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Sample records for neutron-irradiated molybdenum alloys

  1. Radiation Damages in Aluminum Alloy SAV-1 under Neutron Irradiation

    Science.gov (United States)

    Salikhbaev, Umar; Akhmedzhanov, Farkhad; Alikulov, Sherali; Baytelesov, Sapar; Boltabaev, Azizbek

    2016-05-01

    The aim of this work was to study the effect of neutron irradiation on the kinetics of radiation damages in the SAV-1 alloy, which belongs to the group of aluminum alloys of the ternary system Al-Mg-Si. For fast-neutron irradiation by different doses up to fluence 1019 cm-2 the SAV-1 samples were placed in one of the vertical channels of the research WWR type reactor (Tashkent). The temperature dependence of the electrical resistance of the alloy samples was investigated in the range 290 - 490 K by the four-compensation method with an error about 0.1%. The experimental results were shown that at all the temperatures the dependence of the SAV-1 alloy resistivity on neutron fluence was nonlinear. With increasing neutron fluence the deviation from linearity and the growth rate of resistivity with temperature becomes more appreciable. The observed dependences are explained by means of martensitic transformations and the radiation damages in the studied alloy under neutron irradiation. The mechanisms of radiation modification of the SAV-1 alloy structure are discussed.

  2. Swelling and tensile properties of neutron-irradiated vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.; Smith, D.L.

    1990-07-01

    Vanadium-base alloys are candidates for use as structural material in magnetic fusion reactors. In comparison to other candidate structural materials (e.g., Type 316 stainless and HT-9 ferritic steels), vanadium-base alloys such as V-15Cr-5Ti and V-20Ti have intrinsically lower long-term neutron activation, neutron irradiation after-heat, biological hazard potential, and neutron-induced helium and hydrogen transmutation rates. Moreover, vanadium-base alloys can withstand a higher surface-heat, flux than steels because of their lower thermal stress factor. In addition to having these favorable neutronic and physical properties, a candidate alloy for use as structural material in a fusion reactor must have dimensional stability, i.e., swelling resistance, and resistance to embrittlement during the reactor lifetime at a level of structural strength commensurate with the reactor operating temperature and structural loads. In this paper, we present experimental results on the swelling and tensile properties of several vanadium-base alloys after irradiation at 420, 520, and 600{degree}C to neutron fluences ranging from 0.3 to 1.9 {times} 10{sup 27} neutrons/m{sup 2} (17 to 114 atom displacements per atom (dpa)).

  3. Microstructure and dimensional changes of neutron-irradiated zirconium alloys

    Science.gov (United States)

    Pedraza, A. J.; Fainstein-Pedraza, D.

    1982-08-01

    Experimental observations concerning the neutron-irradiation-induced defect structure in zirconium-based alloys are analyzed within the framework of an irradiation growth theory developed in the past years. The competition of those defects and the microstructure present in the material prior to irradiation as point defect sinks is studied as a function of irradiation temperature and dose. Owing to the different growth behavior of recrystallized and of cold-worked specimens at reactor temperatures, the cellular microstructure of the latter is considered in detail. In view of the highly anisotropic dislocation system in these materials, cell boundaries are reasoned to form with essentially edge components in the walls parallel to the c-axis, while the boundary segments normal to that axis should be of screw type. Since the latter would induce no dimensional change if the cell boundary absorbs defects by dislocation climb, it is argued — on the basis of growth data — that it must behave as a point defect sink/source of a different nature than that of free dislocations. The possibility of dislocation segments with a ( c + a)-type Burgers vector in the cell boundary or rapid point defect diffusion along it are also discussed. The existing growth model is then enlarged in order to account quantitatively for the dimensional changes of cold-worked materials, and its results are compared with available experimental data.

  4. Study of the temperature evolution of defect agglomerates in neutron irradiated molybdenum single crystals

    Energy Technology Data Exchange (ETDEWEB)

    Lambri, O.A. [Instituto de Fisica Rosario. Member of the CONICET' s Research Staff, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina); Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina)], E-mail: olambri@fceia.unr.edu.ar; Zelada-Lambri, G.I. [Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, (2000) Rosario, Santa Fe (Argentina); Cuello, G.J. [Institut Laue Langevin, 6, rue Jules Horowitz, BP 156, 38042 Grenoble (France); Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain); Bozzano, P.B. [Laboratorio de Microscopia Electronica. Unidad de Actividad Materiales, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Avda. Gral. Paz 1499, (1650) San Martin (Argentina); Garcia, J.A. [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain)

    2009-04-15

    Small angle neutron scattering as a function of temperature, differential thermal analysis, electrical resistivity and transmission electron microscopy studies have been performed in low rate neutron irradiated single crystalline molybdenum, at room temperature, for checking the evolution of the defects agglomerates in the temperature interval between room temperature and 1200 K. The onset of vacancies mobility was found to happen in temperatures within the stage III of recovery. At around 550 K, the agglomerates of vacancies achieve the largest size, as determined from the Guinier approximation for spherical particles. In addition, the decrease of the vacancy concentration together with the dissolution of the agglomerates at temperatures higher than around 920 K was observed, which produce the release of internal stresses in the structure.

  5. Influence of composition, heat treatment and neutron irradiation on the electrical conductivity of copper alloys

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, B.N.

    1998-01-01

    The electrical conductivity of three different types of copper alloys, viz. CuNiBe, CuCrZr and Cu-Al(2)O(3) as well as of pure copper are reported. The alloys have undergone different pre-irradiation heat treatments and have been fission-neutron irradiated up to 0.3 dpa. In some cases post...

  6. Mechanical and microstructural properties of neutron irradiated Fe-Cr-C alloys

    Energy Technology Data Exchange (ETDEWEB)

    Konstantinovic, M.J.; Renterghem, W. van; Matijasevic, M.; Minov, B.; Lambrecht, M.; Chiapetto, M.; Malerba, L. [Studiecentrum voor Kernenergie/Centre d' Etude de l' Energie Nucleaire (SCK-CEN), Mol (Belgium); Toyama, T. [Institute for Materials Research, Tohoku University, Sendai (Japan)

    2016-11-15

    Defect properties of neutron irradiated Fe-Cr-C alloys and their influence on the mechanical behavior are studied by combining mechanical tests, microstructural examination, and the results of models. It is found that the initial microstructure of these alloys, determined by the Cr and C concentrations, as well as by the thermal treatment, can account for different defect formation and distribution after neutron irradiation. On the basis of these results, a correlation between defect properties and macroscopic mechanical behavior is proposed. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Correlation between shear punch and tensile data for neutron-irradiated aluminum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, M.L.; Edwards, D.J. [Pacific Northwest Laboratory, Richland, WA (United States); Toloczko, M.B. [Univ. of California, Santa Barbara, CA (United States)] [and others

    1995-04-01

    This work was performed to determine whether shear punch and tensile data obtained on neutron irradiated aluminum alloys exhibited the same type of relationship as had been seen in other work and to assess the validity of extrapolating the results to proton-irradiated alloys. This work was also meant to be the first of a series of similar test matrices designed to determine whether the shear punch/tensile relationship varied or was the same for different alloy classes.

  8. Correlative Microscopy of Alpha Prime Precipitation in Neutron-Irradiated Fe-Cr-Al Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States)

    2016-12-01

    Fe-Cr-Al alloys are currently being considered for accident tolerant light water reactor fuel cladding applications due to their superior high temperature oxidation and corrosion resistance compared to Zr-based alloys. This work represents the current state-of-the-art on both techniques for analysis of α' precipitate microstructures and the processes and mechanisms governing its formation in neutron-irradiated Fe-Cr-Al alloys.

  9. Embrittlement of molybdenum-rhenium welds under low and high temperature neutron irradiation

    Science.gov (United States)

    Krajnikov, A. V.; Morito, F.; Danylenko, M. I.

    2014-01-01

    The effect of low- and high-temperature neutron irradiation on the tensile strength, microhardness, and fracture mode has been studied for a series of Mo-Re welds with various Re concentrations. Radiation-induced hardening and concurrent ductility reduction are the key after-effects of neutron exposure. Low-temperature irradiation usually leads to a very hard embrittlement. The hardening effect is rather limited and unstable because of the lack of ductility. Irradiated specimens fail by brittle intergranular or transgranular fracture. The damaging effect of neutrons is less pronounced after high-temperature irradiation. The hardening of the matrix is rather high, but irradiated specimens still keep residual plasticity. High-temperature irradiation intensifies homogeneous nucleation of Re-rich phases, and this effect equalises the difference in mechanical properties between the different weld zones. A characteristic ductility loss exposure temperature was found to separate the temperature fields of absolutely brittle and relatively ductile behaviour. It usually varies between 850 K and 1000 K depending on the alloy composition and irradiation conditions.

  10. Mechanical energy losses in plastically deformed and electron plus neutron irradiated high purity single crystalline molybdenum at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Zelada, Griselda I. [Laboratorio de Materiales, Escuela de Ingenieria Electrica, Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Avda. Pellegrini 250, 2000 Rosario (Argentina); Lambri, Osvaldo Agustin [Laboratorio de Materiales, Escuela de Ingenieria Electrica, Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Avda. Pellegrini 250, 2000 Rosario (Argentina); Instituto de Fisica Rosario - CONICET, Member of the CONICET& #x27; s Research Staff, Avda. Pellegrini 250, 2000 Rosario (Argentina); Bozzano, Patricia B. [Laboratorio de Microscopia Electronica, Unidad de Actividad Materiales, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Avda. Gral. Paz 1499, 1650 San Martin (Argentina); Garcia, Jose Angel [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain)

    2012-10-15

    Mechanical spectroscopy (MS) and transmission electron microscopy (TEM) studies have been performed in plastically deformed and electron plus neutron irradiated high purity single crystalline molybdenum, oriented for single slip, in order to study the dislocation dynamics in the temperature range within one third of the melting temperature. A damping peak related to the interaction of dislocation lines with both prismatic loops and tangles of dislocations was found. The peak temperature ranges between 900 and 1050 K, for an oscillating frequency of about 1 Hz. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. Subtask 12F3: Effects of neutron irradiation on tensile properties of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the tensile properties of candidate vanadium-base alloys. Vanadium-base alloys of the V-Cr-Ti system are attractive candidates for use as structural materials in fusion reactors. The current focus of the U.S. program of research on these alloys is on the V-(4-6)Cr-(3-6)Ti-(0.05-0.1)Si (in wt.%) alloys. In this paper, we present experimental results on the effects of neutron irradiation on tensile properties of selected candidate alloys after irradiation at 400{degrees}C-600{degrees}C in lithium in fast fission reactors to displacement damages of up to {approx}120 displacement per atom (dpa). Effects of irradiation temperature and dose on yield and ultimate tensile strengths and uniform and total elongations are given for tensile test temperatures of 25{degrees}C, 420{degrees}C, 500{degrees}, and 600{degrees}C. Effects of neutron damage on tensile properties of the U.S. reference alloy V-4Cr-4Ti are examined in detail. 7 refs., 10 figs., 1 tab.

  12. 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.

  13. Embrittlement behaviour of different international low activation alloys after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2001-05-01

    The embrittlement behaviour of ferritic/martensitic steels after irradiation in the Petten high flux reactor (HFR) was investigated by instrumented Charpy-V tests with subsize specimens. The main objective, apart from studying effects of particularly low doses, was a comparison of low activation alloys (LAA) from various countries with different Cr contents and different types and concentrations of minor alloying elements and impurities. In the present report, the results of another three materials (OPTIMAR, OPTIFER-IV, GA3X) obtained within the second phase of the MANITU programme (0.8 dpa, at 250-450°C) were analysed and assessed in comparison to the results of the first irradiation up to 0.8 dpa. The evaluation clearly showed a reduced embrittlement problem for the advanced reduced-activation alloys. Of the examined alloys, the GA3X steel shows the very best embrittlement behaviour after neutron irradiation.

  14. Correlative Microscopy of alpha' Precipitation in Neutron-Irradiated Fe-Cr-Al Alloys

    Science.gov (United States)

    Briggs, Samuel A.

    Fe-Cr-Al alloys are currently being considered for accident tolerant light water reactor fuel cladding applications due to their superior high temperature oxidation and corrosion resistance compared to Zr-based alloys. However, precipitation of the Cr-rich alpha' phase during exposure to LWR operational environments can result in application-limiting hardening and embrittlement. To study this effect, four Fe-Cr-Al model alloys with compositions between 10-18 at.% Cr and 5.8-9.3 at.% Al have been neutron-irradiated in the High Flux Isotope Reactor at a target temperature of 320°C to nominal damage doses of up to 7 dpa in order to emulate typical LWR exposure conditions. A correlative microscopy approach involving atom probe tomography, small-angle neutron scattering, and scanning transmission electron microscopy coupled with energy dispersive x-ray spectroscopy was employed to study the resulting precipitate microstructure. This approach necessitated the development of various analysis techniques to allow for cross-comparison between experimental techniques, including a novel method for correcting for trajectory aberration artifacts in atom probe data sets through phase density comparison. Successful correlation of results from these techniques allows for the individual limitations of each to be overcome and enables the detailed microstructural information gleaned from highly localized atom probe tomography analyses to be extrapolated to bulk alloy behavior. Precipitation response was found to increase with Cr content, while Al additions appeared to partially destabilized the alpha' phase, resulting in precipitate compositions with reduced Cr content compared to binary Fe-Cr systems. Observed precipitate evolution with radiation dose indicates a diffusion-limited coarsening mechanism that is similar to what is observed in the thermally aged system. This work represents the current state-of-the-art on both techniques for analysis of alpha' precipitate

  15. Subtask 12F1: Effect of neutron irradiation on swelling of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Loomis, B.A.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to determine the effects of neutron irradiation on the density change, void distribution, and microstructural evolution of vanadium-base alloys. Swelling behavior and microstructural evolution of V-Ti, V-Cr-Ti, and V-Ti-Si alloys were investigated after irradiation at 420-600{degrees}C up to 114 dpa. The alloys exhibited swelling maxima between 30 and 80 dpa and swelling decreased on irradiation to higher dpa. This is in contrast to the monotonically increasing swelling of binary alloys that contain Fe, Ni, Cr, Mo, W, and Si. Precipitation of dense Ti{sub 5}Si{sub 3} promotes good resistance to swelling of the Ti-containing alloys, and it was concluded that Ti of >3 wt.% and 400-1000 wppm Si are necessary to effectively suppress swelling. Swelling was minimal in V-4Cr-4Ti, identified as the most promising alloy based on good mechanical properties and superior resistance to irradiation embrittlement. 18 refs., 6 figs., 1 tab.

  16. Neutron irradiation effects on the microstructural development of tungsten and tungsten alloys

    Science.gov (United States)

    Hasegawa, Akira; Fukuda, Makoto; Yabuuchi, Kiyohiro; Nogami, Shuhei

    2016-04-01

    Data on the microstructural development of tungsten (W) and tungsten rhenium (Re) alloys were obtained after neutron irradiation at 400-800 °C in the Japan Materials Testing Reactor (JMTR), the experimental fast test reactor Joyo, and the High Flux Isotope Reactor (HFIR) for irradiation damage levels in the range of 0.09-1.54 displacement per atom (dpa). Microstructural observations showed that a small amount of Re (3-5%) in W-Re alloys is effective in suppressing void formation. In W-Re alloys with Re concentrations greater than 10%, acicular precipitates are the primary structural defects. In the HFIR-irradiated specimen, in which a large amount of Re was expected to be produced by the nuclear transmutation of W to Re because of the reactor's high thermal neutron flux, voids were not observed even in pure W. The synergistic effects of displacement damage and solid transmutation elements on microstructural development are discussed, and the microstructural development of tungsten materials utilized in fusion reactors is predicted.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-01

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

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

    Science.gov (United States)

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

    2017-02-01

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

  19. EL2-related defects in neutron irradiated GaAs/sub 1//sub -x/P/sub x/ alloys

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, E.; Garcia, F.; Jimenez, B.; Calleja, E.; Gomez, A.; Alcober, V.

    1985-10-15

    The generation of EL2-related defects in GaAsP alloys by fast neutron irradiation has been studied through deep level transient spectroscopy and photocapacitance techniques. After irradiation p-n junctions were not annealed at high temperatures. In the composition range x>0.4, fast neutrons generate a broad center at E/sub c/-0.7 eV that it is suggested to belong to the EL2 family. The presence of photocapacitance quenching effects has been taken as a preliminary fingerprint to make the above assignment. From computer analysis of the nonexponential transient capacitance waveforms, evidence that neutron irradiation creates a family of midgap levels, EL2-related, is found.

  20. A replica technique for extracting precipitates from neutron-irradiated or thermal-aged vanadium alloys for TEM analysis

    Energy Technology Data Exchange (ETDEWEB)

    Fukumoto, K., E-mail: fukumoto@u-fukui.ac.jp; Iwasaki, M.

    2014-06-01

    A carbon replica technique has been developed to extract precipitates from vanadium alloys. Using this technique, precipitation phases can be extracted from neutron-irradiated or thermal-aged V–4Cr–4Ti alloys. Precipitate identification using EDS X-ray analysis and electron diffraction was facilitated. Only NaCl type of Ti(OCN) precipitate was formed in the thermal-aged V–4Cr–4Ti alloys at 600 °C for 20 h and cation sub-lattice was only occupied by Ti atoms. However, the thin plate of precipitates with NaCl type of crystallographic structure could be seen in the V–4Cr–4Ti alloys irradiated at 593 °C in the JOYO fast reactor. The precipitate contained chromium and vanadium atoms on the cation sub-lattice as well as titanium atoms. It is considered that the phase of MX type (M = Ti, V, Cr and X = O, N, C) is a metastable phase under neutron irradiation.

  1. A replica technique for extracting precipitates from neutron-irradiated or thermal-aged vanadium alloys for TEM analysis

    Science.gov (United States)

    Fukumoto, K.; Iwasaki, M.

    2014-06-01

    A carbon replica technique has been developed to extract precipitates from vanadium alloys. Using this technique, precipitation phases can be extracted from neutron-irradiated or thermal-aged V-4Cr-4Ti alloys. Precipitate identification using EDS X-ray analysis and electron diffraction was facilitated. Only NaCl type of Ti(OCN) precipitate was formed in the thermal-aged V-4Cr-4Ti alloys at 600 °C for 20 h and cation sub-lattice was only occupied by Ti atoms. However, the thin plate of precipitates with NaCl type of crystallographic structure could be seen in the V-4Cr-4Ti alloys irradiated at 593 °C in the JOYO fast reactor. The precipitate contained chromium and vanadium atoms on the cation sub-lattice as well as titanium atoms. It is considered that the phase of MX type (M = Ti, V, Cr and X = O, N, C) is a metastable phase under neutron irradiation.

  2. Microstructural changes of Y-doped V-4Cr-4Ti alloys after ion and neutron irradiation

    Directory of Open Access Journals (Sweden)

    H. Watanabe

    2016-12-01

    Full Text Available High-purity Y-doped V-4Cr-4Ti alloys (0.1–0.2wt. % Y, manufactured by the National Institute for Fusion Science (NIFS, were used for this study. Heavy-ion and fission-neutron irradiation was carried out at temperatures 673–873K. During the ion irradiation at 873K, the microstructure was controlled by the formation of Ti(C,O,N precipitates lying on the (100 plane. Y addition effectively suppressed the growth of Ti(C,O,N precipitates, especially at lower dose irradiation to up to 4 dpa. However, at higher dose levels (12.0 dpa, the number density was almost at the same levels irrespective of the presence of Y. After neutron irradiation at 873K, fine titanium oxides were also observed in all V alloys. However, smaller oxide sizes were observed in the Y-doped samples under the same irradiation conditions. The detailed analysis of EDS showed that the center of the Ti(C,O,N precipitates was mainly enriched by nitrogen. The results showed that the contribution of not only oxygen atoms picked up from the irradiation environment but also nitrogen atoms is essential to understand the microstructural evolution of V-4Cr-4Ti-Y alloys.

  3. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Science.gov (United States)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  4. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: the case of Fe-Cu model alloys

    CERN Document Server

    Subbotina, A V

    2016-01-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. We show that the obtained results are in a good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  5. Subtask 12F4: Effects of neutron irradiation on the impact properties and fracture behavior of vanadium-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Loomis, B.A.; Smith, D.L. [Argonne National Lab., IL (United States)

    1995-03-01

    Up-to-date results on the effects of neutron irradiation on the impact properties and fracture behavior of V, V-Ti, V-Cr-Ti and V-Ti-Si alloys are presented in this paper, with an emphasis on the behavior of the U.S. reference alloys V-4Cr-4Ti containing 500-1000 wppm Si. Database on impact energy and cluctile-brittle transition temperature (DBTT) has been established from Charpy impact tests of one-third-size specimens irradiated at 420{degrees}C-600{degrees}C up to {approx}50 dpa in lithium environment in fast fission reactors. To supplement the Charpy impact tests fracture behavior was also characterized by quantitative SEM fractography on miniature tensile and disk specimens that were irradiated to similar conditions and fractured at -196{degrees}C to 200{degrees}C by multiple bending. For similar irradiation conditions irradiation-induced increase in DBTT was influenced most significantly by Cr content, indicating that irradiation-induced clustering of Cr atoms takes place in high-Cr (Cr {ge} 7 wt.%) alloys. When combined contents of Cr and Ti were {le}10 wt.%, effects of neutron irradiation on impact properties and fracture behavior were negligible. For example, from the Charpy-impact and multiple-bend tests there was no indication of irradiation-induced embrittlement for V-5Ti, V-3Ti-1Si and the U.S. reference alloy V-4Cr-4Ti after irradiation to {approx}34 dpa at 420{degrees}C to 600{degrees}C, and only ductile fracture was observed for temperatures as low as -196{degrees}C. 14 refs., 8 figs., 1 tab.

  6. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Edwards, D.J.; Singh, B.N.; Bilde-Sørensen, Jørgen

    2005-01-01

    The formation of ‘cleared’ channels in neutron irradiated metals and alloys have been frequently reported for more than 40 years. So far, however, no unambiguous and conclusive evidence showing as to how and where these channels are initiated has emerged. In the following we present experimental ...

  7. Effect of neutron irradiation and postradiation annealing on the microstructure and properties of an Al-Mg-Si alloy

    Science.gov (United States)

    Maksimkin, O. P.; Tsai, K. V.; Rofman, O. V.; Sil'nyagina, N. S.

    2016-09-01

    The effect of long-term neutron irradiation and postradiation thermal-induced aging on the microstructure and mechanical properties of an aluminum-based reactor Al-Mg-Si alloy grade SAV-1 has been studied. The material under study is the shell of an automatic fine-control rod used to control the reactivity of the core of a VVR-K research reactor. Successive 1-h annealings of specimens of the SAV-1 alloy irradiated to doses of 0.001 and 5 dpa in the temperature range of 100-550°C have been carried out. The evolution of the fine structure of the material and changes in its mechanical characteristics have been studied. The phenomenon of the acceleration of the aging of the SAV-1 alloy under the effect of a high neutron fluence at an irradiation temperature of 80°C has been observed, which involves the formation of numerous lineage (stitch) Guinier-Preston zones in the alloy. It has been shown that the strength characteristics of the SAV-1 alloy depend significantly on the degree of its radiation- and thermal-induced aging.

  8. Mechanical properties and microstructure of neutron irradiated cold worked Al-6063 alloy

    Energy Technology Data Exchange (ETDEWEB)

    Munitz, A.; Shtechman, A.; Cotler, C.; Dahan, S. [Nuclear Res. Center-Negev, Beer-Sheva (Israel); Talianker, M. [Ben-Gurion Univ., Beer-Sheva (Israel). Dept. of Materials Science

    1998-01-01

    The impact of neutron irradiation on the mechanical properties and fracture morphology of cold worked Al-6063 were studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens (50 mm long and 6 mm wide gauge sections) were punched out from an Al-6063 23% cold worked tubes, which had been exposed to prolonged neutron irradiation of up to 4.5 x 10{sup 25} thermal neutrons/m{sup 2} (E < 0.625 eV). The temperature ranged between 41 and 52 C. The tensile specimens were then tensioned till fracture in an Instron tensiometer with strain rate of 2 x 10{sup -3} s{sup -1}. The uniform elongation and the ultimate tensile strength increase as functions of fluence. Metallographic examination and fractography reveal a decrease in the local area reduction of the final fracture necking. This reduction is accompanied with a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. No voids could be observed up to the maximum fluence. The dislocation density of cold worked Al decreases with the thermal neutron fluence. Prolonged annealing of unirradiated cold worked Al-6063 at 52 C revealed similar results. It thus appears that under our irradiation conditions the temperature during irradiation is the major factor influencing the mechanical properties and the microstructure during irradiation. (orig.). 23 refs.

  9. Nano-scale chemical evolution in a proton-and neutron-irradiated Zr alloy

    Science.gov (United States)

    Harte, Allan; Topping, M.; Frankel, P.; Jädernäs, D.; Romero, J.; Hallstadius, L.; Darby, E. C.; Preuss, M.

    2017-04-01

    Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr)2 and Zr2(Fe,Ni). This is accomplished through ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr)2, predominantly from the edge region, and homogeneously in the case of Zr2(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr2(Fe,Ni) SPP with respect to the Zr(Fe,Cr)2. As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed.

  10. Mechanical properties and microstructure of neutron irradiated cold worked Al-6063 alloy

    Science.gov (United States)

    Munitz, A.; Shtechman, A.; Cotler, C.; Talianker, M.; Dahan, S.

    1998-01-01

    The impact of neutron irradiation on the mechanical properties and fracture morphology of cold worked Al-6063 were studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens (50 mm long and 6 mm wide gauge sections) were punched out from an Al-6063 23% cold worked tubes, which had been exposed to prolonged neutron irradiation of up to 4.5 × 10 25 thermal neutrons/m 2 ( E < 0.625 eV). The temperature ranged between 41 and 52°C. The tensile specimens were then tensioned till fracture in an Instron tensiometer with strain rate of 2 × 10 -3 s -1. The uniform elongation and the ultimate tensile strength increase as functions of fluence. Metallographic examination and fractography reveal a decrease in the local area reduction of the final fracture necking. This reduction is accompanied with a morphology transition from ductile transgranular shear rupture to a combination of transgranular shear with intergranular dimpled rupture. The intergranular rupture area increases with fluence. No voids could be observed up to the maximum fluence. The dislocation density of cold worked Al decreases with the thermal neutron fluence. Prolonged annealing of unirradiated cold worked Al-6063 at 52°C revealed similar results. It thus appears that under our irradiation conditions the temperature during irradiation is the major factor influencing the mechanical properties and the microstructure during irradiation.

