Meador, Michael A.

    1998-01-01

    PMR-15 is a processable, high-temperature polymer developed at the NASA Lewis Research Center in the 1970's principally for aeropropulsion applications. Use of fiber-reinforced polymer matrix composites in these applications can lead to substantial weight savings, thereby leading to improved fuel economy, increased passenger and payload capacity, and better maneuverability. PMR-15 is used fairly extensively in military and commercial aircraft engines components seeing service temperatures as high as 500 F (260 C), such as the outer bypass duct for the F-404 engine. The current world-wide market for PMR-15 materials (resins, adhesives, and composites) is on the order of $6 to 10 million annually.

  3. CONFINEMENT OF HIGH TEMPERATURE PLASMA

    Science.gov (United States)

    Koenig, H.R.

    1963-05-01

    The confinement of a high temperature plasma in a stellarator in which the magnetic confinement has tended to shift the plasma from the center of the curved, U-shaped end loops is described. Magnetic means are provided for counteracting this tendency of the plasma to be shifted away from the center of the end loops, and in one embodiment this magnetic means is a longitudinally extending magnetic field such as is provided by two sets of parallel conductors bent to follow the U-shaped curvature of the end loops and energized oppositely on the inside and outside of this curvature. (AEC)

  4. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi.

    1991-02-01

    High temperature divertor plasma operation has been proposed, which is expected to enhance the core energy confinement and eliminates the heat removal problem. In this approach, the heat flux is guided through divertor channel to a remote area with a large target surface, resulting in low heat load on the target plate. This allows pumping of the particles escaping from the core and hence maintaining of the high divertor temperature, which is comparable to the core temperature. The energy confinement is then determined by the diffusion coefficient of the core plasma, which has been observed to be much lower than the thermal diffusivity. (author)

  5. High temperature superconductors and method

    International Nuclear Information System (INIS)

    Ruvalds, J.J.

    1977-01-01

    This invention comprises a superconductive compound having the formula: Ni/sub 1-x/M/sub x/Z/sub y/ wherein M is a metal which will destroy the magnetic character of nickel (preferably copper, silver or gold); Z is hydrogen or deuterium; x is 0.1 to 0.9; and y, correspondingly, 0.9 to 0.1, and method of conducting electric current with no resistance at relatively high temperature of T>1 0 K comprising a conductor consisting essentially of the superconducting compound noted above

  6. High temperature thermoelectric energy conversion

    International Nuclear Information System (INIS)

    Wood, C.

    1986-01-01

    Considerable advances were made in the late '50's and early early '60's in the theory and development of materials for high-temperature thermoelectric energy conversion. This early work culminated in a variety of materials, spanning a range of temperatures, with the product of the figure of merit, Z, and temperature, T, i.e., the dimensionless figure of merit, ZT, of the order of one. This experimental limitation appeared to be universal and led a number of investigators to explore the possibility that a ZT - also represents a theoretical limitation. It was found not to be so

  7. Sphaleron rate at high temperature in 1+1 dimensions

    International Nuclear Information System (INIS)

    Smit, Jan; Tang, W.H.

    1999-01-01

    We resolve the controversy in the high temperature behavior of the sphaleron rate in the abelian Higgs model in 1+1 dimensions. The T 2 behavior at intermediate lattice spacings is found to change into T ((2)/(3)) behavior in the continuum limit. The results are supported by analytic arguments that the classical approximation is good for this model

  8. Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2012-03-15

    The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

  9. Studies of high temperature superconductors

    International Nuclear Information System (INIS)

    Narlikar, A.

    1989-01-01

    The high temperature superconductors (HTSCs) discovered are from the family of ceramic oxides. Their large scale utilization in electrical utilities and in microelectronic devices are the frontal challenges which can perhaps be effectively met only through consolidated efforts and expertise of a multidisciplinary nature. During the last two years the growth of the new field has occurred on an international scale and perhaps has been more rapid than in most other fields. There has been an extraordinary rush of data and results which are continually being published as short texts dispersed in many excellent journals, some of which were started to ensure rapid publication exclusively in this field. As a result, the literature on HTSCs has indeed become so massive and so diffuse that it is becoming increasingly difficult to keep abreast with the important and reliable facets of this fast-growing field. This provided the motivation to evolve a process whereby both professional investigators and students can have ready access to up-to- date in-depth accounts of major technical advances happening in this field. The present series Studies of High Temperature Superconductors has been launched to, at least in part, fulfill this need