  11. Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy

    Science.gov (United States)

    Field, Kevin G.; Briggs, Samuel A.; Hu, Xunxiang; Yamamoto, Yukinori; Howard, Richard H.; Sridharan, Kumar

    2017-01-01

    FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental compatibility compared with more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300 and 400 °C have shown post-irradiation microstructures containing dislocation loops and a Cr-rich α‧ phase. Although these initial studies established the post-irradiation microstructures, there was little to no focus on understanding the influence of pre-irradiation microstructures on this response. In this study, a well-annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 displacements per atom (dpa) at 382 °C and then the effect of random high-angle grain boundaries on the spatial distribution and size of a dislocation loops, a/2 dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with a/2 dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a dislocation loops exhibiting an increased size in the vicinity of the grain boundary. These results suggest the importance of the pre-irradiation microstructure and, specifically, defect sink density spacing to the radiation tolerance of FeCrAl alloys.

  12. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    Science.gov (United States)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  13. Effect of fission neutron irradiation on the tensile and electrical properties of copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Institute, St. Petersburg (Russian Federation); Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

    The objective of this study is to evaluate the properties of several copper alloys following fission reactor irradiation at ITER-relevant temperatures of 80 to 200{degrees}C. This study provides some of the data needed for the ITER research and development Task T213. These low temperature irradiations caused significant radiation hardening and a dramatic decrease in the work hardening ability of copper and copper alloys. The uniform elongation was higher at 200{degree}C compared to 100{degree}C, but still remained below 1% for most of the copper alloys.

  14. Database on Performance of Neutron Irradiated FeCrAl Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Briggs, Samuel A. [Univ. of Wisconsin, Madison, WI (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    The present report summarizes and discusses the database on radiation tolerance for Generation I, Generation II, and commercial FeCrAl alloys. This database has been built upon mechanical testing and microstructural characterization on selected alloys irradiated within the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) up to doses of 13.8 dpa at temperatures ranging from 200°C to 550°C. The structure and performance of these irradiated alloys were characterized using advanced microstructural characterization techniques and mechanical testing. The primary objective of developing this database is to enhance the rapid development of a mechanistic understanding on the radiation tolerance of FeCrAl alloys, thereby enabling informed decisions on the optimization of composition and microstructure of FeCrAl alloys for application as an accident tolerant fuel (ATF) cladding. This report is structured to provide a brief summary of critical results related to the database on radiation tolerance of FeCrAl alloys.

  15. Embrittlement behaviour of low-activation alloys with reduced boron content after neutron irradiation

    Science.gov (United States)

    Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2003-09-01

    Ferritic/martensitic steels for fusion applications have been irradiated up to 2.4 dpa in the Petten high flux reactor (HFR); their embrittlement behaviour was investigated by instrumented Charpy-V tests with subsize specimens. The aim of this mid-dose range programme was a comparison of low-activation alloys subjected to different heat treatments and with reduced B contents (down to 2 wt ppm). In the present report, the results of different OPTIFER alloys (Ia, II, IV, V, VI), as obtained in Phases IA and IB of the HFR-irradiation programme (2.4 dpa, at 250-450 °C), are analysed and assessed in comparison to the results of the former MANITU programme. The evaluation clearly shows the eliminated embrittlement problem for the advanced European reduced-activation alloys in comparison to international reference steels. This improvement can be clearly correlated to the reduction of the boron content. Furthermore, the influence of different heat treatments on the impact properties is presented.

  16. Impact property of low-activation vanadium alloy after laser welding and heavy neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Takuya, E-mail: nagasaka@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Muroga, Takeo [National Institute for Fusion Science, Toki, Gifu (Japan); The Graduate University for Advanced Studies, Toki, Gifu (Japan); Watanabe, Hideo [Research Institute for Applied Mechanics, Kyushu University, Kasuga (Japan); Miyazawa, Takeshi [The Graduate University for Advanced Studies, Toki, Gifu (Japan); Yamazaki, Masanori [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki (Japan); Shinozaki, Kenji [Department of Mechanical System Engineering, Graduate School of Engineering, Hiroshima University, Higashi Hiroshima (Japan)

    2013-11-15

    Weld specimens of the reference low activation vanadium alloy, NIFS-HEAT-2, were irradiated up to a neutron fluence of 1.5 × 10{sup 25} n m{sup −2} (E > 0.1 MeV) (1.2 dpa) at 670 K and 1.3 × 10{sup 26} n m{sup −2} (5.3 dpa) at 720 K in the JOYO reactor in Japan. The base metal exhibited superior irradiation resistance with the ductile-to-brittle transition temperature (DBTT) much lower than room temperature (RT) for both irradiation conditions. The weld metal kept the DBTT below RT after the 1.2 dpa irradiation; however, it showed enhanced irradiation embrittlement with much higher DBTT than RT after the 5.3 dpa irradiation. The high DBTT for the weld metal was effectively recovered by a post-irradiation annealing at 873 K for 1 h. Mechanisms of the irradiation embrittlement and its recovery are discussed, based on characterization of the radiation defects and irradiation-induced precipitation.

  17. Standard Specification for Nickel-Chromium-Molybdenum-Columbium Alloy (UNS N06625), Nickel-Chromium-Molybdenum-Silicon Alloy (UNS N06219), and Nickel-Chromium-Molybdenum-Tungsten Alloy (UNS N06650) Rod and Bar

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2014-01-01

    Standard Specification for Nickel-Chromium-Molybdenum-Columbium Alloy (UNS N06625), Nickel-Chromium-Molybdenum-Silicon Alloy (UNS N06219), and Nickel-Chromium-Molybdenum-Tungsten Alloy (UNS N06650) Rod and Bar

  18. Initiation and propagation of cleared channels in neutron-irradiated pure copper and a precipitation hardened CuCrZr alloy

    DEFF Research Database (Denmark)

    Singh, B.N; Edwards, D.J.; Bilde-Sørensen, Jørgen

    2004-01-01

    has emerged. Recently we have studied the problem of initiation and propagation of cleared channels during post-irradiation tensile tests of pure copper and a copper alloy irradiated with fission neutrons.Tensile specimens of pure copper and a precipitation hardened copper alloy (CuCrZr) were neutron...... irradiated at 323 and 373K to displacement doses in the range of 0.01 to 0.3 dpa (displacement per atom) and tensile tested at the irradiation temperature.The stress-strain curves clearly indicated the occurrence of a yield drop. The post-deformation microstructural examinations revealed that the channels...... throughout the whole tensile test, no clear evidencehas been found for the operation of Frank-Read sources in the volume between the channels. Channels have been observed to penetrate through annealing twins, in some cases stopping at the opposite twin boundary and in other cases penetrating even throughthe...

  19. Alloy hardening and softening in binary molybdenum alloys as related to electron concentration

    Science.gov (United States)

    Stephens, J. R.; Witzke, W. R.

    1972-01-01

    An investigation was conducted to determine the effects of alloy additions of hafnium, tantalum, tungsten, rhenium, osmium, iridium, and platinum on hardness of molybdenum. Special emphasis was placed on alloy softening in these binary molybdenum alloys. Results showed that alloy softening was produced by those elements having an excess of s+d electrons compared to molybdenum, while those elements having an equal number or fewer s+d electrons that molybdenum failed to produce alloy softening. Alloy softening and alloy hardening can be correlated with the difference in number of s+d electrons of the solute element and molybdenum.

  20. Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M.

    1997-04-01

    Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

  1. Hot rolling of thick uranium molybdenum alloys

    Science.gov (United States)

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  2. The Densification of Molybdenum and Molybdenum Alloy Powders Using Hot Isostatic Pressing.

    Science.gov (United States)

    1985-08-01

    TECHNICAL REPORT ARLCB-TR-85025 00 THE DENSIFICATION OF MOLYBDENUM (n AND MOLYBDENUM ALLOY POWDERS USING HOT ISOSTATIC PRESSING J. BARRANCO I. AHMAD S...ISOSTATIC PRESSING Final 6. PERFORMING ORG. REPORT NUMBER 7. AUTHOR(o) . CONTRACT OR GRANT NUMBER(e) J. Barranco , I. Ahmad, S. Isserow, and R. Warenchak

  3. Radiation resistance and parameters of activation of aluminium-magnesium-scandium and aluminium-magnesium-vanadium alloys under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, L.I.; Ivanov, V.V.; Lazorenko, V.M.; Platov, Yu.M.; Tovtin, V.I.; Toropova, L.S. (A.A. Baikov Inst. of Metallurgy, Academy of Sciences, Moscow (Russia))

    1992-09-01

    Alloys Al-2.24Mg-0.223Sc-0.04Zr, Al-2.24Mg-0.12Sc-0.04Zr, and Al-2.24Mg-0.05V (at.%) annealed at 150deg C and 400deg C were irradiated at [approx equal] 70 and [approx equal] 150deg C in the SM-2 reactor. The maximum neutron fluence was 4.7x10[sup 24] m[sup -2] (E > 0.1 MeV). The tensile tests were carried out in the temperature range 20 to 350deg C. Alloy Al-2.24 Mg-0.23Sc-0.04Zr annealed at 400deg C and alloy Al-2.24Mg-0.12Sc-0.04Zr annealed at 150deg C at all test temperatures retained good mechanical properties after irradiation. The mechanisms for the radiation resistance of aluminium-scandium and aluminium-magnesium-scandium alloys are discussed. Calculations of induced radioactivity and its decay behaviour after shutdown in aluminium and Al-2.24Mg-(0.12-0.23)Sc alloys were carried out. Composition of the radionuclides in these materials after irradiation in the SM-2 reactor were also determined using a gamma-spectroscopy technique. (orig.).

  4. Radiation resistance and parameters of activation of aluminium-magnesium-scandium and aluminium-magnesium-vanadium alloys under neutron irradiation

    Science.gov (United States)

    Ivanov, L. I.; Ivanov, V. V.; Lazorenko, V. M.; Platov, Yu. M.; Tovtin, V. I.; Toropova, L. S.

    1992-09-01

    Alloys Al2.24Mg0.23Sc0.04Zr, Al2.24Mg0.12Sc0.04Zr, and Al2.24Mg0.05V (at.)) annealed at 150°C and 400°C were irradiated ≈70 and ≈150°C in the SM-2 reactor. The maximum neutron fluence was 4.7×1024 m-2 (E > 0.1 MeV). The tensile tests were carried out in the temperature range 20 to 350°C. Alloy Al2.24Mg0.23Sc0.04Zr annealed at 400°C and alloy Al2.24Mg0.12Sc0.04Zr annealed at 150°C at all test temperature, retained good mechanical properties after irradiation. The mechanisms for the radiation resistance of aluminiumscandium and aluminiummagnesiumscandium alloys are discussed. Calculations of induced radioactivity and its decay behaviour after shutdown in aluminium and Al2.24Mg(0.12-0.23)Sc alloys were carried out. Composition of the radionuclides in these materials after irradiation in the SM-2 reactor were also determined using a gamma-spectroscopy technique.

  5. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  6. Effects of neutron irradiation on microstructure and deformation behaviour of mono- and polycrystalline molybdenum and its alloys

    DEFF Research Database (Denmark)

    Singh, B.N.; Evans, J.H.; Horsewell, A.

    1998-01-01

    specimens were tensile tested at 295 K. Post-irradiation microstructures were quantitatively characterized using a transmission electron microscope (TEM). Fracture surfaces were examined in a scanning electron microscope (SEM). The results of tensile testing as well as of transmission and scanning...

  7. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  8. Posterior magnetic effect on the pure and doped Fe-Ni alloy under neutron irradiation; Efeito magnetico posterior na liga Fe-Ni pura e dopada, sob irradiacao neutronica

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Iris

    1974-07-01

    Polycrystalline specimens of unirradiated and neutron irradiated Fe-Ni alloys have been studied in the temperature range RT - 500 deg C. The study was carried out in pure (50-50) as well as in Si, A1, Cr and Mo doped samples. Initial magnetic permeability was measured in unirradiated (virgin)and in neutron irradiated samples, during isochronal and linear thermal treatments. The main results are: a magnetic After Effect (MAE) is detected in the temperature range 370 deg C - Tc, where Tc is the Curie Temperature. In this range an activation energy of 3.2 {+-} 0.2 eV was determined for the Cr doped Fe-Ni alloy (impurity content: 0.1%); measurements made in the irradiated samples, during a linear temperature treatment, show the existence of several MAE zones in the temperature range RT - Tc. The isochronal annealing experiments show that these MAE zones are accompanied by a decrease in the room temperature value of the magnetic permeability, for zones between RT and a certain temperature T{sub 1}. Above this range there is a steep increase in the room temperature permeability. Activation energies were determined for pure and Mo-doped (0.1%) samples for the first MAE zone (50 deg C - 120 deg C). The values obtained 1.25 - 0.08 eV and 1.42 {+-} 0.09 eV, respectively; the impurity - doped samples show a different behaviour relative to the pure ones: samples with low impurity content (0.1% and 0.5% of Si, Al or Mo) present an enhancement in the amplitude and also an overlapping of the diffusion stages. On the other hand, samples with higher impurity content (2 and 4% of Mo) show a decrease in these amplitudes. (author)

  9. Standard Specification for Low-Carbon Nickel-Chromium-Molybdenum, Low-Carbon Nickel-Chromium-Molybdenum-Copper, Low-Carbon Nickel-Chromium-Molybdenum-Tantalum, Low-Carbon Nickel-Chromium-Molybdenum-Tungsten, and Low-Carbon Nickel-Molybdenum-Chromium Alloy Plate, Sheet, and Strip

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2015-01-01

    Standard Specification for Low-Carbon Nickel-Chromium-Molybdenum, Low-Carbon Nickel-Chromium-Molybdenum-Copper, Low-Carbon Nickel-Chromium-Molybdenum-Tantalum, Low-Carbon Nickel-Chromium-Molybdenum-Tungsten, and Low-Carbon Nickel-Molybdenum-Chromium Alloy Plate, Sheet, and Strip

  10. Standard Specification for Low-Carbon Nickel-Chromium-Molybdenum, Low-Carbon Nickel-Molybdenum-Chromium, Low-Carbon Nickel-Molybdenum-Chromium-Tantalum, Low-Carbon Nickel-Chromium-Molybdenum-Copper, and Low-Carbon Nickel-Chromium-Molybdenum-Tungsten Alloy Rod

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2015-01-01

    Standard Specification for Low-Carbon Nickel-Chromium-Molybdenum, Low-Carbon Nickel-Molybdenum-Chromium, Low-Carbon Nickel-Molybdenum-Chromium-Tantalum, Low-Carbon Nickel-Chromium-Molybdenum-Copper, and Low-Carbon Nickel-Chromium-Molybdenum-Tungsten Alloy Rod

  11. Role of electron concentration in softening and hardening of ternary molybdenum alloys

    Science.gov (United States)

    Stephens, J. R.; Witzke, W. R.

    1975-01-01

    Effects of various combinations of hafnium, tantalum, rhenium, osmium, iridium, and platinum in ternary molybdenum alloys on alloy softening and hardening were determined. Hardness tests were conducted at four test temperatures over the temperature range 77 to 411 K. Results showed that hardness data for ternary molybdenum alloys could be correlated with anticipated results from binary data based upon expressions involving the number of s and d electrons contributed by the solute elements. The correlation indicated that electron concentration plays a dominant role in controlling the hardness of ternary molybdenum alloys.

  12. Microstructures and Hardness/Wear Performance of High-Carbon Stellite Alloys Containing Molybdenum

    Science.gov (United States)

    Liu, Rong; Yao, J. H.; Zhang, Q. L.; Yao, M. X.; Collier, Rachel

    2015-12-01

    Conventional high-carbon Stellite alloys contain a certain amount of tungsten which mainly serves to provide strengthening to the solid solution matrix. These alloys are designed for combating severe wear. High-carbon molybdenum-containing Stellite alloys are newly developed 700 series of Stellite family, with molybdenum replacing tungsten, which are particularly employed in severe wear condition with corrosion also involved. Three high-carbon Stellite alloys, designated as Stellite 706, Stellite 712, and Stellite 720, with different carbon and molybdenum contents, are studied experimentally in this research, focusing on microstructure and phases, hardness, and wear resistance, using SEM/EDX/XRD techniques, a Rockwell hardness tester, and a pin-on-disk tribometer. It is found that both carbon and molybdenum contents influence the microstructures of these alloys significantly. The former determines the volume fraction of carbides in the alloys, and the latter governs the amount of molybdenum-rich carbides precipitated in the alloys. The hardness and wear resistance of these alloys are increased with the carbide volume fraction. However, with the same or similar carbon content, high-carbon CoCrMo Stellite alloys exhibit worse wear resistance than high-carbon CoCrW Stellite alloys.

  13. Neutron-Induced Microstructural Evolution of Fe-15Cr-16Ni Alloys at ~400 C During Neutron Irradiation in the FFTF Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.; Wolfer, W. G.; Isobe, Yoshihiro

    2001-06-30

    An experiment conducted at ~400 degrees C on simple model austenitic alloys (Fe-15Cr-16Ni and Fe-15Cr-16Ni-0.25Ti, both with and without 500 appm boron) irradiated in the FFTF fast reactor at seven different dpa rates clearly shows that lowering of the atomic displacement rate leads to a pronounced reduction in the transient regime of void swelling. While the steady state swelling rate (~1%/dpa) of these alloys is unaffected by changes in the dpa rate, the transient regime of swelling can vary from <1 to ~60 dpa when the dpa rate varies over more than two orders of magnitude. This range of dpa rates covers the full span of fusion, PWR and fast reactor rates. The origin of the flux sensitivity of swelling arises first in the evolution of the Frank dislocation loop population, its unfaulting, and the subsequent evolution of the dislocation network. There also appears to be some flux sensitivity to the void nucleation process. Most interestingly, the addition of titanium suppresses the void nucleation process somewhat, but does not alter the duration of the transient regime of swelling or its sensitivity to dpa rate. Side-by-side irradiation of boron-modified model alloys in this same experiment shows that higher helium generation rates homogenize the swelling somewhat, but do not significantly change its magnitude or flux sensitivity. The results of this study support the prediction that austenitic alloys irradiated at PWR-relevant displacement rates will most likely swell more than when irradiated at higher rates characteristic of fast reactors. Thus, the use of swelling data accumulated in fast reactors may possibly lead to an under-prediction of swelling in lower-flux PWRs and fusion devices.

  14. Formation and evolution of MnNi clusters in neutron irradiated dilute Fe alloys modelled by a first principle-based AKMC method

    Energy Technology Data Exchange (ETDEWEB)

    Ngayam-Happy, R. [EDF-R and D, Departement Materiaux et Mecanique des Composants (MMC), Les Renardieres, F-77818 Moret sur Loing Cedex (France); Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Becquart, C.S., E-mail: charlotte.becquart@univ-lille1.fr [Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France); Domain, C. [EDF-R and D, Departement Materiaux et Mecanique des Composants (MMC), Les Renardieres, F-77818 Moret sur Loing Cedex (France); Unite Materiaux et Transformations (UMET), UMR CNRS 8207, Universite de Lille 1, ENSCL, F-59655 Villeneuve d' Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modelisation des Microstructures pour le Vieillissement des Materiaux (EM2VM) (France)

    2012-07-15

    An atomistic Monte Carlo model parameterised on electronic structure calculations data has been used to study the formation and evolution under irradiation of solute clusters in Fe-MnNi ternary and Fe-CuMnNi quaternary alloys. Two populations of solute rich clusters have been observed, which can be discriminated by whether or not the solute atoms are associated with self-interstitial clusters. Mn-Ni-rich clusters are observed at a very early stage of the irradiation in both modelled alloys, whereas the quaternary alloys contain also Cu-containing clusters. Mn-Ni-rich clusters nucleate very early via a self-interstitial-driven mechanism, earlier than Cu-rich clusters; the latter, however, which are likely to form via a vacancy-driven mechanism, grow in number much faster than the former, helped by the thermodynamic driving force to Cu precipitation in Fe, thereby becoming dominant in the low dose regime. The kinetics of the number density increase of the two populations is thus significantly different. Finally the main conclusion suggested by this work is that the so-called late blooming phases might as well be neither late, nor phases.

  15. Experimental investigation of high He/dpa microstructural effects in neutron irradiated B-alloyed Eurofer97 steel by means of small angle neutron scattering (SANS and electron microscopy

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2016-12-01

    Full Text Available High He/dpa microstructural effects have been investigated, by means of small-angle neutron scattering (SANS and transmission electron microscopy (TEM, in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2V, 1.08W wt%, B contents variable between 10 and 1000ppm, neutron irradiated at the High Flux Reactor of the JRC-Petten at temperatures between 250 °C and 450 °C, up do a dose level of 16 dpa. Under these irradiation parameters, B activation is expected to produce corresponding helium contents variable between 80 and 5600appm, with helium bubble distributions relevant for the technological applications. The SANS measurements were carried out under magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. Increasing the estimated helium content from 400 to 5600appm, the analysis of the SANS cross-sections yields an increase in the volume fraction, attributed to helium bubbles, of almost one order of magnitude (from 0.007 to 0.038; furthermore, the difference between nuclear and magnetic SANS components is strongly reduced. These results are discussed in correlation with TEM observations of the same samples and are tentatively attributed to the effect of drastic microstructural changes in Eurofer97-1 for high He/dpa ratio values, possibly relating to the dissolution of large B-carbides due to transmutation reactions.

  16. Unusual response of the binary V-2Si alloy to neutron irradiation in FFTF at 430-600{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Ohnuki, S.; Konoshita, H.; Takahaski, H. [Hokkaido Univ., Sapparo (Japan); Garner, F.A. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-04-01

    When V-2Si was irradiated in FFTF at 430, 500 and 600C to doses as high as 80 dpa, a very unusual swelling response was observed in which the swelling appeared to saturate rather quickly at {approx}35% at 430 and 540C, but approached this swelling same level much more slowly at 600C. The possible causes of this phenomenon are discussed as well as the implications of these findings on the swelling behavior of other high swelling vanadium binary alloys.

  17. Acidic ammonothermal growth of gallium nitride in a liner-free molybdenum alloy autoclave

    Science.gov (United States)

    Malkowski, Thomas F.; Pimputkar, Siddha; Speck, James S.; DenBaars, Steven P.; Nakamura, Shuji

    2016-12-01

    This paper discusses promising materials for use as internal, non-load bearing components as well as molybdenum-based alloys for autoclave structural components for an ammonothermal autoclave. An autoclave was constructed from the commercial titanium-zirconium-molybdenum (TZM) alloy and was found to be chemically inert and mechanically stable under acidic ammonothermal conditions. Preliminary seeded growth of GaN was demonstrated with negligible incorporation of transition metals (including molybdenum) into the grown material (560 °C). The possibility of a 'universal', inexpensive, liner-free ammonothermal autoclave capable of exposure to basic and acidic chemistry is demonstrated.

  18. The in vitro toxicity of cobalt-chrome-molybdenum alloy and its constituent metals.

    Science.gov (United States)

    Evans, E J; Thomas, I T

    1986-01-01

    Cobalt-chrome-molybdenum alloys are widely used in orthopaedic implants. Although they are relatively well tolerated, complications (including loosening and tissue necrosis) still occur and sometimes appear to be due to incomplete biocompatibility of the alloy. To investigate the local effect of the alloy on cells derived from the musculo-skeletal system, primary lines of fibroblastic cells from newborn rats were exposed to powders of cobalt-chrome-molybdenum alloy and its main constituents cobalt, chromium nickel and molybdenum. The toxicity of the metals was determined by counts of total cell number and of abnormal cells at intervals from 2 to 12 d. The alloy was much less toxic than cobalt or nickel and the pattern of toxicity was different for each metal. The results emphasize the difficulty of devising a single tissue culture test of toxicity which will measure the toxicity of any potential implant material.

  19. Creep-fatigue deformation behaviour of OFHC-copper and CuCrZr alloy with different heat treatments and with and without neutron irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Li, M.; Stubbins, J.F.

    2005-01-01

    The creep-fatigue interaction behaviour of a precipitation hardened CuCrZr alloy was investigated at 295 and 573 K. To determine the effect of irradiation a number of fatigue specimens were irradiated at 333 and 573 K to a dose level in the range of 0.2 -0.3 dpa and were tested at room temperature...... with a frequency of 0.5 Hz. Holdtimes of up to 1000 seconds were used. Creep-fatigue experiments were carried out using strain, load and extension controlled modesof cyclic loading. In addition, a number of “interrupted” creep-fatigue tests were performed on the prime aged CuCuZr specimens in the strain controlled...

  20. Aqueous corrosion behavior of uranium-molybdenum alloys

    Science.gov (United States)

    Gardner, Levi D.

    Nuclear fuel characterization requires understanding of the various conditions to which materials are exposed in-reactor. One of these important conditions is corrosion, particularly that of fuel constituents. Therefore, corrosion behavior is of special interest and an essential part of nuclear materials characterization efforts. In support of the Office of Material Management and Minimization's Reactor Conversion Program, monolithic uranium-10 wt% molybdenum alloy (U-Mo) is being investigated as a low enriched uranium alternative to highly enriched uranium dispersion fuel currently used in domestic high performance research reactors. The aqueous corrosion behavior of U-Mo is being examined at Pacific Northwest National Laboratory (PNNL) as part of U-Mo fuel fabrication capability activity. No prior study adequately represents this behavior given the current state of alloy composition and thermomechanical processing methods, and research reactor water chemistry. Two main measurement techniques were employed to evaluate U-Mo corrosion behavior. Low-temperature corrosion rate values were determined by means of U-Mo immersion testing and subsequent mass-loss measurements. The electrochemical behavior of each processing condition was also qualitatively examined using the techniques of corrosion potential and anodic potentiodynamic polarization. Scanning electron microscopy (SEM) and optical metallography (OM) imagery and hardness measurements provided supplemental corrosion analysis in an effort to relate material corrosion behavior to processing. The processing effects investigated as part of this were those of homogenization heat treatment (employed to mitigate the effects of coring in castings) and sub-eutectoid heat treatment, meant to represent additional steps in fabrication (such as hot isostatic pressing) performed at similar temperatures. Immersion mass loss measurements and electrochemical results both showed very little appreciable difference between

  1. Standard Specification for Pressure Consolidated Powder Metallurgy Iron-Nickel-Chromium-Molybdenum (UNS N08367), Nickel-Chromium-Molybdenum-Columbium (Nb) (UNS N06625), Nickel-Chromium-Iron Alloys (UNS N06600 and N06690), and Nickel-Chromium-Iron-Columbium-Molybdenum (UNS N07718) Alloy Pipe Flanges, Fittings, Valves, and Parts

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2015-01-01

    Standard Specification for Pressure Consolidated Powder Metallurgy Iron-Nickel-Chromium-Molybdenum (UNS N08367), Nickel-Chromium-Molybdenum-Columbium (Nb) (UNS N06625), Nickel-Chromium-Iron Alloys (UNS N06600 and N06690), and Nickel-Chromium-Iron-Columbium-Molybdenum (UNS N07718) Alloy Pipe Flanges, Fittings, Valves, and Parts

  2. Simple spectrophotometric method for determination of zirconium or hafnium in selected molybdenum-base alloys.

    Science.gov (United States)

    Dupraw, W A

    1972-06-01

    A simple analytical procedure is described for determining zirconium or hafnium in molybdenum-base alloys by formation of the Arsenazo III complex of zirconium or hafnium in 9 M hydrochloric acid medium. The absorbance is measured at 670 nm. Molybdenum (10 mg), titanium (1 mg), and rhenium (10 mg) have no adverse effect. No prior separation is needed. The relative standard deviation is 1.3-2.7%.

  3. REDUCTION AND CONSOLIDATION OF SUPERIOR QUALITY MOLYBDENUM ALLOYS

    Science.gov (United States)

    diameter bomb. Techniques were developed for the electron beam melting of hydrogen-reduced molybdenum powder. Although this material contains low... beam melting of the thermitically-reduced molybdenum presented difficulties; primarily because of the melting configuration. The use of a remote...interstitial elements, it exhibited severe grain boundary brittleness. Physical properties of these single crystals are being determined. The electron

  4. Thermal annealing in neutron-irradiated tribromobenzenes

    DEFF Research Database (Denmark)

    Siekierska, K.E.; Halpern, A.; Maddock, A. G.

    1968-01-01

    The distribution of 82Br among various products in neutron-irradiated isomers of tribromobenzene has been investigated, and the effect of thermal annealing examined. Reversed-phase partition chromatography was employed for the determination of radioactive organic products, and atomic bromine...

  5. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    Science.gov (United States)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  6. Investigation of welding and brazing of molybdenum and TZM alloy tubes

    Science.gov (United States)

    Lundblad, Wayne E.

    1991-01-01

    This effort involved investigating the welding and brazing techniques of molybdenum tubes to be used as cartridges in the crystal growth cartridge. Information is given in the form of charts and photomicrographs. It was found that the recrystallization temperature of molybdenum can be increased by alloying it with 0.5 percent titanium and 0.1 percent zirconium. Recrystallization temperatures for this alloy, known as TZM, become significant around 2500 F. A series of microhardness tests were run on samples of virgin and heat soaked TZM. The test results are given in tabular form. It was concluded that powder metallurgy TZM may be an acceptable cartridge material.

  7. PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R; Ann Visser, A; James Laurinat, J

    2007-10-15

    The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum.

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

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

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

  9. Innovative Molybdenum Alloy for Extreme Operating Conditions Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Molybdenum has been identified as a promising material for many high temperature NASA applications due to its high melting temperature, resistance to liquid metals,...

  10. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET

    Energy Technology Data Exchange (ETDEWEB)

    Visser, A; Robert Pierce, R

    2007-06-07

    H-Canyon Engineering requested the Savannah River National Laboratory (SRNL) to perform two solvent extraction experiments using dissolved Super Kukla (SK) material. The SK material is an uranium (U)-molybdenum (Mo) alloy material of 90% U/10% Mo by weight with 20% 235U enrichment. The first series of solvent extraction tests involved a series of batch distribution coefficient measurements with 7.5 vol % tributylphosphate (TBP)/n-paraffin for extraction from 4-5 M nitric acid (HNO{sub 3}), using 4 M HNO{sub 3}-0.02 M ferrous sulfamate (Fe(SO3NH2)2) scrub, 0.01 M HNO3 strip steps with particular emphasis on the distribution of U and Mo in each step. The second set of solvent extraction tests determined whether the 2.5 wt % sodium carbonate (Na2CO3) solvent wash change frequency would need to be modified for the processing of the SK material. The batch distribution coefficient measurements were performed using dissolved SK material diluted to 20 g/L (U + Mo) in 4 M HNO{sub 3} and 5 M HNO{sub 3}. In these experiments, U had a distribution coefficient greater than 2.5 while at least 99% of the nickel (Ni) and greater than 99.9% of the Mo remained in the aqueous phase. After extraction, scrub, and strip steps, the aqueous U product from the strip contains nominally 7.48 {micro}g Mo/g U, significantly less than the maximum allowable limit of 800 {micro}g Mo/g U. Solvent washing experiments were performed to expose a 2.5 wt % Na2CO3 solvent wash solution to the equivalent of 37 solvent wash cycles. The low Mo batch distribution coefficient in this solvent extraction system yields only 0.001-0.005 g/L Mo extracted to the organic. During the solvent washing experiments, the Mo appears to wash from the organic.

  11. Kinetic Monte Carlo modelling of neutron irradiation damage in iron

    Energy Technology Data Exchange (ETDEWEB)

    Gamez, L. [Instituto de Fusion Nuclear, UPM, Madrid (Spain); Departamento de Fisica Aplicada, ETSII, UPM, Madrid (Spain)], E-mail: linarejos.gamez@upm.es; Martinez, E. [Instituto de Fusion Nuclear, UPM, Madrid (Spain); Lawrence Livermore National Laboratory, LLNL, CA 94550 (United States); Perlado, J.M.; Cepas, P. [Instituto de Fusion Nuclear, UPM, Madrid (Spain); Caturla, M.J. [Departamento de Fisica Aplicada, Universidad de Alicante, Alicante (Spain); Victoria, M. [Instituto de Fusion Nuclear, UPM, Madrid (Spain); Marian, J. [Lawrence Livermore National Laboratory, LLNL, CA 94550 (United States); Arevalo, C. [Instituto de Fusion Nuclear, UPM, Madrid (Spain); Hernandez, M.; Gomez, D. [CIEMAT, Madrid (Spain)

    2007-10-15

    Ferritic steels (FeCr based alloys) are key materials needed to fulfill the requirements expected in future nuclear fusion facilities, both for magnetic and inertial confinement, and advanced fission reactors (GIV) and transmutation systems. Research in such field is actually a critical aspect in the European research program and abroad. Experimental and multiscale simulation methodologies are going hand by hand in increasing the knowledge of materials performance. At DENIM, it is progressing in some specific part of the well-linked simulation methodology both for defects energetics and diffusion, and for dislocation dynamics. In this study, results obtained from kinetic Monte Carlo simulations of neutron irradiated Fe under different conditions are presented, using modified ad hoc parameters. A significant agreement with experimental measurements has been found for some of the parameterization and mechanisms considered. The results of these simulations are discussed and compared with previous calculations.

  12. First-principles studies of chromium line-ordered alloys in a molybdenum disulfide monolayer

    Science.gov (United States)

    Andriambelaza, N. F.; Mapasha, R. E.; Chetty, N.

    2017-08-01

    Density functional theory calculations have been performed to study the thermodynamic stability, structural and electronic properties of various chromium (Cr) line-ordered alloy configurations in a molybdenum disulfide (MoS2) hexagonal monolayer for band gap engineering. Only the molybdenum (Mo) sites were substituted at each concentration in this study. For comparison purposes, different Cr line-ordered alloy and random alloy configurations were studied and the most thermodynamically stable ones at each concentration were identified. The configurations formed by the nearest neighbor pair of Cr atoms are energetically most favorable. The line-ordered alloys are constantly lower in formation energy than the random alloys at each concentration. An increase in Cr concentration reduces the lattice constant of the MoS2 system following the Vegard’s law. From density of states analysis, we found that the MoS2 band gap is tunable by both the Cr line-ordered alloys and random alloys with the same magnitudes. The reduction of the band gap is mainly due to the hybridization of the Cr 3d and Mo 4d orbitals at the vicinity of the band edges. The band gap engineering and magnitudes (1.65 eV to 0.86 eV) suggest that the Cr alloys in a MoS2 monolayer are good candidates for nanotechnology devices.

  13. Spectrographic analysis of uranium-molybdenum alloys; Analisis espectrografico de aleaciones uranio-molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Roca, M.

    1967-07-01

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO{sub 3}. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO{sub 3}. (Author) 5 refs.

  14. Nanostructures obtained from a mechanically alloyed and heat treated molybdenum carbide

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Barriga Arceo, L. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico) and ESIQIE-UPALM, IPN Apdo Postal 118-395, C.P. 07051 D.F. Mexico (Mexico)]. E-mail: luchell@yahoo.com; Orozco, E. [Instituto de Fisica UNAM, Apdo Postal 20-364, C.P. 01000 D.F. Mexico (Mexico)]. E-mail: eorozco@fisica.unam.mx; Mendoza-Leon, H. [ESIQIE-UPALM, IPN Apdo Postal 118-395, C.P. 07051 D.F. Mexico (Mexico)]. E-mail: luchell@yahoo.com; Palacios Gonzalez, E. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: epalacio@imp.mx; Leyte Guerrero, F. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: fleyte@imp.mx; Garibay Febles, V. [Programa de Ingenieria Molecular, I.M.P. Lazaro Cardenas 152, C.P. 07730 D.F. Mexico (Mexico)]. E-mail: vgaribay@imp.mx

    2007-05-31

    Mechanical alloying was used to prepare molybdenum carbide. Microstructural characterization of samples was performed by X-ray diffraction (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM) methods. Molybdenum carbide was heated at 800 {sup o}C for 15 min in order to produce carbon nanotubes. Nanoparticles of about 50-140 nm in diameter and nanotubes with diameters of about 70-260 nm and 0.18-0.3 {mu}m in length were obtained after heating at 800 {sup o}C, by means of this process.

  15. Effects of low-temperature neutron irradiation on the mechanical properties of BCC metals

    Science.gov (United States)

    Kitajima, K.; Abe, H.; Aono, Y.; Kuramoto, E.; Takamura, S.

    1982-08-01

    Tensile properties, together with the effects of point-irradiation annealings on them, were measured on single crystals of pure iron, iron containing 200 at. ppm carbon, and pure molybdenum, which were irradiated at 5 K in reactor JRR-3 and stored at 77 K, at the test temperatures of 4.2-800 K. Their measurements were compared with those irradiated by 2.5 and 28 MeV electrons at 77 K to elucidate the characteristics of neutron irradiation. Interpretations were then presented for the mechanisms of softening and hardening based on the interactions of defects and defect clusters formed in various annealing stages with screw dislocation in bcc metals.

  16. Design aspects of a cold neutron irradiator

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, A.G.; Clark, D.D.; Hossain, T.Z.; Spern, S.A. [Cornell Univ., Ithaca, NY (United States)

    1995-12-31

    Design work on a cold-neutron irradiator (CNI) is being pursued at Cornell University. Prompt gamma neutron activation analysis (PGNAA) by means of cold neutron absorption is the objective of the CNI. Using cold neutrons instead of thermal neutrons to cause neutron capture in the sample, the CNI is a logical extension of the concept of a thermal neutron irradiator. Since the neutron capture cross section for most nuclei varies as 1/v, augmentation of the neutron capture reaction rate is achieved in the sample by a factor of {approximately}2.3. The statistical precision with which one can measure the mass of a particular element in the sample is enhanced in a CNI, in comparison with a thermal neutron irradiator, by a factor of between 2.3 and the square of 2.3. The exact factor by which the statistical precision is enhanced depends on the energy of the PGNAA photopeak at which one is looking and on the extent to which the photon background measured by the photon detector is dominated by either the {sup 252}Cf spontaneous fission photons or by the neutron capture photons from the CNI structural materials. Within the context of the optimization of the elemental sensitivity of the CNI system, the CNI must efficiently deliver cold neutrons from the {sup 252}Cf fast neutron source to the sample and must efficiently deliver the PGNAA gamma rays of the sample to the high-purity germanium (HPGe) photon detector while maintaining reasonable fast neutron and gamma-ray backgrounds at the detector.

  17. A simple spectrophotometric method for determination of zirconium or hafnium in selected molybdenum-base alloys

    Science.gov (United States)

    Dupraw, W. A.

    1972-01-01

    A simple analytical procedure is described for accurately and precisely determining the zirconium or hafnium content of molybdenum-base alloys. The procedure is based on the reaction of the reagent Arsenazo III with zirconium or hafnium in strong hydrochloric acid solution. The colored complexes of zirconium or hafnium are formed in the presence of molybdenum. Titanium or rhenium in the alloy have no adverse effect on the zirconium or hafnium complex at the following levels in the selected aliquot: Mo, 10 mg; Re, 10 mg; Ti, 1 mg. The spectrophotometric measurement of the zirconium or hafnium complex is accomplished without prior separation with a relative standard deviation of 1.3 to 2.7 percent.

  18. Effects of heat treatments on the thermal diffusivity of Uranium-Molybdenum alloy

    Science.gov (United States)

    Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Pedrosa, T. A.

    2016-07-01

    U-Mo alloys are the most investigated nuclear fuel material to be used in research reactors. The addition of molybdenum stabilizes the gamma phase of uranium and increases its melting point. A research program under development at Nuclear Technology Development Center (CDTN) aims the obtaining of uranium-molybdenum alloys to enable the high enriched uranium (HEU) to low enriched uranium (LEU) conversions. U-Mo ingots with 10% by weight were induction melted and heat treated at 300 °C for 72 h, 120 h and 240 h. Thermal diffusivity was determined by the laser flash method and thermal quadrupole method, from room temperature to 300 oC and 400oC. It was observed that the thermal diffusivity tends to increase with increasing temperature.

  19. Embrittlement behavior of neutron irradiated RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 {sup o}C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 {sup o}C and 300 {sup o}C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 {sup o}C and 350 {sup o}C.

  20. Embrittlement behavior of neutron irradiated RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 °C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 °C and 300 °C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 °C and 350 °C.

  1. Microstructures and oxidation behavior of some Molybdenum based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ray, Pratik Kumar [Iowa State Univ., Ames, IA (United States)

    2011-01-01

    The advent of Ni based superalloys revolutionized the high temperature alloy industry. These materials are capable of operating in extremely harsh environments, comprising of temperatures around 1050 C, under oxidative conditions. Demands for increased fuel efficiency, however, has highlighted the need for materials that can be used under oxidative conditions at temperatures in excess of 1200 C. The Ni based superalloys are restricted to lower temperatures due to the presence of a number of low melting phases that melt in the 1250 - 1450 C, resulting in softening of the alloys above 1000 C. Therefore, recent research directions have been skewed towards exploring and developing newer alloy systems. This thesis comprises a part of such an effort. Techniques for rapid thermodynamic assessments were developed and applied to two different systems - Mo-Si alloys with transition metal substitutions (and this forms the first part of the thesis) and Ni-Al alloys with added components for providing high temperature strength and ductility. A hierarchical approach towards alloy design indicated the Mo-Ni-Al system as a prospective candidate for high temperature applications. Investigations on microstructures and oxidation behavior, under both isothermal and cyclic conditions, of these alloys constitute the second part of this thesis. It was seen that refractory metal systems show a marked microstructure dependence of oxidation.

  2. Current Amplification Characteristics of BJT on Fast Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Sun, Gwang Min; Baek, Hani [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    BJT (Bipolar Junction Transistor) is a three-terminal device with an important feature in that the current through two terminals can be controlled by small changes we make in the current or voltage at the third terminal. This control feature allows us to amplify small AC signals or to switch the device from an on state and off state and back. Fast neutron irradiation incurs lattice damage in bulk Si. The recombination rate of minority carriers and register are increased by the lattice damage. This study will investigate the current amplification characteristics of a pnp Si BJT through fast neutron irradiation experiments. In this paper, the current amplification characteristics of a pnp Si BJT were investigated for fast neutron irradiation. The experimental results show that base-tocollector current amplification ratio is decreased with an increase in the fast neutron irradiation. These indicate that the lattice damage caused by fast neutron irradiation increases the recombination rate of minority carriers and resistor.

  3. Influence of bath composition on the electrodeposition of cobalt-molybdenum amorphous alloy thin films

    Institute of Scientific and Technical Information of China (English)

    Qiaoying Zhou; Hongliang Ge; Guoying Wei; Qiong Wu

    2008-01-01

    Cobalt-molybdenum (Co-Mo) amorphous alloy thin films were deposited on copper substrates by the electrochemical method at pH 4.0. Among the experimental electrodeposition parameters, only the concentration ratio of molybdate to cobalt ions ([ MoO2-4 ]/[CO2+]) was varied to analyze its influence on the mechanism of induced cobalt-molybdenum codeposition. Voltammetry was one of the main techniques, which was used to examine the voltammetric response, revealing that cobalt-molybdenum codeposi-tion depended on the nature of the species in solution. To correlate the type of the film to the electrochemical response, various co-bait-molybdenum alloy thin films obtained from different [ MoO2-4]/[Co2+] solutions were tested. Crack-free homogeneous films could be easily obtained from the low molybdate concentrations ([ MoO2-4]/[Co2+]≈0.05) applying low deposition potentials.Moreover, the content of molybdenum up to 30wt% could be obtained from high molybdate concentration; in this case, the films showed cracks. The formation of these cracked films could be predicted from the observed distortions in the curves of electric cur-rent-time (j-t) deposition transients. The films with amorphous stmeture were obtained. The hysteresis loops suggested that the easily film were obtained when the deposition potential was -1025 mV, and [ MoO2-4]/[Co2+] was 0.05 in solution, which exhibited a nicer soft-magnetic response.

  4. Effect of molybdenum on structure, microstructure and mechanical properties of biomedical Ti-20Zr-Mo alloys.

    Science.gov (United States)

    Kuroda, Pedro Akira Bazaglia; Buzalaf, Marília Afonso Rabelo; Grandini, Carlos Roberto

    2016-10-01

    Titanium has an allotropic transformation around 883°C. Below this temperature, the crystalline structure is hexagonal close-packed (α phase), changing to body-centered cubic (β phase). Zirconium has the same allotropic transformation around 862°C. Molybdenum has body-centered cubic structure, being a strong β-stabilizer for the formation of titanium alloys. In this paper, the effect of substitutional molybdenum was analyzed on the structure, microstructure and selected mechanical properties of Ti-20Zr-Mo (wt%) alloys to be used in biomedical applications. The samples were prepared by arc-melting and characterized by x-ray diffraction with subsequent refinement by the Rietveld method, optical and scanning electron microscopy. The mechanical properties were analyzed by Vickers microhardness and dynamic elasticity modulus. X-ray measurements and Rietveld analysis revealed the presence of α' phase without molybdenum, α'+α″ phases with 2.5wt% of molybdenum, α″+β phases with 5 and 7.5wt% of molybdenum, and only β phase with 10wt% of molybdenum. These results were corroborated by microscopy results, with a microstructure composed of grains of β phase and lamellae and needles of α' and α″ phase in intra-grain the region. The hardness of the alloy was higher than the commercially pure titanium, due to the action of zirconium and molybdenum as hardening agents. The samples have a smaller elasticity modulus than the commercially pure titanium.

  5. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  6. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  7. Neutron irradiation induced amorphization of silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Snead, L.L.; Hay, J.C. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    This paper provides the first known observation of silicon carbide fully amorphized under neutron irradiation. Both high purity single crystal hcp and high purity, highly faulted (cubic) chemically vapor deposited (CVD) SiC were irradiated at approximately 60 C to a total fast neutron fluence of 2.6 {times} 10{sup 25} n/m{sup 2}. Amorphization was seen in both materials, as evidenced by TEM, electron diffraction, and x-ray diffraction techniques. Physical properties for the amorphized single crystal material are reported including large changes in density ({minus}10.8%), elastic modulus as measured using a nanoindentation technique ({minus}45%), hardness as measured by nanoindentation ({minus}45%), and standard Vickers hardness ({minus}24%). Similar property changes are observed for the critical temperature for amorphization at this neutron dose and flux, above which amorphization is not possible, is estimated to be greater than 130 C.

  8. Diamond Deposition on WC/Co Alloy with a Molybdenum Intermediate Layer

    Science.gov (United States)

    Liu, Sha; Yu, Zhi-Ming; Yi, Dan-Qing

    It is known that in the condition of chemical vapor deposition (CVD) diamond process, molybdenum is capable of forming carbide known as the "glue" which promotes growth of the CVD diamond, and aids its adhesion by (partial) relief of stresses at the interface. Furthermore, the WC grains are reaction bonded to the Mo2C phase. Therefore, molybdenum is a good candidate material for the intermediate layer between WC-Co substrates and diamond coatings. A molybdenum intermediate layer of 1-3 μm thickness was magnetron sputter-deposited on WC/Co alloy prior to the deposition of diamond coatings. Diamond films were deposited by hot filament chemical vapor deposition (HFCVD). The chemical quality, morphology, and crystal structure of the molybdenum intermediate layer and the diamond coatings were characterized by means of SEM, EDX, XRD and Raman spectroscopy. It was found that the continuous Mo intermediate layer emerged in spherical shapes and had grain sizes of 0.5-1.5 μm after 30 min sputter deposition. The diamond grain growth rate was slightly slower as compared with that of uncoated Mo layer on the WC-Co substrate. The morphologies of the diamond films on the WC-Co substrate varied with the amount of Mo and Co on the substrate. The Mo intermediate layer was effective to act as a buffer layer for both Co diffusion and diamond growth.

  9. A chemical approach toward low temperature alloying of immiscible iron and molybdenum metals

    Energy Technology Data Exchange (ETDEWEB)

    Nazir, Rabia [Department of Chemistry, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Applied Chemistry Research Centre, Pakistan Council of Scientific and Industrial Research Laboratories Complex, Lahore 54600 (Pakistan); Ahmed, Sohail [Department of Chemistry, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Mazhar, Muhammad, E-mail: mazhar42pk@yahoo.com [Department of Chemistry, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Akhtar, Muhammad Javed; Siddique, Muhammad [Physics Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan); Khan, Nawazish Ali [Material Science Laboratory, Department of Physics, Quaid-i-Azam University, Islamabad 45320 (Pakistan); Shah, Muhammad Raza [HEJ Research Institute of Chemistry, University of Karachi, Karachi 75270 (Pakistan); Nadeem, Muhammad [Physics Division, PINSTECH, P.O. Nilore, Islamabad (Pakistan)

    2013-11-15

    Graphical abstract: - Highlights: • Low temperature pyrolysis of [Fe(bipy){sub 3}]Cl{sub 2} and [Mo(bipy)Cl{sub 4}] homogeneous powder. • Easy low temperature alloying of immiscible metals like Fe and Mo. • Uniform sized Fe–Mo nanoalloy with particle size of 48–68 nm. • Characterization by EDXRF, AFM, XRPD, magnetometery, {sup 57}Fe Mössbauer and impedance. • Alloy behaves as almost superparamagnetic obeying simple –R(CPE)– circuit. - Abstract: The present research is based on a low temperature operated feasible method for the synthesis of immiscible iron and molybdenum metals’ nanoalloy for technological applications. The nanoalloy has been synthesized by pyrolysis of homogeneous powder precipitated, from a common solvent, of the two complexes, trisbipyridineiron(II)chloride, [Fe(bipy){sub 3}]Cl{sub 2}, and bipyridinemolybedenum(IV) chloride, [Mo(bipy)Cl{sub 4}], followed by heating at 500 °C in an inert atmosphere of flowing argon gas. The resulting nanoalloy has been characterized by using EDXRF, AFM, XRD, magnetometery, {sup 57}Fe Mössbauer and impedance spectroscopies. These results showed that under provided experimental conditions iron and molybdenum metals, with known miscibility barrier, alloy together to give (1:1) single phase material having particle size in the range of 48–66 nm. The magnetism of iron is considerably reduced after alloy formation and shows its trend toward superparamagnetism. The designed chemical synthetic procedure is equally feasible for the fabrication of other immiscible metals.

  10. Hot spray technology of TA7 titanium alloy coated by molybdenum and its bonding strength

    Institute of Scientific and Technical Information of China (English)

    Li Xiaoquan; Du Zeyu; Yang Xuguang

    2006-01-01

    A kind of surface modification test was introduced, by which plasma spray in argon atmosphere with CNC4500 system was applied for TA7 titanium alloy to be coated with molybdenum, and technology to produce metallurgical bonding at interface of coating and base meal was tested by heating in vacuum condition for diffusion after hot spray.With the help of scan electron microscope analysis ( SEM) , the effect of argon inlet pressure and heating temperature on coating structure as well as product of diffusion layer were studied.The glued tensile test method was used to measure bonding strength of base metal to coating.The result has shown that both argon inlet pressure and heating temperature exert some effect on coating structure and the width of diffusion layer.A bonding strength of base metal to coating which is greater than molybdenum coating itself may be attained and can be controlled in more than 50 MPa level with tested hot spray technology.