  10. High temperature PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jianlu; Xie, Zhong; Zhang, Jiujun; Tang, Yanghua; Song, Chaojie; Navessin, Titichai; Shi, Zhiqing; Song, Datong; Wang, Haijiang; Wilkinson, David P.; Liu, Zhong-Sheng; Holdcroft, Steven [Institute for Fuel Cell Innovation, National Research Council Canada, Vancouver, BC (Canada V6T 1W5)

    2006-10-06

    There are several compelling technological and commercial reasons for operating H{sub 2}/air PEM fuel cells at temperatures above 100{sup o}C. Rates of electrochemical kinetics are enhanced, water management and cooling is simplified, useful waste heat can be recovered, and lower quality reformed hydrogen may be used as the fuel. This review paper provides a concise review of high temperature PEM fuel cells (HT-PEMFCs) from the perspective of HT-specific materials, designs, and testing/diagnostics. The review describes the motivation for HT-PEMFC development, the technology gaps, and recent advances. HT-membrane development accounts for {approx}90% of the published research in the field of HT-PEMFCs. Despite this, the status of membrane development for high temperature/low humidity operation is less than satisfactory. A weakness in the development of HT-PEMFC technology is the deficiency in HT-specific fuel cell architectures, test station designs, and testing protocols, and an understanding of the underlying fundamental principles behind these areas. The development of HT-specific PEMFC designs is of key importance that may help mitigate issues of membrane dehydration and MEA degradation. (author)

  11. NSTX High Temperature Sensor Systems

    International Nuclear Information System (INIS)

    McCormack, B.; Kugel, H.W.; Goranson, P.; Kaita, R.

    1999-01-01

    The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed

  12. Can prenatal low-dose irradiation affect behavior of adult rats?

    International Nuclear Information System (INIS)

    Smajda, B.; Tomasova, L.; Kokocova, N.

    2011-01-01

    The aim of our study was to determine whether exposure of very low dose gamma-rays during the critical phase of brain development affects some selected behavioral parameters in adult rats. Pregnant female Wistar rats were irradiated with 1 Gy gamma-rays from a cobalt source at 17 th day of pregnancy. The progeniture of irradiated as well as non-irradiated females have undergone behavioral tests at the age of 3 months. Irradiated animals exhibited lower locomotor and exploratory activity in the open field test. (authors)

  13. HIgh Temperature Photocatalysis over Semiconductors

    Science.gov (United States)

    Westrich, Thomas A.

    Due in large part to in prevalence of solar energy, increasing demand of energy production (from all sources), and the uncertain future of petroleum energy feedstocks, solar energy harvesting and other photochemical systems will play a major role in the developing energy market. This dissertation focuses on a novel photochemical reaction process: high temperature photocatalysis (i.e., photocatalysis conducted above ambient temperatures, T ≥ 100°C). The overarching hypothesis of this process is that photo-generated charge carriers are able to constructively participate in thermo-catalytic chemical reactions, thereby increasing catalytic rates at one temperature, or maintaining catalytic rates at lower temperatures. The photocatalytic oxidation of carbon deposits in an operational hydrocarbon reformer is one envisioned application of high temperature photocatalysis. Carbon build-up during hydrocarbon reforming results in catalyst deactivation, in the worst cases, this was shown to happen in a period of minutes with a liquid hydrocarbon. In the presence of steam, oxygen, and above-ambient temperatures, carbonaceous deposits were photocatalytically oxidized over very long periods (t ≥ 24 hours). This initial experiment exemplified the necessity of a fundamental assessment of high temperature photocatalytic activity. Fundamental understanding of the mechanisms that affect photocatalytic activity as a function of temperatures was achieved using an ethylene photocatalytic oxidation probe reaction. Maximum ethylene photocatalytic oxidation rates were observed between 100 °C and 200 °C; the maximum photocatalytic rates were approximately a factor of 2 larger than photocatalytic rates at ambient temperatures. The loss of photocatalytic activity at temperatures above 200 °C is due to a non-radiative multi-phonon recombination mechanism. Further, it was shown that the fundamental rate of recombination (as a function of temperature) can be effectively modeled as a

  14. High-temperature absorbed dose measurements in the megagray range

    International Nuclear Information System (INIS)

    Balian, P.; Ardonceau, J.; Zuppiroli, L.