  11. Effect of molybdenum on structure, microstructure and mechanical properties of biomedical Ti-20Zr-Mo alloys

    Energy Technology Data Exchange (ETDEWEB)

    Kuroda, Pedro Akira Bazaglia [UNESP - Univ Estadual Paulista, Laboratório de Anelasticidade e Biomateriais, 17.033-360, Bauru, SP (Brazil); IBTN/Br – Institute of Biomaterials, Tribocorrosion and Nanomedicine, Brazilian Branch, 17.033-360 Bauru, SP (Brazil); Buzalaf, Marília Afonso Rabelo [USP – Universidade de São Paulo, Departamento de Ciências Biológicas, 17.012-901, Bauru, SP (Brazil); Grandini, Carlos Roberto, E-mail: betog@fc.unesp.br [UNESP - Univ Estadual Paulista, Laboratório de Anelasticidade e Biomateriais, 17.033-360, Bauru, SP (Brazil); IBTN/Br – Institute of Biomaterials, Tribocorrosion and Nanomedicine, Brazilian Branch, 17.033-360 Bauru, SP (Brazil)

    2016-10-01

    Titanium has an allotropic transformation around 883 °C. Below this temperature, the crystalline structure is hexagonal close-packed (α phase), changing to body-centered cubic (β phase). Zirconium has the same allotropic transformation around 862 °C. Molybdenum has body-centered cubic structure, being a strong β-stabilizer for the formation of titanium alloys. In this paper, the effect of substitutional molybdenum was analyzed on the structure, microstructure and selected mechanical properties of Ti-20 Zr-Mo (wt%) alloys to be used in biomedical applications. The samples were prepared by arc-melting and characterized by x-ray diffraction with subsequent refinement by the Rietveld method, optical and scanning electron microscopy. The mechanical properties were analyzed by Vickers microhardness and dynamic elasticity modulus. X-ray measurements and Rietveld analysis revealed the presence of α′ phase without molybdenum, α′ + α″ phases with 2.5 wt% of molybdenum, α″ + β phases with 5 and 7.5 wt% of molybdenum, and only β phase with 10 wt% of molybdenum. These results were corroborated by microscopy results, with a microstructure composed of grains of β phase and lamellae and needles of α′ and α″ phase in intra-grain the region. The hardness of the alloy was higher than the commercially pure titanium, due to the action of zirconium and molybdenum as hardening agents. The samples have a smaller elasticity modulus than the commercially pure titanium. - Highlights: • Ti-20Zr-Mo system alloys were developed. • β-Stabilizer effect of Zr in the presence of another β-stabilizer element • Alloys with low elastic modulus.

  12. The intrinsic gettering in neutron irradiation Czochralski-silicon

    CERN Document Server

    Li Yang Xian; Niu Ping Juan; Liu Cai Chi; Xu Yue Sheng; Yang Deren; Que Duan Lin

    2002-01-01

    The intrinsic gettering in neutron irradiated Czochralski-silicon is studied. The result shows that a denuded zone at the surface of the neutron irradiated Czochralski-silicon wafer may be formed through one-step short-time annealing. The width of the denuded zone is dependent on the annealing temperature and the dose of neutron irradiation, while it is irrelated to the annealing time in case the denuded zone is formed. The authors conclude that the interaction between the defects induced by neutron irradiation and the oxygen in the silicon accelerates the oxygen precipitation in the bulk, and becomes the dominating factor of the quick formation of intrinsic gettering. It makes the effect of thermal history as the secondary factor

  13. Neutron irradiation effects on superconducting wires and insulating materials

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Arata [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)], E-mail: nishi-a@nifs.ac.jp; Takeuchi, Takao [National Institute for Materials Science, 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Nishijima, Shigehiro [Graduate School of Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Nishijima, Gen; Shikama, Tatsuo [Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai, Naka, Ibaraki 319-1195 (Japan); Koizumi, Norikiyo [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

    2009-06-15

    On the progress of the Deuterium-Deuterium (D-D) or Deuterium-Tritium (D-T) burning plasma devices, the importance of neutron irradiation on superconducting magnet materials increases and the data base is desired to design the next generation devices. To carry out the investigations on the effect of neutron irradiation, neutron irradiation fields are required together with post-irradiation test facilities. In these several years, a collaboration network of neutron irradiation effect on superconducting magnet materials has been constructed. 14 MeV neutron irradiation was carried out at Fusion Neutronics Sources (FNS) in Japan Atomic Energy Agency (JAEA) and fission neutron irradiation was performed at JRR-3 in JAEA. After the irradiation, the Nb{sub 3}Sn, NbTi and Nb{sub 3}Al samples were sent to High Field Laboratory for Superconducting Materials (HFLSM) in Tohoku University and the superconducting properties were evaluated with 28 T hybrid magnet. Also, the organic insulation materials are considered to be weaker than superconducting materials against neutron irradiation and cyanate ester resin composite was fabricated and tested at the fission reactor. One clear result on Nb{sub 3}Sn was the property change of Nb{sub 3}Sn by 14 MeV neutron irradiation over 13 T. The critical current was increased by 1.4 times around 13 T but the increment of the critical current became almost zero at higher magnetic fields and the critical magnetic field of the irradiated sample showed almost the same as non-irradiated one.

  14. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  15. Polyethylene terephthalate degradation under reactor neutron irradiation

    Science.gov (United States)

    Chikaoui, K.; Izerrouken, M.; Djebara, M.; Abdesselam, M.

    2017-01-01

    This paper is devoted to study the defects generated by reactor neutron in polyethylene terephthalate (PET) films. The explored fast neutron fluence ranges from 2.02×1016 to 2.07×1018 n cm-2. The induced damages were investigated using ultraviolet-visible spectrophotometry (UV-vis), Fourier Transform Infrared spectrometry (FTIR) and X-ray diffraction (XRD). The UV-vis spectra show important changes indicating the degradation of the chemical structure and the creation of new chromophores. FTIR spectra reveal that the intensities of the different absorption bands decrease linearly under fast neutron irradiation. The internal reference band at 1410 cm-1 is used to follow the overall damage during irradiation. The 1342 cm-1 band corresponding to CH2 wagging of trans conformation of crystalline phase show a sharpe linear decrease as the fast neutrons fluence goes up. The creation of the monosubstituted benzene, investigated using the 1610 cm-1 band. It shows a linear increase with fast neutron fluence. It is found from XRD analysis that the diffraction peak (100) intensity is drastically reduced after irradiation at 2.02×1016 n cm-2.

  16. Tensile properties of Inconel 718 after low temperature neutron irradiation

    Science.gov (United States)

    Byun, T. S.; Farrell, K.

    2003-05-01

    Tensile properties of Inconel 718 (IN718) have been investigated after neutron irradiation to 0.0006-1.2 dpa at 60-100 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The alloy was exposed in solution-annealed (SA) and precipitation-hardened (PH) conditions. Before irradiation, the yield strength of PH IN718 was about 1170 MPa, which was 3.7 times higher than that of SA IN718. In the SA condition, an almost threefold increase in yield strength was found at 1.2 dpa, but the alloy retained a positive strain-hardening capability and a uniform ductility of more than 20%. Comparisons showed that the strain-hardening behavior of the SA IN718 is similar to that of a SA 316LN austenitic stainless steel. In the PH condition, the IN718 displayed no radiation-induced hardening in yield strength and significant softening in ultimate tensile strength. The strain-hardening capability of the PH IN718 decreased with dose as the radiation-induced dissolution of precipitates occurred, which resulted in the onset of plastic instability at strains less than 1% after irradiation to 0.16 or 1.2 dpa. An analysis on plastic instability indicated that the loss of uniform ductility in PH IN718 was largely due to the reduction in strain-hardening rate, while in SA IN718 and SA 316LN stainless steel it resulted primarily from the increase of yield stress.

  17. Cost Estimate for Molybdenum and Tantalum Refractory Metal Alloy Flow Circuit Concepts

    Science.gov (United States)

    Hickman, Robert R.; Martin, James J.; Schmidt, George R.; Godfroy, Thomas J.; Bryhan, A.J.

    2010-01-01

    The Early Flight Fission-Test Facilities (EFF-TF) team at NASA Marshall Space Flight Center (MSFC) has been tasked by the Naval Reactors Prime Contract Team (NRPCT) to provide a cost and delivery rough order of magnitude estimate for a refractory metal-based lithium (Li) flow circuit. The design is based on the stainless steel Li flow circuit that is currently being assembled for an NRPCT task underway at the EFF-TF. While geometrically the flow circuit is not representative of a final flight prototype, knowledge has been gained to quantify (time and cost) the materials, manufacturing, fabrication, assembly, and operations to produce a testable configuration. This Technical Memorandum (TM) also identifies the following key issues that need to be addressed by the fabrication process: Alloy selection and forming, cost and availability, welding, bending, machining, assembly, and instrumentation. Several candidate materials were identified by NRPCT including molybdenum (Mo) alloy (Mo-47.5 %Re), tantalum (Ta) alloys (T-111, ASTAR-811C), and niobium (Nb) alloy (Nb-1 %Zr). This TM is focused only on the Mo and Ta alloys, since they are of higher concern to the ongoing effort. The initial estimate to complete a Mo-47%Re system ready for testing is =$9,000k over a period of 30 mo. The initial estimate to complete a T-111 or ASTAR-811C system ready for testing is =$12,000k over a period of 36 mo.

  18. Comparison of spring characteristics of titanium-molybdenum alloy and stainless steel.

    Science.gov (United States)

    Sheibaninia, Ahmad; Salehi, Anahita; Asatourian, Armen

    2017-01-01

    Titanium-molybdenum alloy (TMA) and stainless steel (SS) wires are commonly used in orthodontics as arch-wires for tooth movement. However, plastic deformation phenomenon in these arch-wires seems to be a major concern among orthodontists. This study aimed to compare the mechanical properties of TMA and SS wires with different dimensions. Seventy-two wire samples (36 TMA and 36 SS) of three different sizes (19×25, 17×25 and 16×22) were analyzed in vitro, with 12 samples in each group. Various mechanical properties of the wires, including spring-back, bending moment and stiffness were determined using a universal testing machine. Student's t-test showed statistically significant differences in the mean values of all the groups. In addition, metallographic comparison of SS and TMA wires was conducted under an optical microscope. The degree of stiffness of 16×22-sized SS and TMA springs was found to be 12±2 and 5±0.4, respectively, while the bending moment was estimated to be 1927±352 (gm-mm) and 932±16 (gm-mm), respectively; the spring-back index was determined to be 0.61±0.2 and 0.4±.09, respectively (pBending moment, optical microscope, spring-back, stainless steel, stiffness, titanium‒molybdenum alloy.

  19. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  20. Characterization of the uranium--2 weight percent molybdenum alloy. [Treatment to obtain 930 MPa yield strength (0. 2 percent)

    Energy Technology Data Exchange (ETDEWEB)

    Hemperly, V.C.

    1976-05-19

    The uranium-2 wt percent molybdenum alloy was prepared, processed, and age hardened to meet a minimum 930-MPa yield strength (0.2 percent) with a minimum of 10 percent elongation. These mechanical properties were obtained with a carbon level up to 300 ppM in the alloy. The tensile-test ductility is lowered by the humidity of the laboratory atmosphere. (auth)

  1. Tensile and stress-rupture behavior of hafnium carbide dispersed molybdenum and tungsten base alloy wires

    Science.gov (United States)

    Yun, Hee Mann; Titran, Robert H.

    1993-01-01

    The tensile strain rate sensitivity and the stress-rupture strength of Mo-base and W-base alloy wires, 380 microns in diameter, were determined over the temperature range from 1200 K to 1600 K. Three molybdenum alloy wires; Mo + 1.1w/o hafnium carbide (MoHfC), Mo + 25w/o W + 1.1w/o hafnium carbide (MoHfC+25W) and Mo + 45w/o W + 1.1w/o hafnium carbide (MoHfC+45W), and a W + 0.4w/o hafnium carbide (WHfC) tungsten alloy wire were evaluated. The tensile strength of all wires studied was found to have a positive strain rate sensitivity. The strain rate dependency increased with increasing temperature and is associated with grain broadening of the initial fibrous structures. The hafnium carbide dispersed W-base and Mo-base alloys have superior tensile and stress-rupture properties than those without HfC. On a density compensated basis the MoHfC wires exhibit superior tensile and stress-rupture strengths to the WHfC wires up to approximately 1400 K. Addition of tungsten in the Mo-alloy wires was found to increase the long-term stress rupture strength at temperatures above 1400 K. Theoretical calculations indicate that the strength and ductility advantage of the HfC dispersed alloy wires is due to the resistance to recrystallization imparted by the dispersoid.

  2. Alloy hardening and softening in binary molybdenum alloys as related to electron concentration.

    Science.gov (United States)

    Stephens, J. R.; Witzke, W. R.

    1972-01-01

    Determination of the effects of alloy additions of Hf, Ta, W, Re, Os, Ir, and Pt on the hardness of Mo. Special emphasis was placed on alloy softening in these binary Mo alloys. A modified microhardness test unit permitted hardness determinations at homologous temperatures ranging from 0.02 to 0.15, where alloy softening normally occurs in bcc alloys. Results showed that alloy softening was produced by those elements having an excess of s + d electrons compared to Mo while those elements having an equal number or fewer s + d electrons than Mo failed to produce alloy softening. The magnitude of the softening and the amount of solute element at the hardness minimum diminished rapidly with increasing test temperature. At solute concentrations where alloy softening was observed, the temperature sensitivity of hardness was lowered. For solute elements having an excess of s + d electrons or fewer s + d electrons than Mo, alloy softening and alloy hardening can be correlated with the difference in number of s + d electrons of the solute element and Mo.

  3. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.

  4. New facility for post irradiation examination of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo; Kawamura, Hiroshi [Oarai Research Establishment, Ibaraki-Ken (Japan)

    1995-09-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800{degrees}C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and {sup 60}Co;7.4 MBq/day.

  5. The influence of Neutron Irradiation in CR-39 polymer

    Directory of Open Access Journals (Sweden)

    Sangeeta Prasher

    2015-06-01

    Full Text Available The script allocates the influence of neutron irradiations on the structural and optical properties of CR-39. The structural properties of the polymer have been examined using the FTIR spectrum of the pristine and neutron beam irradiated polymer. The studies reveal the increase the intensity of some bands with neutron irradiation pointing the increase in the unsaturated behavior of the polymer. The optical properties analyzed using the UV-Vis spectra made it evident that CR-39 gets easily influenced at a fluence of 1016 n/cm2. The glassy characteristics of the polymer also found to diminish with increasing neutron irradiation. Significant variations in the property profile of the polymer have been observed at higher neutron fluence.

  6. Wear behaviour of cobalt-chromium-molybdenum alloys used in metal-on-metal hip implants

    Science.gov (United States)

    Varano, Rocco

    The influence of carbon (C) content, microstructure, crystallography and mechanical properties on the wear behaviour of metal-on-metal (MM) hip implants made from commercially available cobalt-chromium-molybdenum (CoCrMo) alloys designated as American Society of Testing and Materials (ASTM) grade F1537, F75 and as-cast were studied in this work. The as-received bars of wrought CoCrMo alloys (ASTM F1537 of either about 0.05% or 0.26% C) were each subjected to various heat treatments to develop different microstructures. Pin and plate specimens were fabricated from each bar and were tested against each other using a linear reciprocating pin-on-plate apparatus in 25% by volume bovine serum solution. The applied normal load was 9.81 N and the reciprocating plate had a sinusoidal velocity with an average speed of 26 mm/s. The wear was measured gravimetrically and it was found to be most strongly affected by alloy C content, irrespective of grain size or carbide morphology. More precisely, the wear behaviour was directly correlated to the dissolved C content of the alloys. Increased C in solid-solution coincided with lower volumetric wear since C helps to stabilize the face-centred cubic (FCC) crystal structure thus limiting the amount of strain induced transformation (SIT) to the hexagonal close-packed crystal structure (HCP). Based on the observed surface twinning in and around the contact zone and the potentially detrimental effect of the HCP phase, it was postulated that the MM wear behaviour of CoCrMo alloys in the present study was controlled by a deformation mechanism, rather than corrosion or tribochemical reactions.

  7. Comparison of spring characteristics of titanium-molybdenum alloy and stainless steel

    Science.gov (United States)

    Salehi, Anahita; Asatourian, Armen

    2017-01-01

    Background Titanium-molybdenum alloy (TMA) and stainless steel (SS) wires are commonly used in orthodontics as arch-wires for tooth movement. However, plastic deformation phenomenon in these arch-wires seems to be a major concern among orthodontists. This study aimed to compare the mechanical properties of TMA and SS wires with different dimensions. Material and Methods Seventy-two wire samples (36 TMA and 36 SS) of three different sizes (19×25, 17×25 and 16×22) were analyzed in vitro, with 12 samples in each group. Various mechanical properties of the wires, including spring-back, bending moment and stiffness were determined using a universal testing machine. Student’s t-test showed statistically significant differences in the mean values of all the groups. In addition, metallographic comparison of SS and TMA wires was conducted under an optical microscope. Results The degree of stiffness of 16×22-sized SS and TMA springs was found to be 12±2 and 5±0.4, respectively, while the bending moment was estimated to be 1927±352 (gm-mm) and 932±16 (gm-mm), respectively; the spring-back index was determined to be 0.61±0.2 and 0.4±.09, respectively (p<0.001). There were no statistically significant differences in spring-back index in larger dimensions of the wires. Conclusions Systematic analysis indicated that springs made of TMA were superior compared to those made of SS. Although both from economic and functionality viewpoints the use of TMA is suggested, further clinical investigations are recommended. Key words:Bending moment, optical microscope, spring-back, stainless steel, stiffness, titanium‒molybdenum alloy. PMID:28149469

  8. New decorative applications of alloys at base of aluminium-molybdenum; Nuevas aplicaciones decorativas de aleaciones a base de aluminio-molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Mesa L, V.H.; Hernandez P, C.D.; Alvarez P, M.A.; Guzman, J.; Garcia H, M.; Juarez I, J.A.; Gonzalez, C.; Alvarez F, O. [IIM-UNAM, A.P. 70-360, 04510 Mexico D.F. (Mexico)]. e-mail: oaf@servidor.unam.mx

    2005-07-01

    Decorative properties of aluminum-molybdenum alloys have been analyzed as a function of chemical composition and applied heat treatment. These decorative application are due exclusively for their nano structure nature. The alloys were prepared at room temperature by DC magnetron sputtering technique in argon atmosphere at different deposition time to obtain several thickness and chemical compositions in the range 3 to 30 % of molybdenum metal. (Author)

  9. Fluence dependence of defect evolution in austenitic stainless steels during fission neutron irradiation

    Science.gov (United States)

    Watanabe, H.; Muroga, T.; Yoshida, N.

    To understand microstructural evolution during fission neutron irradiation, a pure Fe-Cr-Ni ternary alloy, phosphorus-containing model austenitic stainless steels and SUS316 were irradiated in a Japanese Material Testing Reactor (JMTR) at 493 and 613 K. At 493 K, the density of defect cluster increased with the irradiation dose, but there was no significant change in loop density and loop size among all the materials. At 613 K, on the other hand, interstitial type dislocation loops and phosphides were formed in pure ternary and phosphorus-containing alloys, respectively, by an early stage of irradiation. These results suggest that the defect cluster formation at 493 and 613 K is mainly controlled by the cascade damage and long-range migration of free point defects, respectively.

  10. Dislocation morphology in deformed and irradiated niobium. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C. P.

    1977-06-01

    Niobium foils of moderate purity were examined for the morphology of dislocations or defect clusters in the deformed or neutron-irradiated state by transmission electron microscopy. New evidence has been found for the dissociation of screw dislocations into partials on the (211) slip plane according to the Crussard mechanism: (a/2) (111) ..-->.. (a/3) (111) + (a/6) (111).

  11. Comparison of properties and microstructures of Trefimetaux and Hycon 3HP{trademark} after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M. [Risoe National Lab., Roskilde (Denmark)

    1998-09-01

    The precipitation strengthened CuNiBe alloys are among three candidate copper alloys being evaluated for application in the first wall, divertor, and limiter components of ITER. Generally, CuNiBe alloys have higher strength but poorer conductivity compared to CuCrZr and CuAl{sub 2}O{sub 3} alloys. Brush-Wellman Inc. has manufactured an improved version of their Hycon CuNiBe alloy that has higher conductivity while maintaining a reasonable level strength. It is of interest, therefore, to investigate the effect of radiation on the physical and mechanical properties of this alloy. In the present work the authors have investigated the physical and mechanical properties of the Hycon 3HP{trademark} alloy both before and after neutron irradiation and have compared its microstructure and properties with the European CuNiBe candidate alloy manufactured by Trefirmetaux. Tensile specimens of both alloys were irradiated in the DR-3 reactor at Risoe to displacement dose levels up to 0.3 dpa at 100, 250 and 350 C. Both alloys were tensile tested in the unirradiated and irradiated conditions at 100, 250 and 350 C. Both pre- and post-irradiation microstructures of the alloys were investigated in detail using transmission electron microscopy. Fracture surfaces were examined under a scanning electron microscope. Electrical resistivity measurements were made on tensile specimens before and after irradiation; all measurements were made at 23 C. At this point it seems unlikely that CuNiBe alloys can be recommended for applications in neutron environments where the irradiation temperature exceeds 200 C. Applications at temperatures below 200 C might be plausible, but only after careful experiments have determined the dose dependence of the mechanical properties and the effect of sudden temperature excursions on the material to establish the limits on the use of the alloy.

  12. Effect of molybdenum addition on the mechanical properties of sinter-forged Fe–Cu–C alloys

    Energy Technology Data Exchange (ETDEWEB)

    Rathore, Sanjay S., E-mail: rathore.sanjaysingh@gmail.com; Salve, Milind M., E-mail: milindrowling@gmail.com; Dabhade, Vikram V., E-mail: vvdabfmt@iitr.ac.in

    2015-11-15

    Molybdenum provides solid solution strengthening, enhances hardenability and has thus been used to improve mechanical properties of ferrous alloys significantly. The present study reports the effect of molybdenum addition on the properties of sinter-forged Fe–Cu–C alloys prepared using elemental powders under various heat treatment conditions. The elemental powder mixtures were compacted at a pressure of 500 MPa followed by sintering at 1120 °C in N{sub 2}–20%H{sub 2} atmosphere. Further, the sintered compacts were immediately forged at the sintering temperature in a closed die. The sinter-forged compacts were further homogenized and then heat treated under different cooling rates. Enhancement of the mechanical properties (hardness and tensile strength) were observed with Mo addition and increase in severity of quench. Hardness of air cooled samples was slightly lower than that of the water quenched samples but comparable with oil quenched samples. However, no significant increase in hardness was observed beyond 1.5 wt% Mo addition for all cooling conditions. At higher molybdenum content ductility was retained due to stabilization of ferrite phase by molybdenum. The microstructural study showed mostly ferrite–pearlite structure in furnace cooled condition whereas a complex microstructure was observed in the faster cooling conditions. Grain refinement was also observed with molybdenum addition. - Highlights: • Mo (0.25–4.0 wt%) addition in sinter-forged Fe–2Cu–0.65C alloys was investigated. • Effect of heat treatment on mechanical properties and microstructure was discussed. • Hardness and strength increased with Mo addition at the expense of ductility. • Hardness in air cooled condition was comparable with oil/water cooled conditions.

  13. Dilatometric analysis on shrinkage behavior during non-isothermal sintering of nanocrystalline tungsten mechanically alloyed with molybdenum

    Energy Technology Data Exchange (ETDEWEB)

    Srivastav, Ajeet K., E-mail: ajeetshri@gmail.com [Department of Metallurgical and Materials Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Murty, B.S. [Department of Metallurgical and Materials Engineering, Indian Institute of Technology Madras, Chennai 600036 (India)

    2012-09-25

    Highlights: Black-Right-Pointing-Pointer Alloying with Mo reduces the WC contamination during milling. Black-Right-Pointing-Pointer Kirkendall effect assisted enhanced diffusion influences the densification of W-Mo alloys Black-Right-Pointing-Pointer Densification kinetics improved with Mo alloying in nanocrystalline tungsten. Black-Right-Pointing-Pointer Densification starts with Mo diffusion and later W and Mo both diffuse along the grain boundaries. - Abstract: The paper attempts to study the shrinkage behavior of nanocrystalline tungsten mechanically alloyed with molybdenum (5, 10, 15 and 20 wt.%). The dilatometric analysis was performed by Setsys Evolution TMA (ambient to 1600 Degree-Sign C) using constant heating rate (CHR) method. The significant improvement in shrinkage with alloying of molybdenum is attributed to reduced grain size, lowered tungsten carbide contamination and enhanced diffusion kinetics. The initial stage sintering kinetics of W-20Mo alloy has been investigated. The densification starts with Mo diffusion (calculated activation energy = 128 kJ/mol) and proceeds with the diffusion of both along the grain boundaries (calculated activation energy = 307 {+-} 1 kJ/mol).

  14. Two-dimensional molybdenum tungsten diselenide alloys: photoluminescence, Raman scattering, and electrical transport.

    Science.gov (United States)

    Zhang, Mei; Wu, Juanxia; Zhu, Yiming; Dumcenco, Dumitru O; Hong, Jinhua; Mao, Nannan; Deng, Shibin; Chen, Yanfeng; Yang, Yanlian; Jin, Chuanhong; Chaki, Sunil H; Huang, Ying-Sheng; Zhang, Jin; Xie, Liming

    2014-07-22

    Two-dimensional transition-metal dichalcogenide alloys have attracted intense attention due to their tunable band gaps. In the present work, photoluminescence, Raman scattering, and electrical transport properties of monolayer and few-layer molybdenum tungsten diselenide alloys (Mo1-xWxSe2, 0 ≤ x ≤ 1) are systematically investigated. The strong photoluminescence emissions from Mo1-xWxSe2 monolayers indicate composition-tunable direct band gaps (from 1.56 to 1.65 eV), while weak and broad emissions from the bilayers indicate indirect band gaps. The first-order Raman modes are assigned by polarized Raman spectroscopy. Second-order Raman modes are assigned according to its frequencies. As composition changes in Mo1-xWxSe2 monolayers and few layers, the out-of-plane A1g mode showed one-mode behavior, while B2g(1) (only observed in few layers), in-plane E2g(1), and all observed second-order Raman modes showed two-mode behaviors. Electrical transport measurement revealed n-type semiconducting transport behavior with a high on/off ratio (>10(5)) for Mo1-xWxSe2 monolayers.

  15. Annealing behaviors of vacancy in varied neutron irradiated Czochralski silicon

    Institute of Scientific and Technical Information of China (English)

    CHEN Gui-feng; LI Yang-xian; LIU Li-li; NIU Ping-juan; NIU Sheng-li; CHEN Dong-feng

    2006-01-01

    The difference of annealing behaviors of vacancy-oxygen complex (VO) in varied dose neutron irradiated Czochralski silicon: (S1 5×1017 n/cm3 and S2 1.07×1019 n/cm3) were studied. The results show that the VO is one of the main defects formed in neutron irradiated Czochralski silicon (CZ-Si). In this defect,oxygen atom shares a vacancy,it is bonded to two silicon neighbors. Annealed at 200 ℃,divacancies are trapped by interstitial oxygen(Oi) to form V2O (840 cm-1). With the decrease of the 829 cm-1 (VO) three infrared absorption bands at 825 cm-1 (V2O2),834 cm-1 (V2O3) and 840 cm-1 (V2O) will rise after annealed at temperature range of 200-500 ℃. After annealed at 450-500 ℃ the main absorption bands in S1 sample are 834 cm-1,825 cm-1 and 889 cm-1 (VO2),in S2 is 825 cm-1. Annealing of A-center in varied neutron irradiated CZ-Si is suggested to consist of two processes. The first is due to trapping of VO by Oi in low dose neutron irradiated CZ-Si (S1) and the second is due to capture the wandering vacancy by VO,etc,in high dose neutron irradiated CZ-Si (S2),the VO2 plays an important role in the annealing of A-center. With the increase of the irradiation dose,the annealing behavior of A-center is changed.

  16. Effect of periodic temperature variations on the microstructure of neutron-irradiated metals

    DEFF Research Database (Denmark)

    Zinkle, S.J.; Hashimoto, N.; Hoelzer, D.T.

    2002-01-01

    Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom.......-induced microstructural features consisted of dislocation loops, stacking fault tetrahedra and voids in the stainless steel, Ti-rich precipitates in the V alloy, and voids (along with a low density of stacking fault tetrahedra) in copper.......Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom...... and V–4Cr–4Ti were qualitatively similar to companion specimens that were continuously maintained at 520 °C during the entire irradiation. The microstructural observations on pure copper irradiated at a constant temperature of 340 °C in this experiment are also summarized. The main radiation...

  17. 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.

  18. Accelerated oxygen precipitation in fast neutron irradiated Czochralski silicon

    Institute of Scientific and Technical Information of China (English)

    Ma Qiao-Yun; Li Yang-Xian; Chen Gui-Feng; Yang Shuai; Liu Li-Li; Niu Ping-Juan; Chen Dong-Feng; Li Hong-Tao

    2005-01-01

    Annealing effect of the oxygen precipitation and the induced defects have been investigated on the fast neutron irradiated Czochralski silicon (CZ-Si) by infrared absorption spectrum and the optical microscopy. It is found that the fast neutron irradiation greatly accelerates the oxygen precipitation that leads to a sharp decrease of the interstitial oxygen with the annealing time. At room temperature (RT), the 1107cm-1 infrared absorption band of interstitial oxygen becomes weak and broadens to low energy side. At low temperature, the infrared absorption peaks appear at 1078cm-1, 1096cm-1, and 1182cm-1, related to different shapes of the oxygen precipitates. The bulk microdefects,including stacking faults, dislocations and dislocation loops, were observed by the optical microscopy. New or large stacking faults grow up when the silicon self-interstitial atoms are created and aggregate with oxygen precipitation.

  19. In vitro antileishmanial properties of neutron-irradiated meglumine antimoniate

    Energy Technology Data Exchange (ETDEWEB)

    Borborema, Samanta Etel Treiger; Nascimento, Nanci do [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), SP (Brazil). Lab. de Biologia Molecular]. E-mail: samanta@usp.br; Osso Junior, Joao Alberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), SP (Brazil). Centro de Radiofarmacia]. E-mail: jaosso@ipen.br; Andrade Junior, Heitor Franco de [Instituto de Medicina Tropical de Sao Paulo (IMT-SP), SP (Brazil). Lab. de Protozoologia]. E-mail:hfandrad@usp.br

    2005-10-15

    Pentavalent antimony, as meglumine antimoniate (Glucantime) or sodium stibogluconate (Pentostam), is the main treatment for leishmaniasis, a complex of diseases caused by the protozoan Leishmania, and an endemic and neglected threat in Brazil. Despite over half a century of clinical use, their mechanism of action, toxicity and pharmacokinetic data remain unknown. The analytical methods for determination of antimony in biological systems remain complex and have low sensitivity. Radiotracer studies have a potential in pharmaceutical development. The aim of this study was to obtain a radiotracer for antimony, with suitable physical and biological properties. Meglumine antimoniate was neutron irradiated inside the IEA-R1 nuclear reactor, producing two radioisotopes {sup 122} Sb and {sup 124} Sb, with high radionuclidic purity and good specific activity. This compound showed the same antileishmanial activity as the native compound. The use of the radiotracers, easily created by neutron irradiation, could be an interesting tool to solve important questions in antimonial pharmacology. (author)

  20. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    OpenAIRE

    2016-01-01

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation f...

  1. Neutron irradiation effects on high Nicalon silicon carbide fibers

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, M.C.; Steiner, D.; Snead, L.L. [Oak Ridge National Laboratory, TN (United States)

    1996-10-01

    The effects of neutron irradiation on the mechanical properties and microstructure of SiC and SiC-based fibers is a current focal point for the development of radiation damage resistant SiC/SiC composites. This report discusses the radiation effects on the Nippon Carbon Hi-Nicalon{trademark} fiber system and also discusses an erratum on earlier results published by the authors on this material. The radiation matrix currently under study is also summarized.

  2. The proposed cold neutron irradiation facility at the Breazeale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dimeo, R. M.; Sokol, P. E.; Carpenter, J. M.

    1997-01-01

    We discuss the design considerations of a Cold Neutron Irradiation Facility (CNIF) originally to have been installed at the Penn State Breazeale Reactor (PSBR). The goal of this project was to study the effects of radiation-induced damage to cryogenic moderators and, in particular, solid methane. This work evolved through the design stage undergoing a full safety analysis and received tentative approval from the PSBR Safeguards Committee but was discontinued due to budgetary constraints. (auth)

  3. Study of neutron irradiated structures of ammonothermal GaN

    Science.gov (United States)

    Gaubas, E.; Ceponis, T.; Deveikis, L.; Meskauskaite, D.; Miasojedovas, S.; Mickevicius, J.; Pavlov, J.; Pukas, K.; Vaitkus, J.; Velicka, M.; Zajac, M.; Kucharski, R.

    2017-04-01

    Study of the radiation damage in GaN-based materials becomes an important aspect for possible application of the GaN detectors in the harsh radiation environment at the Large Hadron Collider and at other particle acceleration facilities. Intentionally doped and semi-insulating bulk ammonothermal GaN materials were studied to reveal the dominant defects introduced by reactor neutron irradiations. These radiation defects have been identified by combining electron spin resonance and transmission spectroscopy techniques. Characteristics of carrier lifetime dependence on neutron irradiation fluence were examined. Variations of the response of the capacitor-type sensors with neutron irradiation fluence have been correlated with the carrier lifetime changes. The measurements of the photoconductivity and photoluminescence transients have been used to study the variation of the parameters of radiative and non-radiative recombination. The examined characteristics indicate that AT GaN as a particle sensing material is radiation hard up to high hadron fluences  ⩾1016 cm‑2.

  4. Cobalt-chromium-molybdenum alloy causes metal accumulation and metallothionein up-regulation in rat liver and kidney

    DEFF Research Database (Denmark)

    Jakobsen, Stig Storgaard; Danscher, Gorm; Stoltenberg, Meredin;

    2007-01-01

    Cobalt-chromium-molybdenum (CoCrMo) metal-on-metal hip prosthesis has had a revival due to their excellent wear properties. However, particulate wear debris and metal ions liberated from the CoCrMo alloys might cause carcinogenicity, hypersensitivity, local and general tissue toxicity, genotoxicity...... and inflammation-generating qualities. Nine months after implanting small pieces of CoCrMo alloy intramuscularly and intraperitoneally in rats, we analysed the accumulation of metals with a multi-element analysis, and the levels of metallothionein I/II with real-time reverse transcriptase-polymerase chain reaction...... in liver and kidney. We found that metal ions are liberated from CoCrMo alloys and suggest that they are released by dissolucytosis, a process where macrophages causes the metallic surface to release metal ions. Animals with intramuscular implants accumulated metal in liver and kidney and metallohionein I...

  5. The effect of molybdenum on niobium, titanium carbonitride precipitate evolution and grain refinement in high-temperature vacuum carburizing alloys

    Science.gov (United States)

    Enloe, Charles M.

    Existing commercial carburizing alloys can be processed at higher temperatures and shorter processing times utilizing vacuum carburizing due to the suppression of intergranular oxidation. To provide resistance to undesired grain coarsening at these elevated temperatures and associated reduction in fatigue performance, microalloyed steel variants have been developed which employ fine Ti- and Nb-carbonitrides to suppress grain growth. Grain coarsening resistance is believed to be limited by the coarsening resistance of the precipitates themselves at high temperature, so further alloy/processing developments to enhance microalloy precipitate coarsening resistance based on a greater mechanistic understanding of solute interaction with microalloy precipitates would be beneficial. Molybdenum is known to affect microalloy precipitate evolution during processing in ferrite and austenite, but a unified explanation of the role of Mo in precipitate evolution is still lacking. Accordingly, the effect of molybdenum on microalloy precipitate size and composition evolutions and the associated onset of abnormal grain growth in austenite was investigated in Mo-bearing and Mo-free, Nb,Ti-microalloyed SAE 4120 steels. Molybdenum additions of 0.30 wt pct to alloys containing Nb additions of 0.05 and 0.10 wt pct Nb delayed the onset of abnormal grain growth in hot-rolled alloys reheated and soaked at 1050 °C and 1100 °C. The coarsening rate of microalloy precipitates was also reduced in Mo-bearing alloys relative to Mo-free alloys during isothermal soaking at 1050 °C, 1100 °C, and 1150 °C. The observed microalloy precipitate coarsening rates exceeded those predicted by the Lifshitz-Slyozov-Wagner relation for volume-diffusion-controlled coarsening, which is attributed to an initial bimodal precipitate size distribution prior to reheating to elevated temperature. Heat-treatments of hot-rolled alloys (tempering and solutionizing) prior to reheating to elevated temperature in

  6. Development of Direct Alloying by Molybdenum Oxides%氧化钼直接合金化炼钢的发展

    Institute of Scientific and Technical Information of China (English)

    李正邦; 朱航宇; 杨海森

    2013-01-01

    Theoretical basis of direct alloying by use of molybdenum oxides and the better period for adding molybdenum oxides to EAF (electric arc furnace) was introduced, and resent research results of stabilizing molybdenum trioxide to inhibit its sublimation and difference of direct alloying by molybdenum oxides was both compared at home and abroad. The suggested conclusion is that the use of calcium oxide, calcium carbonate, magnesium oxide or iron oxide mixed with molybdenum trioxide adding to EAF can inhibit volatilization of molybdenum trioxide significantly, molybdenum yield can be improved accordingly.%介绍了氧化钼直接合金化炼钢的理论依据和电炉炼钢过程中的最佳加入时期,总结了抑制氧化钼挥发的研究成果并对比分析了国内外氧化钼炼钢工艺流程的差别.由此得出,采用氧化钙、碳酸钙、氧化镁或氧化铁和三氧化钼混加的方式能显著抑制三氧化钼的挥发,提高钼的收得率.

  7. Molybdenum carbide supported nickel-molybdenum alloys for synthesis gas production via partial oxidation of surrogate biodiesel

    Science.gov (United States)

    Shah, Shreya; Marin-Flores, Oscar G.; Norton, M. Grant; Ha, Su

    2015-10-01

    In this study, NiMo alloys supported on Mo2C are synthesized by wet impregnation for partial oxidation of methyl oleate, a surrogate biodiesel, to produce syngas. When compared to single phase Mo2C, the H2 yield increases from 70% up to >95% at the carbon conversion of ∼100% for NiMo alloy nanoparticles that are dispersed over the Mo2C surface. Supported NiMo alloy samples are prepared at two different calcination temperatures in order to determine its effect on particle dispersion, crystalline phase and catalytic properties. The reforming test data indicate that catalyst prepared at lower calcination temperature shows better nanoparticle dispersion over the Mo2C surface, which leads to higher initial performance when compared to catalysts synthesized at higher calcination temperature. Activity tests using the supported NiMo alloy on Mo2C that are calcined at the lower temperature of 400 °C shows 100% carbon conversion with 90% H2 yield without deactivation due to coking over 24 h time-on-stream.

  8. Dry sliding friction and wear characteristics of Fe-C-Cu alloy containing molybdenum di sulphide

    Energy Technology Data Exchange (ETDEWEB)

    Dhanasekaran, S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600 036 (India); Gnanamoorthy, R. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600 036 (India)]. E-mail: gmoorthy@iitm.ac.in

    2007-07-01

    Sintered steels find increasing application as bearings and gears due to economical and technical reasons. Materials used for making these machine elements need to have high strength, good wear resistance and low coefficient of friction. An attempt is made to develop molybdenum di sulphide added iron-copper-carbon sintered steels using simple single stage compaction and sintering elemental powders. Friction and wear characteristics of the developed materials were evaluated using cylindrical specimens in a pin-on-disc sliding apparatus. Addition of molybdenum di sulphide increases the compressibility and increases the part density. Strength and hardness of the molybdenum di sulphide added compositions are better than the base composition. Addition of the 3% molybdenum di sulphide is found to be beneficial in improving friction and wear characteristics. Higher amount of brittle phases in the 5% molybdenum di sulphide added sample contributes to the reduction in the wear resistance.

  9. Microstructure and mechanical properties of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, E.; Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Terai, T.; Tanaka, S.

    1998-01-01

    Microstructure and mechanical properties of the neutron irradiated beryllium with total fast neutron fluences of 1.3 - 4.3 x 10{sup 21} n/cm{sup 2} (E>1 MeV) at 327 - 616degC were studied. Swelling increased by high irradiation temperature, high fluence, and by the small grain size and high impurity. Obvious decreasing of the fracture stress was observed in the bending test and in small grain specimens which had many helium bubbles on the grain boundary. Decreasing of the fracture stress for small grain specimens was presumably caused by crack propagation on the grain boundaries which weekend by helium bubbles. (author)

  10. Impurities effect on the swelling of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Donne, M.D.; Scaffidi-Argentina, F. [Institut fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany)

    1995-09-01

    An important factor controlling the swelling behaviour of fast neutron irradiated beryllium is the impurity content which can strongly affect both the surface tension and the creep strength of this material. Being the volume swelling of the old beryllium (early sixties) systematically higher than that of the more modem one (end of the seventies), a sensitivity analysis with the aid of the computer code ANFIBE (ANalysis of Fusion Irradiated BEryllium) to investigate the effect of these material properties on the swelling behaviour of neutron irradiated beryllium has been performed. Two sets of experimental data have been selected: the first one named Western refers to quite recently produced Western beryllium, whilst the second one, named Russian refers to relatively old (early sixties) Russian beryllium containing a higher impurity rate than the Western one. The results obtained with the ANFIBE Code were assessed by comparison with experimental data and the used material properties were compared with the data available in the literature. Good agreement between calculated and measured values has been found.

  11. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    In order to investigate the role of He in cavity nucleation in neutron irradiated iron and steel, pure iron and Eurofer-97 steel have been He implanted and neutron irradiated in a systematic way at different temperatures, to different He and neutron doses and with different He implantation rates...

  12. Optical absorption and luminescence in neutron-irradiated, silica-based fibers

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, D.W.; Farnum, E.H.; Clinard, F.W. [Los Alamos National Lab., CA (United States)] [and others

    1995-04-01

    The objectives of this work are to assess the effects of thermal annealing and photobleaching on the optical absorption of neutron-irradiated, silica fibers of the type proposed for use in ITER diagnostics, and to measure x-ray induced luminescence of unirradiated (virgin) and neutron-irradiated fibers.

  13. Friction stir surfacing of cast A356 aluminium–silicon alloy with boron carbide and molybdenum disulphide powders

    Directory of Open Access Journals (Sweden)

    R. Srinivasu

    2015-06-01

    Full Text Available Good castability and high strength properties of Al–Si alloys are useful in defence applications like torpedoes, manufacture of Missile bodies, and parts of automobile such as engine cylinders and pistons. Poor wear resistance of the alloys is major limitation for their use. Friction stir processing (FSP is a recognized surfacing technique as it overcomes the problems of fusion route surface modification methods. Keeping in view of the requirement of improving wear resistance of cast aluminium–silicon alloy, friction stir processing was attempted for surface modification with boron carbide (B4C and molybdenum disulfide (MoS2 powders. Metallography, micro compositional analysis, hardness and pin-on-disc wear testing were used for characterizing the surface composite coating. Microscopic study revealed breaking of coarse silicon needles and uniformly distributed carbides in the A356 alloy matrix after FSP. Improvement and uniformity in hardness was obtained in surface composite layer. Higher wear resistance was achieved in friction stir processed coating with carbide powders. Addition of solid lubricant MoS2 powder was found to improve wear resistance of the base metal significantly.

  14. Design, Construction, and Modeling of a 252Cf Neutron Irradiator

    Directory of Open Access Journals (Sweden)

    Blake C. Anderson

    2016-01-01

    Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

  15. Tensile and fracture toughness test results of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1998-01-01

    Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

  16. Proceedings of neutron irradiation technical meeting on BNCT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

  17. ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Cetiner, N. O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; McDuffee, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.

  18. Microstructural defects in EUROFER 97 after different neutron irradiation conditions

    Directory of Open Access Journals (Sweden)

    Christian Dethloff

    2016-12-01

    Full Text Available Characterization of irradiation induced microstructural evolution is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM is used to determine the defect microstructure after different neutron irradiation conditions. In particular dislocation loops, voids and precipitates are analyzed concerning defect nature, density and size distribution after irradiation to 15 dpa at 300 °C in the mixed spectrum High Flux Reactor (HFR. New results are combined with previously obtained data from irradiation in the fast spectrum BOR-60 reactor (15 and 32 dpa, 330 °C, which allows for assessment of dose and dose rate effects on the aforementioned irradiation induced defects and microstructural characteristics.

  19. Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

  20. Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045 and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) plate, sheet and strip

  1. Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nickel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) rod, bar, and wire

  2. Neutron irradiation effects on AlGaN/GaN high electron mobility transistors

    Institute of Scientific and Technical Information of China (English)

    Lü Ling; Zhang Jin-Cheng; Xue Jun-Shuai; Ma Xiao-Hua; Zhang Wei; Bi Zhi-Wei; Zhang Yue; Hao Yue

    2012-01-01

    AlGaN/GaN high electron mobility transistors (HEMTs) were exposed to 1 MeV neutron irradiation at a neutron fluence of 1 × 1015 cm-2.The dc characteristics of the devices,such as the drain saturation current and the maximum transconductance,decreased after neutron irradiation. The gate leakage currents increased obviously after neutron irradiation.However,the rf characteristics,such as the cut-off frequency and the maximum frequency,were hardly affected by neutron irradiation.The AlGaN/GaN heterojunctions have been employed for the better understanding of the degradation mechanism.It is shown in the Hall measurements and capacitance-voltage tests that the mobility and concentration of two-dimensional electron gas (2DEG) decreased after neutron irradiation.There was no evidence of the full-width at half-maximum of X-ray diffraction (XRD) rocking curve changing after irradiation,so the dislocation was not influenced by neutron irradiation.It is concluded that the point defects induced in AlGaN and GaN by neutron irradiation are the dominant mechanisms responsible for performance degradations of AlGaN/GaN HEMT devices.

  3. Evaluation of mechanical properties in stainless alloy ferritic with 5 % molybdenum; Avaliacao das propriedades mecanicas em ligas inoxidaveis ferriticas com 5% de molibdenio

    Energy Technology Data Exchange (ETDEWEB)

    Lima Filho, V.X.; Gomes, F.H.F.; Guimaraes, R.F.; Saboia, F.H.C.; Abreu, H.F.G. de [Instituto Federal de Educacao, Ciencia e Tecnologia do Ceara (IFCE). Campus Maracanau, CE (Brazil)], e-mail: venceslau@ifce.edu.br

    2010-07-01

    The deterioration of equipment in the oil industry is caused by high aggressiveness in processing the same. One solution to this problem would increase the content of molybdenum (Mo) alloys, since this improves the corrosion resistance. As the increase of Mo content causes changes in mechanical properties, we sought to evaluate the mechanical properties of alloys with 5% Mo and different levels of chromium (Cr). Were performed metallography and hardness measurement of the alloys in the annealed condition. Subsequent tests were performed tensile and Charpy-V, both at room temperature. The results showed that 2% difference in the content of Cr did not significantly alter the mechanical properties of alloys. The alloys studied had higher values in measured properties when compared to commercial ferritic alloys with similar percentages of Cr. The high content of Mo resulted in a brittle at room temperature but ductile at temperatures above 70 degree C. (author)

  4. Mineral resource of the month: molybdenum

    Science.gov (United States)

    Polyak, Désire E.

    2011-01-01

    The article offers information about the mineral molybdenum. Sources includes byproduct or coproduct copper-molybdenum deposits in the Western Cordillera of North and South America. Among the uses of molybdenum are stainless steel applications, as an alloy material for manufacturing vessels and as lubricants, pigments or chemicals. Also noted is the role played by molybdenum in renewable energy technology.

  5. Experimental investigation of the behaviour of tungsten and molybdenum alloys at high strain-rate and temperature

    Directory of Open Access Journals (Sweden)

    Scapin Martina

    2015-01-01

    Full Text Available The introduction in recent years of new, extremely energetic particle accelerators such as the Large Hadron Collider (LHC gives impulse to the development and testing of refractory metals and alloys based on molybdenum and tungsten to be used as structural materials. In this perspective, in this work the experimental results of a tests campaign on Inermet®  IT180 and pure Molybdenum (sintered by two different producers are presented. The investigation of the mechanical behaviour was performed in tension varying the strain-rates, the temperatures and both of them. Overall six orders of magnitude in strain-rate (between 10−3 and 103 s−1 were covered, starting from quasi-static up to high dynamic loading conditions. The high strain-rate tests were performed using a direct Hopkinson Bar setup. Both in quasi-static and high strain-rate conditions, the heating of the specimens was obtained with an induction coil system, controlled in feedback loop, based on measurements from thermocouples directly welded on the specimen. The temperature range varied between 25 and 1000°C. The experimental data were, finally, used to extract the parameters of the Zerilli-Armstrong model used to reproduce the mechanical behaviour of the investigated materials.

  6. Experimental investigation of the behaviour of tungsten and molybdenum alloys at high strain-rate and temperature

    CERN Document Server

    Scapin, Martina; Carra, Federico; Peroni, Lorenzo

    2015-01-01

    The introduction in recent years of new, extremely energetic particle accelerators such as the Large Hadron Collider (LHC) gives impulse to the development and testing of refractory metals and alloys based on molybdenum and tungsten to be used as structural materials. In this perspective, in this work the experimental results of a tests campaign on Inermet® IT180 and pure Molybdenum (sintered by two different producers) are presented. The investigation of the mechanical behaviour was performed in tension varying the strain-rates, the temperatures and both of them. Overall six orders of magnitude in strain-rate (between 10−3 and 103 s−1) were covered, starting from quasi-static up to high dynamic loading conditions. The high strain-rate tests were performed using a direct Hopkinson Bar setup. Both in quasi-static and high strain-rate conditions, the heating of the specimens was obtained with an induction coil system, controlled in feedback loop, based on measurements from thermocouples directly welded on...

  7. Effect of electrodeposition current density on the microstructure and magnetic properties of nickel-cobalt-molybdenum alloy powders

    Directory of Open Access Journals (Sweden)

    Pešić O.

    2014-01-01

    Full Text Available Nanostructured nickel-cobalt-molybdenum alloy powders were electrodeposited from an ammonium sulfate bath. The powders mostly consist of an amorphous phase and a very small amount of nanocrystals with an mean size of less than 3 nm. An increase in deposition current density increases the amorphous phase percentage, the density of chaotically distributed dislocations and internal microstrains in the powders, while decreasing the mean nanocrystal size. The temperature range over which the structural relaxation of the powders deposited at higher current densities occurs is shifted towards lower temperatures. A change in relative magnetic permeability during structural relaxation is higher in powders deposited at higher current densities. Powder crystallization takes place at temperatures above 700ºC. The formation of the stable crystal structure causes a decrease in relative magnetic permeability. [Projekat Ministarstva nauke Republike Srbije, br. 172057

  8. Determination of aluminium in molybdenum and tungsten metals, iron, steel and ferrous and non-ferrous alloys with pyrocatechol violet.

    Science.gov (United States)

    Donaldson, E M

    1971-09-01

    A method for determining 0.001-0.10% of aluminium in molybdenum and tungsten metals is described. After sample dissolution, aluminium is separated from the matrix materials by chloroform extraction of its acetylacetone complex, at pH 6.5, from an ammonium acetate-hydrogen peroxide medium, then back-extracted into 12M hydrochloric add. Following separation of most co-extracted elements, except for beryllium and small amounts of chroinium(III) and copper(II), by a combined ammonium pyrrolidincdithiocarbamate-cupfen-on-chlorofonn extraction, aluminium is determined spectrophotometrically with Pyrocatechol Violet at 578 nm. Chromium interferes during colour development but beryllium, in amounts equivalent to the aluminium concentration, does not cause significant error in the results. Interference from copper(II) is eliminated by reduction with ascorbic acid. The proposed method is also applicable to iron, steel, ferrovanadium, and copper-base alloys after preliminary removal of the matrix elements by a mercury cathode separation.

  9. Electrochemical studies and growth of apatite on molybdenum doped DLC coatings on titanium alloy β-21S

    Science.gov (United States)

    Anandan, C.; Mohan, L.; Babu, P. Dilli

    2014-03-01

    Titanium alloy β-21S (Ti-15Mo-3Nb-3Al-0.2Si) was coated with molybdenum doped DLC by Plasma-enhanced chemical vapor deposition and sputtering. XRD, XPS and Raman spectroscopy show that Mo is present in the form of carbide in the coating. XPS of samples immersed in Hanks' solution shows presence of calcium, phosphorous and oxygen in hydroxide/phosphate form on the substrate and Mo-doped DLC. Potentiodynamic polarization studies show that the corrosion resistance and passivation behavior of Mo-doped DLC is better than that of substrate. Electrochemical impedance spectroscopy (EIS) studies show that Mo-doped DLC samples behave like an ideal capacitor in Hanks' solution.