    1988-01-01

    Organic conductors of the tetraselenotetracene family have been tested as ''high-temperature'' absorbed dose dosimeters. They were heated up to 120 0 C and irradiated at this temperature with 1-MeV electrons in order to simulate, in a short time, a much longer γ-ray irradiation. The electric resistance increase of the crystal can be considered a good measurement of the absorbed dose in the range 10 6 Gy to a few 10 8 Gy and presumably one order of magnitude more. This dosimeter also permits on-line (in-situ) measurements of the absorbed dose without removing the sensor from the irradiation site. The respective advantages of organic and inorganic dosimeters at these temperature and dose ranges are also discussed. In this connection, we outline new, but negative, results concerning the possible use of silica as a high-temperature, high-dose dosimeter. (author)

  15. The high-temperature reactor

    International Nuclear Information System (INIS)

    Kirchner, U.

    1991-01-01

    The book deals with the development of the German high-temperature reactor (pebble-bed), the design of a prototype plant and its (at least provisional) shut-down in 1989. While there is a lot of material on the HTR's competitor, the fast breeder, literature is very incomplete on HTRs. The author describes HTR's history as a development which was characterised by structural divergencies but not effectively steered and monitored. There was no project-oriented 'community' such as there was for the fast breeder. Also, the new technology was difficult to control there were situations where no one quite knew what was going on. The technical conditions however were not taken as facts but as a basis for interpretation, wishes and reservations. The HTR gives an opportunity to consider the conditions under which large technical projects can be carried out today. (orig.) [de

  16. High temperature creep of vanadium

    International Nuclear Information System (INIS)

    Juhasz, A.; Kovacs, I.

    1978-01-01

    The creep behaviour of polycrystalline vanadium of 99.7% purity has been investigated in the temperature range 790-880 0 C in a high temperature microscope. It was found that the creep properties depend strongly on the history of the sample. To take this fact into account some additional properties such as the dependence of the yield stress and the microhardness on the pre-annealing treatment have also been studied. Samples used in creep measurements were selected on the basis of their microhardness. The activation energy of creep depends on the microhardness and on the creep temperature. In samples annealed at 1250 0 C for one hour (HV=160 kgf mm -2 ) the rate of creep is controlled by vacancy diffusion in the temperature range 820-880 0 C with an activation energy of 78+-8 kcal mol -1 . (Auth.)

  17. High temperature industrial heat pumps

    Energy Technology Data Exchange (ETDEWEB)

    Berghmans, J. (Louvain Univ., Heverlee (Belgium). Inst. Mechanica)

    1990-01-01

    The present report intends to describe the state of the art of high temperature industrial heat pumps. A description is given of present systems on the market. In addition the research and development efforts on this subject are described. Compression (open as well as closed cycle) systems, as well as absorption heat pumps (including transformers), are considered. This state of the art description is based upon literature studies performed by a team of researchers from the Katholieke Universiteit Leuven, Belgium. The research team also analysed the economics of heat pumps of different types under the present economic conditions. The heat pumps are compared with conventional heating systems. This analysis was performed in order to evaluate the present condition of the heat pump in the European industry.

  18. Faraday imaging at high temperatures

    Science.gov (United States)

    Hackel, Lloyd A.; Reichert, Patrick

    1997-01-01

    A Faraday filter rejects background light from self-luminous thermal objects, but transmits laser light at the passband wavelength, thus providing an ultra-narrow optical bandpass filter. The filter preserves images so a camera looking through a Faraday filter at a hot target illuminated by a laser will not see the thermal radiation but will see the laser radiation. Faraday filters are useful for monitoring or inspecting the uranium separator chamber in an atomic vapor laser isotope separation process. Other uses include viewing welds, furnaces, plasma jets, combustion chambers, and other high temperature objects. These filters are can be produced at many discrete wavelengths. A Faraday filter consists of a pair of crossed polarizers on either side of a heated vapor cell mounted inside a solenoid.

  19. Faraday imaging at high temperatures

    International Nuclear Information System (INIS)

    Hackel, L.A.; Reichert, P.

    1997-01-01

    A Faraday filter rejects background light from self-luminous thermal objects, but transmits laser light at the passband wavelength, thus providing an ultra-narrow optical bandpass filter. The filter preserves images so a camera looking through a Faraday filter at a hot target illuminated by a laser will not see the thermal radiation but will see the laser radiation. Faraday filters are useful for monitoring or inspecting the uranium separator chamber in an atomic vapor laser isotope separation process. Other uses include viewing welds, furnaces, plasma jets, combustion chambers, and other high temperature objects. These filters are can be produced at many discrete wavelengths. A Faraday filter consists of a pair of crossed polarizers on either side of a heated vapor cell mounted inside a solenoid. 3 figs

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Modern high-temperature superconductivity