  10. Characteristics of the KUR Heavy Water Neutron Irradiation Facility as a neutron irradiation field with variable energy spectra

    Science.gov (United States)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2000-10-01

    The Heavy Water Neutron Irradiation Facility (HWNIF) of the Kyoto University Research Reactor (KUR) was updated in March 1996, mainly for the improvement in neutron capture therapy (NCT). A striking feature of the updated facility is that the energy spectrum of the neutron beam can be controlled from almost pure thermal to epi-thermal, within 5 min by remote control under a continuous reactor operation. This feature is advantageous not only to medical science such as NCT, but also to the other research fields such as physics, engineering, biology, etc. The performance of the updated facility as a neutron irradiation field with variable energy spectra, was characterized. Thermal neutron flux, cadmium ratio, gamma-ray dose rate, etc., at the normal irradiation position for various irradiation modes were determined, mainly on the basis of the measurement using gold activation foils and thermo-luminescent dosimeters (TLDs). The emphasis was on the performance of the new neutron energy spectrum shifter and cadmium thermal neutron filter, that control the mixing ratio of thermal and epi-thermal neutrons, through the change in the heavy water thickness of the spectrum shifter and the aperture size of the cadmium filter. The evaluation of neutron energy spectra at the normal irradiation position was also performed for three representative irradiation modes, in which the neutron intensities are largest of all the irradiation modes. In addition, the irradiation characteristics of two irradiation devices, namely the Irradiation Rail Device and the Remote Patient Carrier, which were updated concurrently with the facility update, were evaluated.

  11. Protecting Intestinal Epithelial Cell Number 6 against Fission Neutron Irradiation through NF-κB Signaling Pathway

    Science.gov (United States)

    Chang, Gong-Min; Gao, Ya-Bing; Wang, Shui-Ming; Xu, Xin-Ping; Zhao, Li; Zhang, Jing; Li, Jin-Feng; Wang, Yun-Liang; Peng, Rui-Yun

    2015-01-01

    The purpose of this paper is to explore the change of NF-κB signaling pathway in intestinal epithelial cell induced by fission neutron irradiation and the influence of the PI3K/Akt pathway inhibitor LY294002. Three groups of IEC-6 cell lines were given: control group, neutron irradiation of 4Gy group, and neutron irradiation of 4Gy with LY294002 treatment group. Except the control group, the other groups were irradiated by neutron of 4Gy. LY294002 was given before 24 hours of neutron irradiation. At 6 h and 24 h after neutron irradiation, the morphologic changes, proliferation ability, apoptosis, and necrosis rates of the IEC-6 cell lines were assayed and the changes of NF-κB and PI3K/Akt pathway were detected. At 6 h and 24 h after neutron irradiation of 4Gy, the proliferation ability of the IEC-6 cells decreased and lots of apoptotic and necrotic cells were found. The injuries in LY294002 treatment and neutron irradiation group were more serious than those in control and neutron irradiation groups. The results suggest that IEC-6 cells were obviously damaged and induced serious apoptosis and necrosis by neutron irradiation of 4Gy; the NF-κB signaling pathway in IEC-6 was activated by neutron irradiation which could protect IEC-6 against injury by neutron irradiation; LY294002 could inhibit the activity of IEC-6 cells. PMID:25866755

  12. Protecting Intestinal Epithelial Cell Number 6 against Fission Neutron Irradiation through NF-κB Signaling Pathway

    Directory of Open Access Journals (Sweden)

    Gong-Min Chang

    2015-01-01

    Full Text Available The purpose of this paper is to explore the change of NF-κB signaling pathway in intestinal epithelial cell induced by fission neutron irradiation and the influence of the PI3K/Akt pathway inhibitor LY294002. Three groups of IEC-6 cell lines were given: control group, neutron irradiation of 4Gy group, and neutron irradiation of 4Gy with LY294002 treatment group. Except the control group, the other groups were irradiated by neutron of 4Gy. LY294002 was given before 24 hours of neutron irradiation. At 6 h and 24 h after neutron irradiation, the morphologic changes, proliferation ability, apoptosis, and necrosis rates of the IEC-6 cell lines were assayed and the changes of NF-κB and PI3K/Akt pathway were detected. At 6 h and 24 h after neutron irradiation of 4Gy, the proliferation ability of the IEC-6 cells decreased and lots of apoptotic and necrotic cells were found. The injuries in LY294002 treatment and neutron irradiation group were more serious than those in control and neutron irradiation groups. The results suggest that IEC-6 cells were obviously damaged and induced serious apoptosis and necrosis by neutron irradiation of 4Gy; the NF-κB signaling pathway in IEC-6 was activated by neutron irradiation which could protect IEC-6 against injury by neutron irradiation; LY294002 could inhibit the activity of IEC-6 cells.

  13. Nano-cluster stability following neutron irradiation in MA957 oxide dispersion strengthened material

    Energy Technology Data Exchange (ETDEWEB)

    Ribis, J., E-mail: joel.ribis@cea.fr [CEA, DEN, DMN, SRMA, F-91191 Gif sur Yvette (France); Lozano-Perez, S. [Department of Materials, University of Oxford, Parks Road, OX1 3PH Oxford (United Kingdom)

    2014-01-15

    ODS steels are promising materials for Sodium cooled Fast Reactors since their fine distribution of nano-clusters confers excellent mechanical properties. However, the nano-feature stability needs to be assessed under neutron irradiation. Before irradiation, the characterizations show that nano-particles are finely distributed within the ferritic matrix and are identified to have a pyrochlore type structure. After irradiation of the MA957 alloy in the Phenix French reactor at 412 °C up to 50 dpa and 430 °C up to 75 dpa, transmission electron microscopy characterization reveals a very slight density fall but no distinguishable difference in nano-features size before and after irradiation. In addition, after both irradiations, the nano-oxides are still (Y, Ti, O) compounds with orientation relationship with the matrix. A multislice simulation of high resolution images suggests that nano-particles still have a fcc pyrochlore type structure after irradiation. A possible change of lattice parameter seems to be highlighted, possibly due to disordering by cascade effect.

  14. Nano-cluster stability following neutron irradiation in MA957 oxide dispersion strengthened material

    Science.gov (United States)

    Ribis, J.; Lozano-Perez, S.

    2014-01-01

    ODS steels are promising materials for Sodium cooled Fast Reactors since their fine distribution of nano-clusters confers excellent mechanical properties. However, the nano-feature stability needs to be assessed under neutron irradiation. Before irradiation, the characterizations show that nano-particles are finely distributed within the ferritic matrix and are identified to have a pyrochlore type structure. After irradiation of the MA957 alloy in the Phenix French reactor at 412 °C up to 50 dpa and 430 °C up to 75 dpa, transmission electron microscopy characterization reveals a very slight density fall but no distinguishable difference in nano-features size before and after irradiation. In addition, after both irradiations, the nano-oxides are still (Y, Ti, O) compounds with orientation relationship with the matrix. A multislice simulation of high resolution images suggests that nano-particles still have a fcc pyrochlore type structure after irradiation. A possible change of lattice parameter seems to be highlighted, possibly due to disordering by cascade effect.

  15. Effects of combined silicon and molybdenum alloying on the size and evolution of microalloy precipitates in HSLA steels containing niobium and titanium

    Energy Technology Data Exchange (ETDEWEB)

    Pavlina, Erik J., E-mail: e.pavlina@deakin.edu.au [Deakin University, Institute for Frontier Materials, 75 Pigdons Road, Waurn Ponds, Victoria (Australia); Van Tyne, C.J.; Speer, J.G. [Colorado School of Mines, Advanced Steel Processing and Products Research Center, 1500 Illinois Street, Golden, CO (United States)

    2015-04-15

    The effects of combined silicon and molybdenum alloying additions on microalloy precipitate formation in austenite after single- and double-step deformations below the austenite no-recrystallization temperature were examined in high-strength low-alloy (HSLA) steels microalloyed with titanium and niobium. The precipitation sequence in austenite was evaluated following an interrupted thermomechanical processing simulation using transmission electron microscopy. Large (~ 105 nm), cuboidal titanium-rich nitride precipitates showed no evolution in size during reheating and simulated thermomechanical processing. The average size and size distribution of these precipitates were also not affected by the combined silicon and molybdenum additions or by deformation. Relatively fine (< 20 nm), irregular-shaped niobium-rich carbonitride precipitates formed in austenite during isothermal holding at 1173 K. Based upon analysis that incorporated precipitate growth and coarsening models, the combined silicon and molybdenum additions were considered to increase the diffusivity of niobium in austenite by over 30% and result in coarser precipitates at 1173 K compared to the lower alloyed steel. Deformation decreased the size of the niobium-rich carbonitride precipitates that formed in austenite. - Highlights: • We examine combined Si and Mo additions on microalloy precipitation in austenite. • Precipitate size tends to decrease with increasing deformation steps. • Combined Si and Mo alloying additions increase the diffusivity of Nb in austenite.

  16. Peculiarities of high-temperature. beta. -phase formation during rapid heating of titanium-molybdenum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gridnev, V.N.; Zhuravlev, A.F.; Zhuravlev, B.F.; Ivasishin, O.M.; Oshkaderov, S.P. (AN Ukrainskoj SSR, Kiev. Inst. Metallofiziki)

    1983-11-01

    In the framework of the diffusion mechanism of ..cap alpha..+..beta.. ..-->.. ..beta.. transformation the model for calculating interface location determining the degree of transformation and concentration of the formed ..beta..-phase during continuous heating under different rates in titanium alloys with ..beta..-isomorphous alloying elements is suggested. On the example of Ti-10% Mo alloy the comparison of calculation and experimental results of determining parameters of ..cap alpha..+..beta.. ..-->.. ..beta.. transformation is performed.

  17. Effect of neutron irradiation on fracture toughness of metal matrix composites

    Science.gov (United States)

    Sato, Shinji; Hamada, Kenichi; Kohyama, Akira

    1992-09-01

    Based on the recent improvement in mechanical properties of unidirectionally reinforced metal matrix composites (MMCs), SiC/Al and C/Al, impact property change due to neutron irradiation has been investigated. This paper details effects of neutron irradiation on fracture toughness of the MMCs. Materials used were formed sheets of SiC/Al and C/Al. Miniaturized Charpy V-notched specimens were tested by an instrumented Charpy impact tester. Neutron irradiation was performed in JMTR(LWR) at Oarai. The Charpy value was increased with increasing test temperature and with neutron irradiation. SiC/Al was rather more neutron fluence insensitive than C/Al and the insensitivity was correlated to differences in interfacial structure between the two systems.

  18. In Vitro antileishmanial properties of neutron-irradiated meglumine antimoniate

    Directory of Open Access Journals (Sweden)

    Samanta Etel Treiger Borborema

    2005-10-01

    Full Text Available Pentavalent antimony, as meglumine antimoniate (Glucantime® or sodium stibogluconate (Pentostam® , is the main treatment for leishmaniasis, a complex of diseases caused by the protozoan Leishmania, and an endemic and neglected threat in Brazil. Despite over half a century of clinical use, their mechanism of action, toxicity and pharmacokinetic data remain unknown. The analytical methods for determination of antimony in biological systems remain complex and have low sensitivity. Radiotracer studies have a potential in pharmaceutical development. The aim of this study was to obtain a radiotracer for antimony, with suitable physical and biological properties. Meglumine antimoniate was neutron irradiated inside the IEA-R1 nuclear reactor, producing two radioisotopes 122Sb and 124Sb, with high radionuclidic purity and good specific activity. This compound showed the same antileishmanial activity as the native compound. The use of the radiotracers, easily created by neutron irradiation, could be an interesting tool to solve important questions in antimonial pharmacology.Os antimoniais pentavalentes, como o antimoniato de meglumina (Glucantime® ou estibogluconato de sódio (Pentostam® , são o principal tratamento para a leishmaniose, um complexo de doenças causadas pelo protozoário parasita Leishmania, uma doença endêmica e negligenciada no Brasil. Apesar do seu uso clínico por mais de meio século, seu mecanismo de ação, toxicidade e dados de farmacocinética permanecem desconhecidos. Os métodos analíticos para determinação de antimônio em sistemas biológicos são complexos e apresentam baixa sensibilidade. Estudos utilizando radiotraçadores têm papel potencial no desenvolvimento farmacológico. O objetivo deste estudo foi desenvolver um radiotraçador de antimônio, com propriedades físicas e biológicas adequadas. O antimoniato de meglumina foi irradiado por nêutrons no reator nuclear IEA-R1, produzindo dois radioisótopos: 122

  19. Embrittlement of low copper VVER 440 surveillance samples neutron-irradiated to high fluences

    Science.gov (United States)

    Miller, M. K.; Russell, K. F.; Kocik, J.; Keilova, E.

    2000-11-01

    An atom probe tomography microstructural characterization of low copper (0.06 at.% Cu) surveillance samples from a VVER 440 reactor has revealed manganese and silicon segregation to dislocations and other ultrafine features in neutron-irradiated base and weld materials (fluences 1×10 25 m-2 and 5×10 24 m-2, E>0.5 MeV, respectively). The results indicate that there is an additional mechanism of embrittlement during neutron irradiation that manifests itself at high fluences.

  20. Mechanical properties of 50Molybdenum-50Rhenium alloys and their assembly by spinal muscular atrophy

    Science.gov (United States)

    Xu, Jianhui

    This study is concerned with the deformation and fracture behaviors, especially strain-rate effect on plasticity in tensile tests, of two 50Mo-50Re alloys at strain rates ranging from 10-6 s-1 to 1 s-1 at room temperature in air. Metallographic observations of the 50Mo-50Re alloys before and after tensile deformation were conducted to understand the relationships among mechanical properties, microstructure and strain rate in these alloys. Understanding the strain-rate effect on mechanical properties of 50Mo-50Re alloys is important for optimizing forming operations, especially sheet forming, of these alloys, which are often used in cathode and aerospace applications. An anomalous strain-rate effect on ductility was observed in the 50Mo-50Re alloys. Ductility was significantly increased by increasing the strain rate from 10-6 s-1 to 1 s-1 in the fully-recrystallized and recovery heat-treated 50Mo-50Re alloys in tension at room temperature. At a low strain rate, fracture was predominantly brittle, while it was more ductile at higher stain rates. At a low strain rate, secondary cracks initiated at grain boundaries and triple junctions were observed in these alloys, which suggested that significant stress concentration was generated by tensile plastic deformation in the vicinity of grain boundaries, especially triple junctions. Electron backscatter diffraction experiments revealed that there was strain concentration at grain boundaries and their triple junctions during tensile deformation in these alloys. The decrease in ductility at low strain rates in the alloys was related to the possible interaction between dislocations and trace interstitial atoms (e.g., H, O, N and C) picked up during production of these alloys. This dissertation also reports the research efforts made to optimize small-scale resistance spot welding (SSRSW) of refractory alloy 50Mo-50Re thin sheet by adjusting seven important welding parameters, including hold time, electrode material, electrode

  1. Compositional effects on mechanical properties of hafnium-carbide-strengthened molybdenum alloys

    Science.gov (United States)

    Witzke, W. R.

    1975-01-01

    The mechanical properties of swaged rod thermomechanically processed from arc melted Mo-2Re-Hf-C alloys containing as much as 0.9-mol% HfC were evaluated. The low-temperature ductilities of these alloys were not influenced by the amount of HfC present but by the amount of Hf in excess of stoichiometry. Maximum ductility occurred at 0.2- to 0.3-at.% excess Hf. At 0.3- to 0.5-mol% HfC, alloy strength varied directly with the Mo content of extracted carbide particles, both decreasing as the amount of excess Hf increased. Additions of 2-at.% Re had little effect on strength or ductility. Tensile and creep strengths of Mo-2Re-0.7Hf-0.5C alloy equaled or exceeded those of other high strength Mo alloys.

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

    Science.gov (United States)

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

    2008-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Z. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)], E-mail: zheng.lu@lboro.ac.uk; Faulkner, R.G.; Morgan, T.S. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)

    2008-12-01

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

  4. Effect of neutron irradiation on the microstructure of tungsten

    Directory of Open Access Journals (Sweden)

    M. Klimenkov

    2016-12-01

    Full Text Available Two grades of pure tungsten, single and polycrystalline, were irradiated for 282 days in the HFR reactor, Petten, at 900 °C to an average damage level of 1.6dpa. Each grade of tungsten was investigated using the transmission electron microscope (TEM to assess the effect of neutron irradiation on tungsten microstructure. Investigations revealed the formation of faceted cavities, whose diameter varies from 4 to 14nm in both materials. The cavities are homogeneously distributed only inside single crystalline tungsten. The local distribution of cavities in polycrystalline tungsten is strongly influenced by grain boundaries. The number densities of cavities were measured to be 4×1021 m−3 for polycrystalline and 2.5×1021 m−3 for single crystalline tungsten. This corresponds to volumetric densities of 0.45% and 0.33% respectively. High-resolution transmission electron microscopy (HRTEM revealed that faces of cavities are oriented in (110 plane. Analytical investigations showed precipitation of rhenium and osmium produced by a transmutation reaction around cavities and at grain boundaries.

  5. Resistivity measurements on the neutron irradiated detector grade silicon materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zheng

    1993-11-01

    Resistivity measurements under the condition of no or low electrical field (electrical neutral bulk or ENB condition) have been made on various device configurations on detector grade silicon materials after neutron irradiation. Results of the measurements have shown that the ENB resistivity increases with neutron fluence ({Phi}{sub n}) at low {phi}{sub n} (<10{sup 13} n/cm{sup 2}) and saturates at a value between 300 and 400 k{Omega}-cm at {phi}{sub n} {approximately}10{sup 13} n/cm{sup 2}. Meanwhile, the effective doping concentration N{sub eff} in the space charge region (SCR) obtained from the C-V measurements of fully depleted p{sup +}/n silicon junction detectors has been found to increase nearly linearly with {phi}{sub n} at high fluences ({phi}{sub n} > 10{sup 13} n/cm{sup 2}). The experimental results are explained by the deep levels crossing the Fermi level in the SCR and near perfect compensation in the ENB by all deep levels, resulting in N{sub eff} (SCR) {ne} n or p (free carrier concentrations in the ENB).

  6. High-dose neutron irradiation embrittlement of RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 deg. C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = {delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates no hardening embrittlement.

  7. The effect of neutron irradiation on silicon carbide fibers

    Energy Technology Data Exchange (ETDEWEB)

    Newsome, G.A. [Lockheed Martin Corp., Schenectady, NY (United States)

    1997-01-01

    Nine types of SiC fiber have been exposed to neutron radiation in the Advanced Test Reactor at 250 C for various lengths of time ranging from 83 to 128 days. The effects of these exposures have been initially determined using scanning electron microscopy. The fibers tested were Nicalon{trademark} CG, Tyranno, Hi-Nicalon{trademark}, Dow Corning SiC, Carborundum SiC, Textron SCS-6, polymethysilane (PMS) derived SiC from the University of Michigan, and two types of MER SiC fiber. This covers a range of fibers from widely used commercial fibers to developmental fibers. Consistent with previous radiation experiments, Nicalon fiber was severely degraded by the neutron irradiation. Similarly, Tyranno suffered severe degradation. The more advanced fibers which approach the composition and properties of SiC performed well under irradiation. Of these, the Carborundum SiC fiber appeared to perform the best. The Hi-Nicalon and Dow Corning Fibers exhibited good general stability, but also appear to have some surface roughening. The MER fibers and the Textron SCS-6 fibers both had carbon cores which adversely influenced the overall stability of the fibers.

  8. Resistivity measurements on the neutron irradiated detector grade silicon materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zheng

    1993-11-01

    Resistivity measurements under the condition of no or low electrical field (electrical neutral bulk or ENB condition) have been made on various device configurations on detector grade silicon materials after neutron irradiation. Results of the measurements have shown that the ENB resistivity increases with neutron fluence ({Phi}{sub n}) at low {phi}{sub n} (<10{sup 13} n/cm{sup 2}) and saturates at a value between 300 and 400 k{Omega}-cm at {phi}{sub n} {approximately}10{sup 13} n/cm{sup 2}. Meanwhile, the effective doping concentration N{sub eff} in the space charge region (SCR) obtained from the C-V measurements of fully depleted p{sup +}/n silicon junction detectors has been found to increase nearly linearly with {phi}{sub n} at high fluences ({phi}{sub n} > 10{sup 13} n/cm{sup 2}). The experimental results are explained by the deep levels crossing the Fermi level in the SCR and near perfect compensation in the ENB by all deep levels, resulting in N{sub eff} (SCR) {ne} n or p (free carrier concentrations in the ENB).

  9. High-dose neutron irradiation embrittlement of RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 °C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = ΔDBTT/Δ σ indicates hardening-dominated embrittlement at irradiation temperatures below 350 °C with 0.17 ⩽ C ⩽ 0.53 °C/MPa. Scattering of C at irradiation temperatures above 400 °C indicates no hardening embrittlement.

  10. Neutron irradiation behavior of ITER candidate beryllium grades

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

    1998-01-01

    Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

  11. Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

  12. Structure and properties of sintered titanium alloyed with aluminium, molybdenum and oxygen

    Energy Technology Data Exchange (ETDEWEB)

    Anokhin, V.M.; Petrunko, A.N. [State Research and Design Titanium Institute, Zaporozhye (Ukraine); Ivasishin, O.M. [Institute for Metal Physics, National Academy of Sciences of Ukraine, 36 Vernadsky St, 142 Kiev (Ukraine)

    1998-03-15

    Titanium alloys of Ti-Al-Mo-O system were manufactured by blended elemental powder method using Ti, Al, Mo and TiO{sub 2} powders as starting materials. It was found that cold compaction pressure of 800 MPa followed by sintering at 1150-1200 C, for 4 h provided sufficient densification of titanium materials. Complete dissolution of alloying elements in the titanium matrix resulted in a good combination of mechanical properties. Examples of alloys chosen for possible application were Ti-(1.5-2.0)%Mo-0.7%TiO{sub 2} and Ti-2%Al-2%Mo. The latter has already been tried for manufacturing parts in automotive industry. (orig.) 3 refs.

  13. Survey of degradation modes of four nickel-chromium-molybdenum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E. [KMI Energy Services, Livermore, CA (United States)

    1991-03-01

    This report examines the degradation modes of four Ni-Cr-Mo alloys under conditions relevant to the Yucca Mountain Site Characterization Project (YMP). The materials considered are Alloys C-276, C-4, C-22, and 625 because they have desirable characteristics for the conceptual design (CD) of the high-level radioactive-waste containers presented in the YMP Site Characterization Plan (SCP). The types of degradation covered in this report are general corrosion; localized corrosion, including pitting and crevice corrosion; stress corrosion cracking in chloride environments; hydrogen embrittlement (HE); and undesirable phase transformations due to a lack of phase stability. Topics not specifically addressed are welding concerns and microbiological corrosion. The four Ni-Cr-Mo alloys have excellent corrosion resistance in chloride environments such as seawater as well as in more aggressive environments. They have significantly better corrosion resistance than the six materials considered for the CD waste container in the YMP SCP. (Those six materials are Types 304L and 3161L stainless steels, Alloy 825, unalloyed copper, Cu(70)-Ni(30), and 7% aluminum bronze.) In seawater, the Ni-Cr-Mo alloys have negligible general corrosion rates and show little evidence of localized corrosion. The four base materials of these alloys are expected to have nearly indistinguishable corrosion resistance in the YMP environments. The strength requirements of the SCP-CD waste container are met by these materials in the annealed condition; in this condition, they are highly resistant to HE. Historically, HE has been noted when these materials have been strengthened (cold-worked) and used in sour gas (H{sub 2}S and CO{sub 2}) well service -- conditions that are not expected for the YMP. Metallurgical phase stability may be a concern under conditions favoring (1) the formation of intermetallics and carbides, and (2) microstructural ordering.

  14. Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Hara, M. [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI (United States); Kobayashi, M. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Watanabe, H. [Research Institute for Applied Mechanics, Kyushu University, Fukuoka (Japan); Tyburska-Pueschel, B. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI (United States); Institute fuer Plasmaphysik, EURATOM Association, Garching (Germany); Ueda, Y. [Graduate School of Engineering, Osaka University, Osaka (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Okuno, K. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan)

    2012-08-15

    The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 Degree-Sign C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe{sup 2+}, 20 MeV W{sup 2+}, and 700 keV H{sup -}) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

  15. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    Science.gov (United States)

    Tobita, Tohru; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-01

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 1024 n/m2 at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic-plastic fracture toughness (JIc) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low JIc values at high temperatures.

  16. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Tohru, E-mail: tobita.tohru@jaea.go.jp; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-15

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 10{sup 24} n/m{sup 2} at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic–plastic fracture toughness (J{sub Ic}) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low J{sub Ic} values at high temperatures.

  17. Resistivity damage rates in fusion-neutron-irradiated metals at 4. 2 K

    Energy Technology Data Exchange (ETDEWEB)

    Guinan, M.W.; Kinney, J.H.

    1981-01-01

    Changes in electrical resistivity at liquid helium temperature have been used to monitor the production of damage in dilute alloys of vanadium, niobium and molybdenum, and pure tungsten, aluminum and copper irradiated with high energy neutrons. The neutrons were produced at the Livermore rotating-target neutron sources (RTNS-I and RTNS-II). Further experiments on V, Nb and Mo were carried out with 30 MeV d-Be neutrons and slightly degraded fission-spectra neutrons. The results for all six materials are compared to those obtained in a pure fission spectrum. The relative damage production rates are in agreement with predictions based on damage energy calculations.

  18. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Directory of Open Access Journals (Sweden)

    Hu J.-P.

    2016-01-01

    Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

  19. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

    2014-05-23

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

  20. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Science.gov (United States)

    Hu, J.-P.; Holden, N. E.; Reciniello, R. N.

    2016-02-01

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4-7% lower than

  1. Commercial Applications at FRM II Based on Neutron Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Gerstenberg, H.; Draack, A.; Kastenmuller, A. [Technische Universitaet Muenchen, Munchen (Germany)

    2013-07-01

    Due to its design as a heavy water moderated reactor with a very compact core FRM II, Germany's most modern and most powerful research reactor, offers excellent conditions for basic research using beam tubes. On the other hand it is equipped with various irradiation facilities to be used mainly for industrial purposes. From the very beginning of reactor operation a dedicated department had been implemented in order to provide a neutron irradiation service to interested parties on a commercial basis. As of today the most widely used application is Si doping. The semiautomatic doping facility accepts ingots with diameters between 125 mm and 200 mm and a maximum height of 500 mm. The irradiation channel is located deep in the heavy water tank and exhibits a ratio of thermal/fast neutron flux density of > 1000. This value allows the doping of Si to a target resistivity as high as 1100 Ωcm within the tight limits regarding accuracy and homogeneity specified by the customer. Typically the throughput of Si doped in FRM II sums up to about 15 t/year. Another topic of growing importance is the use of FRM II aiming the production of radioisotopes mainly for the radiopharmaceutical industry. The maybe most challenging example is the production of Lu-177 n. c. a. based on the irradiation of Yb{sub 2}O{sub 3} to a high fluence of thermal neutrons of typically 1.5E20 cm{sup -2}. The Lu-177 activity delivered to the customer is in the range of 750 GBq. With respect to further processing it turned out to be a highly advantageous to have the laboratories of ITG, the company extracting the Lu-177 from the freshly irradiated Yb{sub 2}O{sub 3} on site FRM II. Further irradiation facilities are available at FRM II in order to allow the activation of samples for analytical purposes or to irradiate samples for geochronological investigations using the fission track technique. Finally a project on the future installation of a facility dedicated to the irradiation of U-targets for

  2. Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U-{sup 235} > 85 wt%) by low enriched uranium (U-{sup 235} < 20 wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

  3. Gallium suboxide vapor attack on chromium, cobalt, molybdenum, tungsten and their alloys at 1200 [degrees] C

    Energy Technology Data Exchange (ETDEWEB)

    Kolman, D. G. (David G.); Taylor, T. N. (Thomas N.); Park, Y. (Youngsoo); Stan, M. (Marius); Butt, D. P. (Darryl P.); Maggiore, C. J. (Carl J.); Tesmer, Joseph R.; Havrilla, G. J. (George J.)

    2004-01-01

    Our prior work elucidated the failure mechanism of furnace materi als (304 SS, 316 SS, and Hastelloy C-276) exposed to gallium suboxide (Ga{sub 2}O) and/or gallium oxide (Ga{sub 2}O{sub 3}) during plutonium - gallium compound processing. Failure was hypothesized to result from concurrent alloy oxidation/Ga compound reduction followed by Ga uptake. The aim of the current work is to screen candidate replacement materials. Alloys Haynes 25 (49 Co - 20 Cr - 15 W - 10 Ni - 3 Fe - 2 Mn - 0.4 Si, wt%), 52 Mo - 48 Re (wt%), 62 W - 38 Cu (wt%), and commercially pure Cr, Co, Mo, W, and alumina were examined. Preliminary assessments of commercially pure W and Mo - Re suggest that these materials may be suitable for furnace construction. Thermodynamics calculations indicating that materials containing Al, Cr, Mn, Si, and V would be susceptible to oxidation in the presence of Ga{sub 2}O were validated by experimental results. In contrast to that reported previously, an alternate reaction mechanism for Ga uptake, which does not require concurrent alloy oxidation, controls Ga uptake for certain materials. A correlation between Ga solubility and uptake was noted.

  4. Low temperature testing and neutron irradiation of a swept charge device on board the HXMT satellite

    Institute of Scientific and Technical Information of China (English)

    WANG Yu-Sa; CHEN Tian-Xiang; LI Cheng-Kui; HUO Jia; LI Zheng-Wei; LI Wei; HU Wei; ZHANG Yi; LU Bo; ZHU Yue; LIU Yan; CHEN Yong; WU Di; SUN Qing-Rong; ZHANG Zi-Liang; XU Yu-Peng; YANG Yan-Ji; CUI Wei-Wei; LI Mao-Shun; LIU Xiao-Yan; WANG Juan; HAN Da-Wei

    2012-01-01

    We present the low temperature testing of an SCD detector,investigating its performance such as readout noise,energy resolution at 5.9 keV and dark current.The SCD's performance is closely related to temperature,and the temperature range of -80 ℃ to -50 ℃ is the best choice,where the FWHM at 5.9 keV is about 130 eV.The influence of the neutron irradiation from an electrostatic accelerator with fluence up to 1 × 109 cm-2 has been examined.We find the SCD is not vulnerable to neutron irradiation.The detailed operations of the SCD and the test results of low temperature are reported,and the results of neutron irradiation are discussed.

  5. Quantitative TEM analysis of precipitation and grain boundary segregation in neutron irradiated EUROFER 97

    Science.gov (United States)

    Dethloff, Christian; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-01

    Characterization of irradiation induced microstructural defects is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM) is used to analyze the types and structure of precipitates, and the evolution of their size distributions and densities caused by neutron irradiation to a dose of 32 displacements per atom (dpa) at 330-340 °C in the irradiation experiment ARBOR 1. A significant growth of MX and M23C6 type precipitates is observed after neutron irradiation, while the precipitate density remains unchanged. Hardening caused by MX and M23C6 precipitate growth is assessed by applying the Dispersed Barrier Hardening (DBH) model, and shown to be of minor importance when compared to other irradiation effects like dislocation loop formation. Additionally, grain boundary segregation of chromium induced by neutron irradiation was investigated and detected in irradiated specimens.

  6. Deuterium Depth Profile in Neutron-Irradiated Tungsten Exposed to Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Masashi Shimada; G. Cao; Y. Hatano; T. Oda; Y. Oya; M. Hara; P. Calderoni

    2011-05-01

    The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

  7. Status of Neutron Irradiation of Non-Nuclear Materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Kee-Nam; Cho, Man-Soon; Shin, Yoon-Taek; Park, Seng-Jae; Kang, Young Hwan; Jun, Byung-Hyuk; Kim, Chan-Joong; Park, Sang-Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The irradiation facilities have been mostly utilized for the KAERI nuclear research projects relevant to a commercial nuclear power reactor such as the ageing management and safety evaluation of the components. Based on the accumulated experience, HANARO has recently supported national R and D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. As neutron irradiation affects the structure of a material, radiation induced modification of materials has become a perspective method for the purposeful changes in material properties. Some irradiation tests of electro-magnetic materials were also performed at HANARO for scientific research of universities and the demand for neutron irradiation of the materials is increasing rapidly. Another research reactor that will specialize in radio-isotope and neutron transmutation doping (NTD) Si production and the demonstration of reactor design is under construction in Korea. Therefore, HANARO will specialize more on irradiation research. Based on its accumulated irradiation experience, HANARO has recently started new support of R and D relevant to the irradiation of electro-magnetic materials. In this paper, the status of utilization of the HANARO irradiation facilities for non-nuclear materials and the possibility of researches on new electro-magnetic materials using neutron irradiation are surveyed to encourage the utilization of HANARO. HANARO irradiation facilities have been actively utilized for various material irradiation tests requested by users. Although most irradiation tests have been related to national R and D relevant to nuclear power, demand for neutron irradiation of electro-magnetic materials is rapidly increasing at HANARO. Another research reactor, the KIJANG research reactor (KJRR), is under construction in Korea. New irradiation facilities including Neutron Transmutation Doping (NTD) facilities for power

  8. Corrosion resistance of stainless steel, nickel-titanium, titanium molybdenum alloy, and ion-implanted titanium molybdenum alloy archwires in acidic fluoride-containing artificial saliva: An in vitro study

    Directory of Open Access Journals (Sweden)

    Venith Jojee Pulikkottil

    2016-01-01

    Full Text Available Objective: (1 To evaluate the corrosion resistance of four different orthodontic archwires and to determine the effect of 0.5% NaF (simulating high fluoride-containing toothpaste of about 2250 ppm on corrosion resistance of these archwires. (2 To assess whether surface roughness (Ra is the primary factor influencing the corrosion resistance of these archwires. Materials and Methods: Four different archwires (stainless steel [SS], nickel-titanium [NiTi], titanium molybdenum alloy [TMA], and ion-implanted TMA were considered for this study. Surface characteristics were analyzed using scanning electron microscopy, atomic force microscopy (AFM, and energy dispersive spectroscopy. Linear polarization test, a fast electrochemical technique, was used to evaluate the corrosion resistance, in terms of polarization resistance of four different archwires in artificial saliva with NaF concentrations of 0% and 0.5%. Statistical analysis was performed by one-way analysis of variance. Results: The potentiostatic study reveals that the corrosion resistance of low-friction TMA (L-TMA > TMA > NiTi > SS. AFM analysis showed the surface Ra of TMA > NiTi > L-TMA > SS. This indicates that the chemical composition of the wire is the primary influential factor to have high corrosion resistance and surface Ra is only secondary. The corrosion resistance of all wires had reduced significantly in 0.5% acidic fluoride-containing artificial saliva due to formation of fluoride complex compound. Conclusion: The presence of 0.5% NaF in artificial saliva was detrimental to the corrosion resistance of the orthodontic archwires. Therefore, complete removal of residual high-fluorinated toothpastes from the crevice between archwire and bracket during tooth brushing is mandatory.

  9. Corrosion resistance of stainless steel, nickel-titanium, titanium molybdenum alloy, and ion-implanted titanium molybdenum alloy archwires in acidic fluoride-containing artificial saliva: An in vitro study

    Science.gov (United States)

    Pulikkottil, Venith Jojee; Chidambaram, S.; Bejoy, P. U.; Femin, P. K.; Paul, Parson; Rishad, Mohamed

    2016-01-01

    Objective: (1) To evaluate the corrosion resistance of four different orthodontic archwires and to determine the effect of 0.5% NaF (simulating high fluoride-containing toothpaste of about 2250 ppm) on corrosion resistance of these archwires. (2) To assess whether surface roughness (Ra) is the primary factor influencing the corrosion resistance of these archwires. Materials and Methods: Four different archwires (stainless steel [SS], nickel-titanium [NiTi], titanium molybdenum alloy [TMA], and ion-implanted TMA) were considered for this study. Surface characteristics were analyzed using scanning electron microscopy, atomic force microscopy (AFM), and energy dispersive spectroscopy. Linear polarization test, a fast electrochemical technique, was used to evaluate the corrosion resistance, in terms of polarization resistance of four different archwires in artificial saliva with NaF concentrations of 0% and 0.5%. Statistical analysis was performed by one-way analysis of variance. Results: The potentiostatic study reveals that the corrosion resistance of low-friction TMA (L-TMA) > TMA > NiTi > SS. AFM analysis showed the surface Ra of TMA > NiTi > L-TMA > SS. This indicates that the chemical composition of the wire is the primary influential factor to have high corrosion resistance and surface Ra is only secondary. The corrosion resistance of all wires had reduced significantly in 0.5% acidic fluoride-containing artificial saliva due to formation of fluoride complex compound. Conclusion: The presence of 0.5% NaF in artificial saliva was detrimental to the corrosion resistance of the orthodontic archwires. Therefore, complete removal of residual high-fluorinated toothpastes from the crevice between archwire and bracket during tooth brushing is mandatory. PMID:27829756

  10. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Science.gov (United States)

    Clarke, A. J.; Clarke, K. D.; McCabe, R. J.; Necker, C. T.; Papin, P. A.; Field, R. D.; Kelly, A. M.; Tucker, T. J.; Forsyth, R. T.; Dickerson, P. O.; Foley, J. C.; Swenson, H.; Aikin, R. M.; Dombrowski, D. E.

    2015-10-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time-temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2-4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

  12. Migration and accumulation at dislocations of transmutation helium in austenitic steels upon neutron irradiation

    Science.gov (United States)

    Kozlov, A. V.; Portnykh, I. A.

    2016-04-01

    The model of the migration and accumulation at dislocations of transmutation helium and the formation of helium-vacancy pore nuclei in austenitic steels upon neutron irradiation has been proposed. As illustrations of its application, the dependences of the characteristics of pore nuclei on the temperature of neutron irradiation have been calculated. The results of the calculations have been compared with the experimental data in the literature on measuring the characteristics of radiation-induced porosity that arises upon the irradiation of shells of fuel elements of a 16Cr-19Ni-2Mo-2Mn-Si-Ti-Nb-V-B steel in a fast BN600 neutron reactor at different temperatures.

  13. The optimization study of Bonner sphere in the epi-thermal neutron irradiation field for BNCT.

    Science.gov (United States)

    Ueda, H; Tanaka, H; Maruhashi, A; Ono, K; Sakurai, Y

    2011-12-01

    The optimization study on the Bonner sphere in the epi-thermal neutron irradiation field for BNCT was done for the moderator material, moderator size, and activation foils as a neutron detector in the sphere. The saturated activity for the activation foil was obtained from the calculated response, and the effective energy range for each Bonner sphere was determined from the saturated activity. We can see that boric acid solution moderator is suitable for the spectrum measurement of a epi-thermal neutron irradiation field.

  14. Generation of c-component edge dislocations in α-zirconium during neutron irradiation - An atomistic study

    Science.gov (United States)

    Woo, C. H.; Liu, Xiangli

    2009-09-01

    The nucleation and multiplication of c-component edge dislocation segments during neutron irradiation in zirconium and its alloys is known to have important consequences to their in-reactor deformation behavior. Although there are ample experimental observations showing the close correlation between the edge-type and the screw-type of c-dislocations, the relation between them is unclear. In this paper, we performed atomistic study of the interaction between a [0 0 0 1] screw dislocation and a vacancy cluster in the form of a platelet on the basal plane. The local minimum-energy configuration was obtained using the conjugate-gradient method, with boundary relaxation achieved via a modified Green's function method. Under stress-free conditions, the vacancy clusters maintained their cavity nature. With a [0 0 0 1] screw dislocation in the close neighborhood, vacancy clusters containing more than 23 vacancies collapse into faulted vacancy loops. Interaction at even closer range leads to the disappearance of the vacancy cluster and the development of an edge component on the originally straight screw dislocation in the form of a helical line. The implications of these findings are discussed in relation to the experimentally observed behavior of growth acceleration in zirconium and its alloys.

  15. Effect of Neutron Irradiation on Beam-Column Interaction of Reinforced Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Park, Jiho; Kim, Jun Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, HyungTae; Park, Kyoungsoo [Yonsei University, Seoul (Korea, Republic of); Kim, Sang-Ho [Hyundai Engineering, Seoul (Korea, Republic of)

    2015-10-15

    Age-related effects on such RC structures have been extensively studied in detail. However, the effect of neutron irradiation requires further studies from its limited database. Most of RC structures have been regarded as sound as the neutron fluence below 1.0x10{sup 19} n/cm{sup 2}. The reduction of strength is not considered in a periodic inspection program at aging NPPs. However, RC structures, such as biological shields and supports for a reactor vessel, could be exposed to see the critical level of neutron fluence at years of operation. In this regard, beam-column interaction of a typical RC member is numerically investigated as a result of neutron irradiation. The effect of neutron irradiation on beam-column interaction is evaluated. ACI318 requires the strength reduction factor, ϕ=0.70, for the compression controlled area and the higher up to 0.9 as the tensile strain in steel reinforcement goes higher. This concept works well with this example. However, this does not take into account the energy dissipation capacity of the member but it only expresses the ultimate strength. Therefore, the current strength evaluation concept may be misleading when the material behavior of steel reinforcement becomes brittle due to the neutron irradiation. In such case, even for the transient and tension controlled area, the strength reduction factor needs to be modified to account for the potential ductility loss.

  16. Magnetic susceptibility and magnetoresistance of neutron-irradiated doped SI whiskers

    Energy Technology Data Exchange (ETDEWEB)

    Druzhinin, A.A., E-mail: druzh@polynet.lviv.ua [Lviv Polytechnic National University, S. Bandera Str., 12, Lviv 79013 (Ukraine); International Laboratory of High Magnetic Fields and Low Temperatures, Wroclaw (Poland); Ostrovskii, I.P.; Khoverko, Yu.M. [Lviv Polytechnic National University, S. Bandera Str., 12, Lviv 79013 (Ukraine); International Laboratory of High Magnetic Fields and Low Temperatures, Wroclaw (Poland); Rogacki, K. [International Laboratory of High Magnetic Fields and Low Temperatures, Wroclaw (Poland); Litovchenko, P.G.; Pavlovska, N.T. [Institute of Nuclear Researches, NAS of Ukraine, 47, Prospect Nauky, 03028 Kyiv (Ukraine); Pavlovskyy, Yu.V.; Ugrin, Yu.O. [Ivan Franko Drohobych State Pedagogical University, 24, Franko str., 82100 Drohobych (Ukraine)

    2015-11-01

    The effect of 8.6·10{sup 17} n/cm{sup 2} fast neutron irradiation on the magnetic susceptibility and magnetoresistance of Si whiskers with impurity concentration near metal–insulator transition (MIT) has been studied. Neutron irradiated specimens with boron concentration away of MIT are mainly diamagnetic with a small amount of paramagnetic centers originated from dangling bonds on the whisker surface. It has been established that at temperatures near 4.2 K, a significant contribution to the conductivity is made by light charge carriers of low concentration but with high mobility. The as grown whiskers with impurity concentration correspondent to MIT showed hysteresis loops in magnetization at temperature of liquid helium. Besides hysteresis loops in magnetoresistance was observed for whiskers under compression stress at low temperature up to 7 K. The possible reason of the effect can be magnetic interaction between impurities centers in subsurface region of the whisker with the orbital moment of dangle bounds in the whisker core–shell interstices. - Highlights: • Neutron irradiation influence on magnetic susceptibility of Si whiskers is studied. • Neutron irradiated Si whiskers with boron concentration away of MIT are diamagnetic. • Whiskers in the vicinity to MIT showed hysteresis loops in magnetoresistance. • Whiskers in the vicinity to MIT showed hysteresis loops in magnetic susceptibility.

  17. Characterization of defect accumulation in neutron-irradiated Mo by positron annihilation spectroscopy

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Li, Meimei; Snead, L.L.

    2008-01-01

    Positron annihilation lifetime spectroscopy measurements were performed on neutron-irradiated low carbon arc cast Mo. Irradiation took place in the high flux isotope reactor, Oak Ridge National Laboratory, at a temperature of 80 +/- 10 degrees C. Neutron fluences ranged from 2 x 10(21) to 8 x 10...

  18. Preparation and Purification of 125I With Neutron Irradiated Xenon in a Vacuum Circular system

    Institute of Scientific and Technical Information of China (English)

    MIAOZeng-xing; LIYu-cheng; YUNing-wen; WUJie; XIANGXue-qin; ZHAOXiu-yan

    2003-01-01

    This paper describes the preparation and purification of 125I with neutron irradiated xenon in a vacuum circular system, which is specially designed with an irradiate chamber set inside of the reactor and three decay chambers set outside of the reactor. The xenon is filled in this system and recurrently circulates between the irradiate chamber and the decay chambers during the reactor is operating.

  19. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    Science.gov (United States)

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  20. Fission neutron irradiation of copper containing implanted and transmutation produced helium

    DEFF Research Database (Denmark)

    Singh, B.N.; Horsewell, A.; Eldrup, Morten Mostgaard

    1992-01-01

    . The distributions of helium prior to fission neutron irradiation were determined by a combination of transmission electron microscopy (TEM) and positron annihilation techniques (PAT). These specimens, together with pure copper, were then irradiated with fission neutrons in a single capsule in fast flux test...

  1. Corrosion and high temperature resistant coatings for molybdenum, made out of iron and nickel alloys and applied by explosive welding

    Energy Technology Data Exchange (ETDEWEB)

    Pruemmer, R.; Henne, R.

    1980-04-01

    The impact parameters of the explosive welding of molybdenum with Inconel 601 were determined. The combination Mo and Inconel 601 was considered as nonweldable. It can be applied in solar radiation concentrating devices, allowing a higher operating temperature and higher energy conversion efficiency. The usual velocities of the explosive welding process (collision velocities of 2200 m/sec) lead at best to samples affected by cracks, due to the insufficient workability of molybdenum. At higher velocities cracks no longer occur, molybdenum being a strain rate sensitive material. Layer composite materials can be manufactured in flat as well as in tube form. (ESA)

  2. Uranium-molybdenum alloys containing 0,5 to 3 per cent by weight of molybdenum; Alliages uranium-molybdene de 0,5 a 3 pour cent en poids de molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-07-01

    The following properties have been determined in the new cast state of uranium alloys containing 0.5-1-1.8-2 and 3.5 per cent of molybdenum: micro-graphical aspect, crystalline structure, thermal expansion, the mechanical characteristics, behaviour when subjected to cyclic temperature variations, and heat treatment. The transformation curves have been established for continuous cooling at rates varying between 2.5 and 200 deg. C per minute, using a dilaon method for the alloys containing 1.0, 2.0 and 3.0 per cent Mo. T.T.T. curves have been traced for 0.5 and 1.0 per cent Mo alloys and the Ms points determined for alloys containing 2.0 and 3.0 par cent Mo. In this way it has been possible to show the different results of transformation, brought about either by nucleation and diffusion or by shear - the alloy containing 1 per cent Mo, give two martensites {alpha}' and {alpha}'' and the alloys containing 2 and 3 per cent Mo give one martensite with a band structure. (author) [French] Les differentes caracteristiques ont ete determinees sur les alliages 0,5-1-1,8-2 et 3,5 en poids de molybdene, a l'etat brut de coulee: aspect micrographique, structure cristalline, coefficients de dilatation, caracteristiques mecaniques, comportement au cyclage thermique et aux traitements thermiques. Les courbes de transformations au cours de refroidissements continus a des vitesses allant de 2,5 a environ 200 deg. C/mm ont ete etablies a l'aide d'une methode dilatometrique (alliages 1,2 et 3 %). Les courbes TTT ont ete tracees pour les alliages 0,5 et 1 % et les points Ms determines pour les alliages 2 et 3%. Ceci a permis de mettre en evidence differents resultats de transformation, s'operant soit par germination et diffusion, soit par cisaillement (deux martensites: {alpha}' et {alpha}'' pour l'alliage a 1 %, une martensite a structure de bandes pour les alliages 2 et 3%). (auteur)

  3. Fusion neutron irradiation induced ordering and defect production in Cu/sub 3/Au at high temperatures

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-08-01

    We irradiate three Cu/sub 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.

  4. Study on changes of sperm count and testis tissue in black mouse after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ki Jung; Seo, Won Sook [KAERI, Daejeon (Korea, Republic of); Son, Hwa Young [Chungnam National Univ., Daejeon (Korea, Republic of)

    2006-03-15

    For the purpose of the biological effect in black mouse by neutron irradiation, mice were irradiated with 16 or 32 Gy neutron (flux: 1.036739E+09) by lying flat pose at BNCT facility on HANARO Reactors. And 90 days later of irradiation, physical changes of testis and testis tissue were examined. There were no weight changes but a little bit volume changes and sperm counts in the tests. Atrophy of seminiferous tubules irradiated with 32 Gy neutron is increased in number and severity and those in stage VI showed depletion of spermatogonia and pachytene spermatocytes compared to the non-irradiated control group. Testis damage of black mouse was not recovered after long time by 32 Gy neutron irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  6. Grain boundary segregation in neutron-irradiated 304 stainless steel studied by atom probe tomography

    Energy Technology Data Exchange (ETDEWEB)

    Toyama, T., E-mail: ttoyama@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nozawa, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Van Renterghem, W. [SCK Bullet CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Matsukawa, Y.; Hatakeyama, M.; Nagai, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Al Mazouzi, A. [EDF R and D, Avenue des Renardieres Ecuelles, 77818 Moret sur Loing Cedex (France); Van Dyck, S. [SCK Bullet CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium)

    2012-06-15

    Radiation-induced segregation (RIS) of solute atoms at a grain boundary (GB) in 304 stainless steel (SS), neutron-irradiated to a dose of 24 dpa at 300 Degree-Sign C in the fuel wrapper plates of a commercial pressurized water reactor, was investigated using laser-assisted atom probe tomography (APT). Ni, Si, and P enrichment and Cr and Fe depletion at the GB were evident. The full-width at half-maximum of the RIS region was {approx}3 nm for the concentration profile peaks of Ni and Si. The atomic percentages of Ni, Si, and Cr at the GB were {approx}19%, {approx}7%, and {approx}14%, respectively, in agreement with previously-reported values for neutron-irradiated SS. A high number density of intra-granular Ni-Si rich precipitates formed in the matrix. A precipitate-denuded zone with a width of {approx}10 nm appeared on both sides of the GB.

  7. Hydrogen absorption into neutron-irradiated graphite and estimation of the trapping effect

    Energy Technology Data Exchange (ETDEWEB)

    Atsumi, H [Department of Electric and Electronic Engineering, Kinki University, Kowakae 3-4-1, Higashi-Osaka 577-8502 (Japan); Shibata, N [Molecular and Material Engineering, Graduate School of Kinki University, Kowakae 3-4-1, Higashi-Osaka 577-8502 (Japan); Tanabe, T [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Fukuoka 812-8581 (Japan); Shikama, T [Institute for Materials Research, Tohoku University, Sendai, 980-8577 (Japan)

    2007-03-15

    Bulk hydrogen retention and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various conditions. Two kinds of hydrogen trapping sites may exist and be additionally produced during irradiation: interstitial cluster loop edge sites (trap 1) and carbon dangling bonds at edge surfaces of crystallites (trap 2). Neutron irradiation preferably creates trap 2 sites at lower fluences and trap 1 sites at a higher fluence. Trap 2 tends to be annealed out at high temperatures, although trap 1 is hardly decreased even at 1873 K. The activation energy of hydrogen diffusion is found to be increased from 1.04 to 1.60 eV by neutron irradiation.

  8. Influence of rapid thermal process on intrinsic gettering in fast neutron irradiated Czochralski silicon

    Institute of Scientific and Technical Information of China (English)

    CHEN Gui-feng; LI Yang-xian; LI Xing-hua; CAI Li-li; MA Qiao-yun; NIU Ping-juan; NIU Sheng-li; CHEN Dong-feng

    2006-01-01

    A rapid thermal process (RTP) was first introduced into the intrinsic gettering (IG) processes of fast neutron irradiated Czochralski (CZ) silicon. The effect of RTP conditions on bulk microdefects (BMDs) and denuded zone (DZ) was investigated. Fourier transform infrared absorption spectrometer (FTIR) was used to measure the concentration of interstitial oxygen ([Oi]). Bulk microdefects were observed by optical microscope. The results show that,according to the variation of [Oi],it is found that RTP doesn't change the processes of oxygen precipitation in fast neutron irradiated Czochralski silicon. Perfect denuded zone,dense oxygen precipitates and defects form in the bulk of irradiated samples. With increasing temperature of RTP,the width of denuded zone decreases. Increasing RTP cooling rate,the density of Bulk microdefects increases. DZ forms in the sample that annealed in nitrogen atmosphere.