    International Nuclear Information System (INIS)

    Ching Wu Chu

    1988-01-01

    Ever since the discovery of superconductivity in 1911, its unusual scientific challenge and great technological potential have been recognized. For the past three-quarters of a century, superconductivity has done well on the science front. This is because sueprconductivity is interesting not only just in its own right but also in its ability to act as a probe to many exciting nonsuperconducting phenomena. For instance, it has continued to provide bases for vigorous activities in condensed matter science. Among the more recent examples are heavy-fermion systems and organic superconductors. During this same period of time, superconductivity has also performed admirably in the applied area. Many ideas have been conceived and tested, making use of the unique characteristics of superconductivity - zero resistivity, quantum interference phenomena, and the Meissner effect. In fact, it was not until late January 1987 that it became possible to achieve superconductivity with the mere use of liquid nitrogen - which is plentiful, cheap, efficient, and easy to handle - following the discovery of supercondictivity above 90 K in Y-Ba-Cu-O, the first genuine quaternary superconductor. Superconductivity above 90 K poses scientific and technological challenges not previously encountered: no existing theories can adequately describe superconductivity above 40 K and no known techniques can economically process the materials for full-scale applications. In this paper, therefore, the author recalls a few events leading to the discovery of the new class of quaternary compounds with a superconducting transition temperature T c in the 90 K range, describes the current experimental status of high-temperature superconductivity and, finally, discusses the prospect of very-high-temperature superconductivity, i.e., with a T c substantially higher than 100 K. 97 refs., 7 figs

  2. High temperature incineration. Densification of granules from high temperature incineration

    International Nuclear Information System (INIS)

    Voorde, N. van de; Claes, J.; Taeymans, A.; Hennart, D.; Gijbels, J.; Balleux, W.; Geenen, G.; Vangeel, J.

    1982-01-01

    The incineration system of radioactive waste discussed in this report, is an ''integral'' system, which directly transforms a definite mixture of burnable and unburnable radioactive waste in a final product with a sufficient insolubility to be safely disposed of. At the same time, a significant volume reduction occurs by this treatment. The essential part of the system is a high temperature incinerator. The construction of this oven started in 1974, and while different tests with simulated inactive or very low-level active waste were carried out, the whole system was progressively and continuously extended and adapted, ending finally in an installation with completely remote control, enclosed in an alpha-tight room. In this report, a whole description of the plant and of its auxiliary installations will be given; then the already gained experimental results will be summarized. Finally, the planning for industrial operation will be briefly outlined. An extended test with radioactive waste, which was carried out in March 1981, will be discussed in the appendix

  3. High temperature oxidation-sulfidation behavior of Cr-Al{sub 2}O{sub 3} and Nb-Al{sub 2}O{sub 3} composites densified by spark plasma sintering

    Energy Technology Data Exchange (ETDEWEB)

    Saucedo-Acuna, R.A. [Instituto e Ingenieria y Tecnologia, Universidad Autonoma de Cd. Juarez, Av. Del Charro 450 Norte, Col. Partido Romero, C.P. 32310, Cd. Juarez, Chihuahua (Mexico); Monreal-Romero, H.; Martinez-Villafane, A. [Centro de Investigacion en Materiales Avanzados, Departamento de Fisica de Materiales, Miguel de Cervantes 120, Complejo Industrial Chihuahua, C.P. 31109, Chihuahua (Mexico); Chacon-Nava, J.G. [Centro de Investigacion en Materiales Avanzados, Departamento de Fisica de Materiales, Miguel de Cervantes 120, Complejo Industrial Chihuahua, C.P. 31109, Chihuahua (Mexico)], E-mail: jose.chacon@cimav.edu.mx; Arce-Colunga, U. [Centro de Investigacion en Materiales Avanzados, Departamento de Fisica de Materiales, Miguel de Cervantes 120, Complejo Industrial Chihuahua, C.P. 31109, Chihuahua (Mexico); Universidad Autonoma de Tamaulipas, Matamoros 8 y 9 Col. Centro C.P. 87110, Cd. Victoria, Tamaulipas (Mexico); Gaona-Tiburcio, C. [Centro de Investigacion en Materiales Avanzados, Departamento de Fisica de Materiales, Miguel de Cervantes 120, Complejo Industrial Chihuahua, C.P. 31109, Chihuahua (Mexico); De la Torre, S.D. [Centro de Investigacion e Innovacion Tecnologica (CIITEC)-IPN, D.F. Mexico (Mexico)

    2007-12-15

    The high temperature oxidation-sulfidation behavior of Cr-Al{sub 2}O{sub 3} and Nb-Al{sub 2}O{sub 3} composites prepared by mechanical alloying (MA) and spark plasma sintering (SPS) has been studied. These composite powders have a particular metal-ceramic interpenetrating network and excellent mechanical properties. Oxidation-sulfidation tests were carried out at 900 deg. C, in a 2.5%SO{sub 2} + 3.6%O{sub 2} + N{sub 2}(balance) atmosphere for 48 h. The results revealed the influence of the sintering conditions on the specimens corrosion resistance, i.e. the Cr-Al{sub 2}O{sub 3} and Nb-Al{sub 2}O{sub 3} composite sintered at 1310 deg. C/4 min showed better corrosion resistance (lower weight gains) compared with those found for the 1440 deg. C/5 min conditions. For the former composite, a protective Cr{sub 2}O{sub 3} layer immediately forms upon heating, whereas for the later pest disintegration was noted. Thus, under the same sintering conditions the Nb-Al{sub 2}O{sub 3} composites showed the highest weight gains. The oxidation products were investigated by X-ray diffraction, scanning electron microscopy, and transmission electron microscopy.

  4. Chemical and structural effects on the high-temperature mechanical behavior of (1−x)(Na{sub 1/2}Bi{sub 1/2})TiO{sub 3}-xBaTiO{sub 3} ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Deluca, Marco [Materials Center Leoben Forschung GmbH, Roseggerstraße 12, 8700 Leoben (Austria); Institut für Struktur- und Funktionskeramik, Montanuniversitaet Leoben, Peter Tunner Straße 5, 8700 Leoben (Austria); Picht, Gunnar [Institute of Applied Materials, Ceramics in Mechanical Engineering, Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Robert Bosch GmbH, Corporate Sector Research and Advance Engineering Applied Research 1, Robert Bosch Platz 1, 70839 Gerlingen (Germany); Hoffmann, Michael J. [Institute of Applied Materials, Ceramics in Mechanical Engineering, Karlsruhe Institute of Technology, 76131 Karlsruhe (Germany); Rechtenbach, Annett; Töpfer, Jörg [Department of SciTec, University of Applied Sciences Jena, Carl-Zeiß-Promenade 2, 07745 Jena (Germany); Schader, Florian H.; Webber, Kyle G., E-mail: webber@ceramics.tu-darmstadt.de [Institute of Materials Science, Technische Universität Darmstadt, 64287 Darmstadt (Germany)

    2015-04-07

    Bismuth sodium titanate–barium titanate [(1−x)(Na{sub 1/2}Bi{sub 1/2})TiO{sub 3}-xBaTiO{sub 3}, NBT-100xBT] is one of the most well studied lead-free piezoelectric materials due in large part to the high field-induced strain attainable in compositions near the morphotropic phase boundary (x = 0.06). The BaTiO{sub 3}-rich side of the phase diagram, however, has not yet been as comprehensively studied, although it might be important for piezoelectric and positive temperature coefficient ceramic applications. In this work, we present a thorough study of BaTiO{sub 3}-rich NBT-100xBT by ferroelastic measurements, dielectric permittivity, X-ray diffraction, and Raman spectroscopy. We show that the high-temperature mechanical behavior, i.e., above the Curie temperature, T{sub C}, is influenced by local disorder, which appears also in pure BT. On the other hand, in NBT-100xBT (x < 1.0), lattice distortion, i.e., tetragonality, increases, and this impacts both the mechanical and dielectric properties. This increase in lattice distortion upon chemical substitution is counterintuitive by merely reasoning on the ionic size, and is due to the change in the A-O bond character induced by the Bi{sup 3+} electron lone pair, as indicated by Raman spectroscopy.

  5. New diffusion mechanism for high temperature diffusion in solids

    International Nuclear Information System (INIS)

    Doan, N.V.; Adda, Y.

    1986-09-01

    A new atomic transport mechanism in solids at high temperatures has been discovered by Molecular Dynamics computer simulation. It can be described as a ring sequence of atomic replacements induced by unstable Frenkel pairs. This transport process takes place without stable defects, the atomic migration occurring indeed by simultaneous creation and migration of unstable defects. Starting from the analysis of this mechanism in different solids at high temperature (CaF 2 , Na, Ar) and in irradiated copper by subthreshold collisions, we discuss the role of this mechanism on various diffusion controlled phenomena and also on the atomic processes of defect creation

  6. Present status of high temperature engineering test and research, 1994

    International Nuclear Information System (INIS)

    1994-10-01

    High temperature gas-cooled reactors have excellent features such as the generation of high temperature close to 1000degC, very high inherent safety and high fuel burnup. By the advanced basic research under high temperature irradiation condition, the creation of various new technologies which become the momentum of future technical innovation can be expected. The construction of the high temperature engineering test reactor (HTTR) was decided in 1987, which aims at the thermal output of 30 MW and the coolant temperature at reactor exit of 950degC. The initial criticality is scheduled in 1998. Japan Atomic Energy Research Institute has advanced the high temperature engineering test and research, and plans the safety verifying test of the HTTR, the test of connecting heat utilization plants and so on. In this report, mainly the results obtained for one year from May, 1993 are summarized. The outline of the high temperature engineering test and development of the HTTR technologies are reported. (K.I.)

  7. Evaluation of fuel rods behavior - under irradiation test

    International Nuclear Information System (INIS)

    Lameiras, F.S.; Terra, J.L.; Pinto, L.C.M.; Dias, M.S.; Pinheiro, R.B.

    1981-04-01

    By the accompanying of the irradiation of instrumented test fuel rods simulating the operational conditions in reactors, plus the results of post - irradiation exams, tests, evaluation and calibration of analitic modelling of such fuel rods is done. (E.G.) [pt

  8. Behavioral changes in rats prenatally irradiated with low dose of gamma-rays

    International Nuclear Information System (INIS)

    Kiskova, J.; Smajda, B.; Capicikova, M.; Lievajova, K.

    2006-01-01

    In this work, the effects of prenatal gamma-irradiation on behavior in adult Sprague-Dawley rats were studied. Four months old female rats were irradiated with a dose of 1 Gy of gamma-rays on day 15 of gestation. The offspring of irradiated mothers (n=26) and that of control, non-irradiated mothers (n=36) of both sexes at the age of 3 month were tested in Morris's water maze and in open field test. All experimental groups showed a tendency to shortening the time needed to reach the platform in each trial in Morris water maze. Statistically significant difference between irradiated and control rats was detected only in males on 3 rd experimental day. The ability to remember the position of the platform was not altered in irradiated animals after a 4 day pause. In open field test, statistically significant differences in comparison with controls were detected in number of squares entered and in crossings of the central square (P ≤ 0.05) in males. These findings suggest, when comparing with results of other authors, that irradiation effects on postnatal behavior in rats are extremely dependent on the time point of irradiation and that a correlation exist between the developmental stage of the individual brain structures at time of irradiation and the late behavioral effects. (authors)

  9. 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.

  10. Morphological and behavioral evaluation of central nervous system toxicity after embryonic x-irradiation

    International Nuclear Information System (INIS)

    Schneider, B.F.

    1979-01-01

    The purpose of these studies was to quantify the effects of embryonic x-irradiation on neuronal development and to describe the behavioral effects of such a treatment. Rats were exposed to 125r x-irradiation on gestational day 15

  11. Irradiation behavior of uranium-silicide dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.

    1984-01-01

    This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

  12. Materials for advanced high temperature reactors

    International Nuclear Information System (INIS)

    Graham, L.W.

    1977-01-01

    Materials are studied in advanced applications of high temperature reactors: helium gas turbine and process heat. Long term creep behavior and corrosion tests are conducted in simulated HTR helium up to 1000 deg C with impurities additions in the furnace atmosphere. Corrosion studies on AISI 321 steels at 800-1000 deg C have shown that the O 2 partial pressure is as low as 10 -24+-3 atm, Ni and Fe cannot be oxidised above about 500 and 600 deg C, Cr cease to oxidise at 800 to 900 deg C and Ti at 900 to 1000 deg C depending on alloy composition γ' strengthened superalloys must depend on a protective corrosion mechanism assisted by the presence of Ti and possibly Cr. Carburisation has been identified metallographically in several high temperature materials: Hastelloy X and M21Z. Alloy TZM appears to be inert in HTR Helium at 900 and 1000 deg C. In alloy 800 and Inconel 625 surface cracks initiation is suppressed but crack propagation is accelerated but this was not apparent in AISI steels, Hastelloy X or fine grain Inconel at 750 deg C

  13. Evaluation of high temperature capacitor dielectrics

    Science.gov (United States)

    Hammoud, Ahmad N.; Myers, Ira T.

    1992-01-01

    Experiments were carried out to evaluate four candidate materials for high temperature capacitor dielectric applications. The materials investigated were polybenzimidazole polymer and three aramid papers: Voltex 450, Nomex 410, and Nomex M 418, an aramid paper containing 50 percent mica. The samples were heat treated for six hours at 60 C and the direct current and 60 Hz alternating current breakdown voltages of both dry and impregnated samples were obtained in a temperature range of 20 to 250 C. The samples were also characterized in terms of their dielectric constant, dielectric loss, and conductivity over this temperature range with an electrical stress of 60 Hz, 50 V/mil present. Additional measurements are underway to determine the volume resistivity, thermal shrinkage, and weight loss of the materials. Preliminary data indicate that the heat treatment of the films slightly improves the dielectric properties with no influence on their breakdown behavior. Impregnation of the samples leads to significant increases in both alternating and direct current breakdown strength. The results are discussed and conclusions made concerning their suitability as high temperature capacitor dielectrics.

  14. High temperature superconductor current leads

    International Nuclear Information System (INIS)

    Zeimetz, B.; Liu, H.K.; Dou, S.X.

    1996-01-01

    Full text: The use of superconductors in high electrical current applications (magnets, transformers, generators etc.) usually requires cooling with liquid Helium, which is very expensive. The superconductor itself produces no heat, and the design of Helium dewars is very advanced. Therefore most of the heat loss, i.e. Helium consumption, comes from the current lead which connects the superconductor with its power source at room temperature. The current lead usually consists of a pair of thick copper wires. The discovery of the High Temperature Superconductors makes it possible to replace a part of the copper with superconducting material. This drastically reduces the heat losses because a) the superconductor generates no resistive heat and b) it is a very poor thermal conductor compared with the copper. In this work silver-sheathed superconducting tapes are used as current lead components. The work comprises both the production of the tapes and the overall design of the leads, in order to a) maximize the current capacity ('critical current') of the superconductor, b) minimize the thermal conductivity of the silver clad, and c) optimize the cooling conditions

  15. Container floor at high temperatures

    International Nuclear Information System (INIS)

    Reutler, H.; Klapperich, H.J.; Mueller-Frank, U.

    1978-01-01

    The invention describes a floor for container which is stressed at high, changing temperatures and is intended for use in gas-cooled nuclear reactors. Due to the downward cooling gas flow in these types of reactor, the reactor floor is subjected to considerable dimensional changes during switching on and off. In the heating stage, the whole graphite structure of the reactor core and floor expands. In order to avoid arising constraining forces, sufficiently large expansion spaces must be allowed for furthermore restoring forces must be present to close the gaps again in the cooling phase. These restoring forces must be permanently present to prevent loosening of the core cuits amongst one another and thus uncontrollable relative movement. Spring elements are not suitable due to fast fatigue as a result of high temperatures and radiation exposure. It is suggested to have the floor elements supported on rollers whose rolling planes are downwards inclined to a fixed point for support. The construction is described in detail by means of drawings. (GL) [de

  16. High Temperature Superconductor Accelerator Magnets

    CERN Document Server

    AUTHOR|(CDS)2079328; de Rijk, Gijs; Dhalle, Marc

    2016-11-10

    For future particle accelerators bending dipoles are considered with magnetic fields exceeding $20T$. This can only be achieved using high temperature superconductors (HTS). These exhibit different properties from classical low temperature superconductors and still require significant research and development before they can be applied in a practical accelerator magnet. In order to study HTS in detail, a five tesla demonstrator magnet named Feather-M2 is designed and constructed. The magnet is based on ReBCO coated conductor, which is assembled into a $10kA$ class Roebel cable. A new and optimized Aligned Block layout is used, which takes advantage of the anisotropy of the conductor. This is achieved by providing local alignment of the Roebel cable in the coil windings with the magnetic field lines. A new Network Model capable of analyzing transient electro-magnetic and thermal phenomena in coated conductor cables and coils is developed. This model is necessary to solve critical issues in coated conductor ac...

  17. High Temperature Superconducting Underground Cable

    International Nuclear Information System (INIS)

    Farrell, Roger A.

    2010-01-01

    The purpose of this Project was to design, build, install and demonstrate the technical feasibility of an underground high temperature superconducting (HTS) power cable installed between two utility substations. In the first phase two HTS cables, 320 m and 30 m in length, were constructed using 1st generation BSCCO wire. The two 34.5 kV, 800 Arms, 48 MVA sections were connected together using a superconducting joint in an underground vault. In the second phase the 30 m BSCCO cable was replaced by one constructed with 2nd generation YBCO wire. 2nd generation wire is needed for commercialization because of inherent cost and performance benefits. Primary objectives of the Project were to build and operate an HTS cable system which demonstrates significant progress towards commercial progress and addresses real world utility concerns such as installation, maintenance, reliability and compatibility with the existing grid. Four key technical areas addressed were the HTS cable and terminations (where the cable connects to the grid), cryogenic refrigeration system, underground cable-to-cable joint (needed for replacement of cable sections) and cost-effective 2nd generation HTS wire. This was the worlds first installation and operation of an HTS cable underground, between two utility substations as well as the first to demonstrate a cable-to-cable joint, remote monitoring system and 2nd generation HTS.

  18. High-temperature axion potential

    International Nuclear Information System (INIS)

    Dowrick, N.J.; McDougall, N.A.

    1989-01-01

    We investigate the possibility of new terms in the high-temperature axion potential arising from the dynamical nature of the axion field and from higher-order corrections to the θ dependence in the free energy of the quark-gluon plasma. We find that the dynamical nature of the axion field does not affect the potential but that the higher-order effects lead to new terms in the potential which are larger than the term previously considered. However, neither the magnitude nor the sign of the potential can be calculated by a perturbative expansion of the free energy since the coupling is too large. We show that a change in the magnitude of the potential does not significantly affect the bound on the axion decay constant but that the sign of the potential is of crucial importance. By investigating the formal properties of the functional integral within the instanton dilute-gas approximation, we find that the sign of the potential does not change and that the minimum remains at θ=0. We conclude that the standard calculation of the axion energy today is not significantly modified by this investigation

  19. Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Hofman, G.L. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2012-06-15

    Highlights: Black-Right-Pointing-Pointer We in-pile tested U-Mo dispersion in Al matrix. Black-Right-Pointing-Pointer We observed interaction layer growth between U-Mo and Al and pore formation there. Black-Right-Pointing-Pointer Pores degrades thermal conductivity and structural integrity of the fueled zone. Black-Right-Pointing-Pointer The amorphous behavior of interaction layers is thought to be the main reason for unstable large pore growth. Black-Right-Pointing-Pointer A mechanism for pore formation and possible remedy to prevent it are proposed. - Abstract: Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

  20. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    Onimus, F.

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  1. Irradiation behavior of graphite shielding materials for FBR

    International Nuclear Information System (INIS)

    Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

    1994-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

  2. High Temperature Chemistry at NASA: Hot Topics

    Science.gov (United States)

    Jacobson, Nathan S.

    2014-01-01

    High Temperature issues in aircraft engines Hot section: Ni and Co based Superalloys Oxidation and Corrosion (Durability) at high temperatures. Thermal protection system (TPS) and RCC (Reinforced Carbon-Carbon) on the Space Shuttle Orbiter. High temperatures in other worlds: Planets close to their stars.

  3. High temperature turbine engine structure

    Energy Technology Data Exchange (ETDEWEB)

    Carruthers, W.D.; Boyd, G.L.

    1993-07-20

    A hybrid ceramic/metallic gas turbine is described comprising; a housing defining an inlet, an outlet, and a flow path communicating the inlet with the outlet for conveying a flow of fluid through the housing, a rotor member journaled by the housing in the flow path, the rotor member including a compressor rotor portion rotatively inducting ambient air via the inlet and delivering this air pressurized to the flow path downstream of the compressor rotor, a combustor disposed in the flow path downstream of the compressor receiving the pressurized air along with a supply of fuel to maintain combustion providing a flow of high temperature pressurized combustion products in the flow path downstream thereof, the rotor member including a turbine rotor portion disposed in the flow path downstream of the combustor and rotatively expanding the combustion products toward ambient for flow from the turbine engine via the outlet, the turbine rotor portion providing shaft power driving the compressor rotor portion and an output shaft portion of the rotor member, a disk-like metallic housing portion journaling the rotor member to define a rotational axis therefore, and a disk-like annular ceramic turbine shroud member bounding the flow path downstream of the combustor and circumscribing the turbine rotor portion to define a running clearance therewith, the disk-like ceramic turbine shroud member having a reference axis coaxial with the rotational axis and being spaced axially from the metallic housing portion in mutually parallel concentric relation therewith and a plurality of spacers disposed between ceramic disk-like shroud member and the metallic disk-like housing portion and circumferentially spaced apart, each of the spacers having a first and second end portion having an end surface adjacent the shroud member and the housing portion respectively, the end surfaces having a cylindrical curvature extending transversely relative to the shroud member and the housing portion.

  4. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer Syst