  9. High-resolution photoinduced transient spectroscopy of neutron irradiated bulk silicon

    CERN Document Server

    Kozlowski, R; Nossarzhevska, E

    2002-01-01

    High-resolution photoinduced transient spectroscopy has been employed in a study on the formation of defects in bulk silicon due to 1 MeV neutron irradiation. Apart from divacancies in various charge states, complexes involving interstitial carbon and oxygen were revealed. The defect structure of float zone and Czochralski-grown material exposed to fluences of 2x10 sup 1 sup 4 and 6.75x10 sup 1 sup 4 cm sup - sup 2 is compared.

  10. Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

    Science.gov (United States)

    Miller, M. K.; Chernobaeva, A. A.; Shtrombakh, Y. I.; Russell, K. F.; Nanstad, R. K.; Erak, D. Y.; Zabusov, O. O.

    2009-04-01

    A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10 23 m -2 ( E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10 23 m -2 ( E > 0.5 MeV). High number densities of ˜2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT 41 J ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.

  11. Identification of neutron irradiation induced strain rate sensitivity change using inverse FEM analysis of Charpy test

    Science.gov (United States)

    Haušild, Petr; Materna, Aleš; Kytka, Miloš

    2015-04-01

    A simple methodology how to obtain additional information about the mechanical behaviour of neutron-irradiated WWER 440 reactor pressure vessel steel was developed. Using inverse identification, the instrumented Charpy test data records were compared with the finite element computations in order to estimate the strain rate sensitivity of 15Ch2MFA steel irradiated with different neutron fluences. The results are interpreted in terms of activation volume change.

  12. Analysis of microstress in neutron irradiated polyester fibre by X-ray diffraction technique

    Indian Academy of Sciences (India)

    B Mallick; R C Behera; T Patel

    2005-10-01

    Microstresses developed in the crystallites of polymeric material due to irradiation of high-energy particle causes peak broadening and shifting of X-ray diffraction lines to lower angle. Neutron irradiation significantly changes the material properties by displacement of lattice atoms and the generation of helium and hydrogen by nuclear transmutation. Another important aspect of neutron irradiation is that the fast neutron can produce dense ionization at deep levels in the materials. The polyethylene terephthalate (PET) fibre of raw denier value, 78.2, were irradiated by fast neutron of energy, 4.44 MeV, at different fluences ranging from 1 × 109 n/cm2 to 1 × 1012 n/cm2. In the present work, the radiation heating microstresses developed in PET micro-crystallites was investigated applying X’Pert-MPD Philips Analytical X-ray diffractometer and the effects of microstresses in tensile strength of fibre measured by Instron have also been reported. The shift of 0.45 cm-1 in the Raman peak position of 1614.65 cm-1 to a higher value confirmed the development of microstresses due to neutron irradiation using micro-Raman technique. The defects due to irradiation were observed by SEM micrographs of single fibre for virgin and all irradiated samples.

  13. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  14. The effects of fast neutron irradiation on oxygen in Czochralski silicon

    Institute of Scientific and Technical Information of China (English)

    Chen Gui-Feng; Yan Wen-Bo; Chen Hong-Jian; Li Xing-Hua; Li Yang-Xian

    2009-01-01

    The effects of fast neutron irradiation on oxygen atoms in Czochralski silicon (CZ-Si) are investigated systemically by using Fourier transform infrared (FTIR) spectrometer and positron annihilation technique (PAT). Through isochronal annealing, it is found that the trend of variation in interstitial oxygen concentration ([Oi]) in fast neutrons irradiated CZ-Si fluctuates largely with temperature increasing, especially between 500 and 700℃. After the CZ-Si is annealed at 600℃, the V4 appearing as three-dimensional vacancy clusters causes the formation of the molecule-like oxygen clusters, and more importantly these dimers with small binding energies (0.1-1.0eV) can diffuse into the Si lattices more easily than single oxygen atoms, thereby leading to the strong oxygen agglomerations. When the CZ-Si is annealed at temperature increasing up to 700℃, three-dimensional vacancy clusters disappear and the oxygen agglomerations decompose into single oxygen atoms (O) at interstitial sites. Results from FTIR spectrometer and PAT provide an insight into the nature of the [Oi] at temperatures between 500 and 700℃. It turns out that the large fluctuation of [Oi] after short-time annealing from 500 to 700℃ results from the transformation of fast neutron irradiation defects.

  15. Quantitative TEM analysis of precipitation and grain boundary segregation in neutron irradiated EUROFER 97

    Energy Technology Data Exchange (ETDEWEB)

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-15

    Characterization of irradiation induced microstructural defects is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM) is used to analyze the types and structure of precipitates, and the evolution of their size distributions and densities caused by neutron irradiation to a dose of 32 displacements per atom (dpa) at 330–340 °C in the irradiation experiment ARBOR 1. A significant growth of MX and M{sub 23}C{sub 6} type precipitates is observed after neutron irradiation, while the precipitate density remains unchanged. Hardening caused by MX and M{sub 23}C{sub 6} precipitate growth is assessed by applying the Dispersed Barrier Hardening (DBH) model, and shown to be of minor importance when compared to other irradiation effects like dislocation loop formation. Additionally, grain boundary segregation of chromium induced by neutron irradiation was investigated and detected in irradiated specimens.

  16. The Production of Nickel-Chromium-Molybdenum Alloy with Open Pore Structure as an Implant and the Investigation of Its Biocompatibility In Vivo

    Directory of Open Access Journals (Sweden)

    Yusuf Er

    2013-01-01

    Full Text Available A dental crown material, Nickel-Chrome-Molybdenum alloy, is manufactured using precision casting method from a polyurethane foam model in a regular and open-pore form, as a hard tissue implant for orthopedic applications. The samples produced have 10, 20, and 30 (±3 pores per inch of pore densities and 0.0008, 0.0017, and 0.0027 g/mm3 densities, respectively. Samples were implanted in six dogs and observed for a period of two, four, and six months for the histopathological examinations. The dogs were examined radiologically in 15-day intervals and clinically in certain intervals. The implants were taken out with surrounding tissue at the end of these periods. Implants and surrounding tissues were examined histopathologically in terms of biocompatibility. As a result, it is seen that new bone tissue was formed, in pores of the porous implant at the head of the tibia in dogs implanted. Any pathology, inflammation, and reaction in old and new tissues were not observed. It was concluded that a dental alloy (Ni-Cr-Mo alloy could also be used as a biocompatible hard tissue implant material for orthopedics.

  17. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    2013-01-01

    The present work concerns investigations of damage accumulation during helium implantation of pure iron and the reduced activation ferritic-martensitic steel 'EUROFER 97' at 323K and 623K as well as during neutron irradiation with or without prior helium implantation. The defect microstructure......, in particular the cavities, was characterized using Positron Annihilation Lifetime Spectroscopy (PALS) and Transmission Electron Microscopy (TEM). The PALS investigations reveal a clear difference between the He implantation effects in Fe and EUROFER 97 at both temperatures. For both materials the mean positron...

  18. Effects of helium content of microstructural development in Type 316 stainless steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.

    1985-11-01

    This work investigated the sensitivity of microstructural evolution, particularly precipitate development, to increased helium content during thermal aging and during neutron irradiation. Helium (110 at. ppM) was cold preinjected into solution annealed (SA) DO-heat type 316 stainess steel (316) via cyclotron irradiation. These specimens were then exposed side by side with uninjected samples. Continuous helium generation was increased considerably relative to EBR-II irradiation by irradiation in HFIR. Data were obtained from quantitative analytical electron microscopy (AEM) in thin foils and on extraction replicas. 480 refs., 86 figs., 19 tabs.

  19. Shielding design studies for a neutron irradiator system based on a 252Cf source.

    Science.gov (United States)

    da Silva, A X; Crispim, V R

    2001-01-01

    This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation.

  20. Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Masashi Shimada; M. Hara; T. Otsuka; Y. Oya; Y. Hatano

    2014-05-01

    Accurately estimating tritium retention in plasma facing components (PFCs) and minimizing its uncertainty are key safety issues for licensing future fusion power reactors. D-T fusion reactions produce 14.1 MeV neutrons that activate PFCs and create radiation defects throughout the bulk of the material of these components. Recent studies show that tritium migrates and is trapped in bulk (>> 10 µm) tungsten beyond the detection range of nuclear reaction analysis technique [1-2], and thermal desorption spectroscopy (TDS) technique becomes the only established diagnostic that can reveal hydrogen isotope behavior in in bulk (>> 10 µm) tungsten. Radiation damage and its recovery mechanisms in neutron-irradiated tungsten are still poorly understood, and neutron-irradiation data of tungsten is very limited. In this paper, systematic investigations with repeated plasma exposures and thermal desorption are performed to study defect annealing and thermal desorption of deuterium in low dose neutron-irradiated tungsten. Three tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to high flux (ion flux of (0.5-1.0)x1022 m-2s-1 and ion fluence of 1x1026 m-2) deuterium plasma at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy (TDS) was performed with a ramp rate of 10 °C/min up to 900 °C, and the samples were annealed at 900 °C for 0.5 hour. These procedures were repeated three (for 100 and 200 °C samples) and four (for 500 °C sample) times to uncover damage recovery mechanisms and its effects on deuterium behavior. The results show that deuterium retention decreases approximately 90, 75, and 66 % for 100, 200, and 500 °C, respectively after each annealing. When subjected to the same TDS recipe, the desorption temperature shifts from 800 °C to 600 °C after 1st annealing

  1. Low-temperature properties of neutron irradiated CuGeO3 single crystals

    Science.gov (United States)

    Gladczuk, L.; Mosiniewicz-Szablewska, E.; Dabkowska, H.; Baran, M.; Pytel, B.; Szymczak, R.; Szymczak, H.

    2000-07-01

    The effect of neutron irradiation on the magnetic properties of CuGeO3 single crystal which shows the spin-Peierls transition below T sp=14 K was investigated by means of electron paramagnetic resonance (EPR) and susceptibility measurements. It was found that the irradiation led to a decrease of the spin-Peierls transition temperature and induced appreciable changes in the EPR signal intensity, resonance linewidth, g-factor and magnetic susceptibility of this material. These changes may be associated with a partial suppression of both the energy gap and the dimerization within the Cu chains.

  2. Defect-induced magnetism in neutron irradiated 6H-SiC single crystals.

    Science.gov (United States)

    Liu, Yu; Wang, Gang; Wang, Shunchong; Yang, Jianhui; Chen, Liang; Qin, Xiubo; Song, Bo; Wang, Baoyi; Chen, Xiaolong

    2011-02-25

    Defect-induced magnetism is firstly observed in neutron irradiated SiC single crystals. We demonstrated that the intentionally created defects dominated by divacancies (V(Si)V(C)) are responsible for the observed magnetism. First-principles calculations revealed that defect states favor the formation of local moments and the extended tails of defect wave functions make long-range spin couplings possible. Our results confirm the existence of defect-induced magnetism, implying the possibility of tuning the magnetism of wide band-gap semiconductors by defect engineering.

  3. High temperature nanoindentation hardness and Young's modulus measurement in a neutron-irradiated fuel cladding material

    Science.gov (United States)

    Kese, K.; Olsson, P. A. T.; Alvarez Holston, A.-M.; Broitman, E.

    2017-04-01

    Nanoindentation, in combination with scanning probe microscopy, has been used to measure the hardness and Young's modulus in the hydride and matrix of a high burn-up neutron-irradiated Zircaloy-2 cladding material in the temperature range 25-300 °C. The matrix hardness was found to decrease only slightly with increasing temperature while the hydride hardness was essentially constant within the temperature range. Young's modulus decreased with increasing temperature for both the hydride and the matrix of the high burn-up fuel cladding material. The hydride Young's modulus and hardness were higher than those of the matrix in the temperature range.

  4. Defects in Fast-Neutron Irradiated Nitrogen-Doped Czochralski Silicon after Annealing at High Temperature

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Fast-neutron irradiated nitrogen-doped Czochralski silicon (NCZ-Si) was annealed at 1100 ℃ for different time, then FTIR and optical microscope were used to study the behavior of oxygen. It is found that [Oi] increase at the early stage then decrease along with the increasing of anneal time. High density induced-defects can be found in the cleavage plane. By comparing NCZ-Si with Czochralski silicon (CZ-Si), [Oi] in NCZ-Si decrease more after anneal 24 h.

  5. 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...... 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. Defectmicrostructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations...

  6. Friction stir surfacing of cast A356 aluminium–silicon alloy with boron carbide and molybdenum disulphide powders

    OpenAIRE

    R. Srinivasu; A.Sambasiva Rao; Madhusudhan Reddy, G.; Srinivasa Rao, K.

    2015-01-01

    Good castability and high strength properties of Al–Si alloys are useful in defence applications like torpedoes, manufacture of Missile bodies, and parts of automobile such as engine cylinders and pistons. Poor wear resistance of the alloys is major limitation for their use. Friction stir processing (FSP) is a recognized surfacing technique as it overcomes the problems of fusion route surface modification methods. Keeping in view of the requirement of improving wear resistance of cast alumini...

  7. Evaluation of gamma and neutron irradiation effects on the properties of mica film capacitors

    Indian Academy of Sciences (India)

    Rajesh Roy; Arun Pandya

    2005-12-01

    We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 m (0.7874 and 1.5748 mil) with two different areas, 01 and 04 cm2. The capacitance has been measured at room temperature in the frequency range 100 Hz–10 MHz. Negligible change in the capacitance due to high gamma dose of 60Co, 15 kGy at dose rate 0.25 kGy/h, has been observed. However, appreciable change in the capacitance has been observed due to low doses of fast neutrons (cumulative dose, 115 cGy) with flux ∼ 9.925 × 107 neutrons/cm2 h from 252Cf neutron source of fluence, 2.5 × 107 neutrons/s. We have also observed that the impact of gamma and neutron irradiation is more at frequencies higher than 10 kHz. These results show that the mica capacitors do not show any radiation response below 10 kHz. The study shows the radiation response of mica film capacitors to gamma and fast neutron radiations. Mica capacitors show low gamma radiation response in comparison to fast neutron radiation, because a total dose of kGy order has been given by gamma source and only few cGy dose has been given by fast neutron source.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alsabbagh, Ahmad, E-mail: ahalsabb@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Sarkar, Apu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Miller, Brandon [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Burns, Jatuporn [Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Squires, Leah; Porter, Douglas; Cole, James I. [ATR National Scientific User Facility, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Murty, K.L. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2014-10-06

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) have been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.37 dpa. Atom probe tomography revealed manganese and silicon-enriched clusters in both UFG and CG steel after neutron irradiation. Mechanical properties were characterized using microhardness and tensile tests, and irradiation of UFG carbon steel revealed minute radiation effects in contrast to the distinct radiation hardening and reduction of ductility in its CG counterpart. After irradiation, micro hardness indicated increases of around 9% for UFG versus 62% for CG steel. Similarly, tensile strength revealed increases of 8% and 94% respectively for UFG and CG steels while corresponding decreases in ductility were 56% versus 82%. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation while no significant change was observed in UFG steel, revealing better radiation tolerance. Quantitative correlations between experimental results and modeling were demonstrated based on irradiation induced precipitate strengthening and dislocation forest hardening mechanisms.

  9. Investigations of void formation in neutron irradiated iron and F82H steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    2002-01-01

    In the present work pure iron and low activation steel F82H have been neutron irradiated at temperatures in the interval 50 deg.C - 350 deg.C to a dose of 0.23 dpa (displacements per atom). The formation of defects has been investigated by the use ofpositron annihilation spectroscopy (PAS......). In addition iron has been irradiated to different doses in the range 0.01 - 0.4 dpa at 50oC and 100oC and the dose dependence of the electrical conductivity determined. The results demonstrated that theformation of voids takes place during neutron irradiation of pure iron in the whole temperature range....... For irradiation temperatures of 50 deg.C and 100 deg.C also a high density of micro-voids was observed. Voids and micro-voids were also detected in lowactivation F82H steel for a low irradiation temperature (50 deg.C), while for irradiation close to the temperature of annealing stage V (250 deg.C), no voids...

  10. Evaluation of thermal neutron irradiation field using a cyclotron-based neutron source for alpha autoradiography.

    Science.gov (United States)

    Tanaka, H; Sakurai, Y; Suzuki, M; Masunaga, S; Mitsumoto, T; Kinashi, Y; Kondo, N; Narabayashi, M; Nakagawa, Y; Watanabe, T; Fujimoto, N; Maruhashi, A; Ono, K

    2014-06-01

    It is important to measure the microdistribution of (10)B in a cell to predict the cell-killing effect of new boron compounds in the field of boron neutron capture therapy. Alpha autoradiography has generally been used to detect the microdistribution of (10)B in a cell. Although it has been performed using a reactor-based neutron source, the realization of an accelerator-based thermal neutron irradiation field is anticipated because of its easy installation at any location and stable operation. Therefore, we propose a method using a cyclotron-based epithermal neutron source in combination with a water phantom to produce a thermal neutron irradiation field for alpha autoradiography. This system can supply a uniform thermal neutron field with an intensity of 1.7×10(9) (cm(-2)s(-1)) and an area of 40mm in diameter. In this paper, we give an overview of our proposed system and describe a demonstration test using a mouse liver sample injected with 500mg/kg of boronophenyl-alanine.

  11. An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Motohashi, Yoshinobu [Ibaraki Univ., Mito (Japan)

    2000-05-01

    A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10{sup 20} n/cm{sup 2} and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10{sup 10} Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10{sup 20} n/cm{sup 2} and that it decays to about 1/100 in a year. (author)

  12. Neutron irradiation and damage assessment of plastic scintillators of the Tile Calorimeter

    Science.gov (United States)

    Mdhluli, J. E.; Mellado, B.; Sideras-Haddad, E.

    2017-01-01

    Following the comparative study of proton induced radiation damage on various plastic scintillator samples from the ATLAS-CERN detector, a study on neutron irradiation and damage assessment on the same type of samples will be conducted. The samples will be irradiated with different dose rates of neutrons produced in favourable nuclear reactions using a radiofrequency linear particle accelerator as well as from the SAFARI nuclear reactor at NECSA. The MCNP 5 code will be utilized in simulating the neutron transport for determining the dose rate. Light transmission and light yield tests will be performed in order to assess the radiation damage on the scintillators. In addition, Raman spectroscopy and Electron Paramagnetic Resonance (EPR) analysis will be used to characterize the samples after irradiation. The project aims to extent these studies to include radiation assessment damage of any component that processes the scintillating light and deteriorates the quantum efficiency of the Tilecal detector, namely, photomultiplier tubes, wavelength shifting optical fibres and the readout electronics. They will also be exposed to neutron irradiation and the damage assessed in the same manner.

  13. Point defects in 4H–SiC epilayers introduced by neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hazdra, Pavel, E-mail: hazdra@fel.cvut.cz [Department of Microelectronics, Faculty of Electrical Engineering, Czech Technical University in Prague, Technická 2, CZ-16627 Prague 6 (Czech Republic); Záhlava, Vít [Department of Microelectronics, Faculty of Electrical Engineering, Czech Technical University in Prague, Technická 2, CZ-16627 Prague 6 (Czech Republic); Vobecký, Jan [Department of Microelectronics, Faculty of Electrical Engineering, Czech Technical University in Prague, Technická 2, CZ-16627 Prague 6 (Czech Republic); ABB Switzerland Ltd., Semiconductors, Fabrikstrasse 3, CH-5600 Lenzburg (Switzerland)

    2014-05-01

    Electronic properties of radiation damage produced in 4H–SiC by neutron irradiation and its effect on electrical parameters of Junction Barrier Schottky (JBS) diodes were investigated. 4H–SiC N-epilayers, which formed the low-doped N-base of JBS power diodes, were irradiated with 1 MeV neutrons with fluences ranging from 1.3 × 10{sup 13} to 4.0 × 10{sup 14} cm{sup −2}. Radiation defects were then characterized by capacitance deep-level transient spectroscopy, I–V and C–V measurement. Results show that neutron irradiation introduces different point defects giving rise to acceptor levels lying 0.61/0.69, 0.88, 1.03, 1.08 and 1.55 eV below the SiC conduction band edge. Introduction rates of these centers are ranging from 0.64 to 4.0 cm{sup −1}. These defects have a negligible effect on blocking and dynamic characteristics of irradiated diodes. However, the acceptor character of introduced deep levels and their fast introduction deteriorate diode’s ON-state resistance already at fluences exceeding 1 × 10{sup 14} cm{sup −2}.

  14. Cluster dynamics modeling of accumulation and diffusion of helium in neutron irradiated tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y.G.; Zhou, W.H.; Huang, L.F. [Key Laboratory for Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zeng, Z., E-mail: zzeng@theory.issp.ac.cn [Key Laboratory for Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Ju, X. [Department of Physics, University of Science and Technology Beijing, Beijing 100083 (China)

    2012-12-15

    A cluster dynamics model based on rate theory has been developed to study the accumulation and diffusion processes of helium in tungsten under synergistic effects of helium implantation and neutron irradiation. By including self-interstitial atoms, vacancies and helium atoms as well as their clusters and further using more reliable parameters, the evolution of different types of defects with time and depth is investigated. The calculated results are comparable with experiments. The cases of helium plasma corresponding to the first wall and to the divertor are taken into account. The accumulation and diffusion behaviors of helium in tungsten are illustrated by the time and depth dependence of helium concentration in tungsten with or without the neutron irradiation, the contribution of different types of helium clusters/complexes to helium concentration and the depth profiles of different mobile defects and helium-vacancy complexes. It is concluded that the competition of trapping and diffusion effects dominates the behavior of helium atoms in tungsten for these two typical cases. Understanding these mechanisms is important for estimating damages to the plasma-facing materials.

  15. Neutron diffraction analysis of Cr-Ni-Mo-Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    Science.gov (United States)

    Voronin, V. I.; Valiev, E. Z.; Berger, I. F.; Goschitskii, B. N.; Proskurnina, N. V.; Sagaradze, V. V.; Kataeva, N. F.

    2015-04-01

    A quantitative assessment is presented of the dislocation density and relative fractions of edge and screw dislocations in reactor-steel samples 16Cr-15Ni-3Mo-1Ti subjected to preliminary cold deformation by rolling and subsequent fast neutron irradiation using neutron diffraction analysis. The Williamson-Hall modified method was used for calculations. It is shown that the fast neutron irradiation leads to a decrease in the density of dislocations that appeared after samples deformation. The applicability of neutron diffraction analysis to the examination of dislocation structure of deformed and irradiated materials is shown.

  16. Modification of the mesoscopic structure in neutron irradiated EPDM viewed through positron annihilation spectroscopy and dynamic mechanical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lambri, O.A., E-mail: olambri@fceia.unr.edu.a [Instituto de Fisica Rosario - CONICET, Avda. 27 de Febrero 210 bis, 2000 Rosario (Argentina); Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, 2000 Rosario (Argentina); Plazaola, F.; Axpe, E. [Elektrizitatea eta Elektronika Saila, Zientzia eta Teknologia Fakultatea, Euskal Herriko Unibertsitatea, P.K. 644, 48080 Bilbao (Spain); Mocellini, R.R.; Zelada-Lambri, G.I. [Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avda. Pellegrini 250, 2000 Rosario (Argentina); Garcia, J.A. [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apdo. 644, 48080 Bilbao, Pais Vasco (Spain); Matteo, C.L.; Sorichetti, P.A. [Departamento de Fisica, Facultad de Ingenieria, Universidad de Buenos Aires, Avda. Paseo Colon 850, 1063 Buenos Aires (Argentina)

    2011-02-01

    This article focuses on the study of the mesoscopic structure in neutron irradiated EPDM both from experimental and theoretical points of view. In this work we reveal completely the modification of the mesostructure of the EPDM due to neutron irradiation, resolving volume fraction, size and distribution of the crystalline zones as a function of the irradiation dose. Positron annihilation spectroscopy and dynamic mechanical analysis techniques are applied and the results are discussed by means of new theoretical results for describing the interaction process between the crystals and amorphous zones in EPDM.

  17. In-situ, Gate Bias Dependent Study of Neutron Irradiation Effects on AlGaN/GaN HFETs

    Science.gov (United States)

    2010-03-01

    Applied Physics Letters , vol.82, no. 22, 2 June 2008. 72 [12] D. M. Sathaiya, et al., "Thermionic trap-assisted tunneling model and its... Letters , vol. 25, no. 3, 1045, 2008. [18] A. Y. Polyakov , et al., “Neutron irradiation effects on electrical properties and deep-level spectra in...undoped n-AlGaN/GaN heterostructures,” Journal of Applied Physics , vol. 98, 033529, 2005. [19] A. Y. Polyakov , et al., “Neutron irradiation effects in

  18. 微合金钢中钛钼复合析出的第一性原理研究%Study on First Principle of Complex Precipitation of Titanium and Molybdenum in Micro-alloyed Steel

    Institute of Scientific and Technical Information of China (English)

    赵冬伟; 曹建春; 周晓龙; 彭谦之

    2012-01-01

    The formation energy of complex carbonitrides of titanium and molybdenum, which may precipitate in low alloyed steel, was computed using materials studiq-CASTEP module. The calculated results indicate that the effect of molybdenum on the formation energy of (Ti1-xMox)C presents a parasol trend. The formation energy of (Ti1-x Mox)C first decreases and then increases with increasing molybdenum content, and the maximum appears when Ti/Mo is 1. The formation energy of titanium carbide is lower than that of molybdenum carbide. Molybdenum can reduce the formation energy of (Ti1-x Mox)(CyN1-y) and the impact becomes sluggish as molybdenum content increasing. The formation energy of nitrogen-rich carbonitrides is lower than that of carbon-rich carbonitrides.%利用Materials Studio中CASTEP模块对钢中出现的钛钼复合析出相的形成能进行计算.结果表明,Mo含量对碳化钛钼形成能的影响趋势呈抛物线状,以Ti/Mo值等于1为界,随着Mo含量的增加,碳化钛钼的形成能先降低后升高;TiC的形成能比MoC的低,Mo会降低碳氮化钛钼的形成能,且降低幅度随钼含量增加逐渐减小;富氮的碳氮化钛钼比富碳的碳氮化钛钼具有更低的形成能.

  19. Inverse magnetocaloric effect in Ce(Fe0.96Ru0.04)2: Effect of fast neutron irradiation

    Science.gov (United States)

    Dube, V.; Mishra, P. K.; Rajarajan, A. K.; Prajapat, C. L.; Sastry, P. U.; Thakare, S. V.; Singh, M. R.; Ravikumar, G.

    2013-02-01

    We have shown the effect of fast neutron irradiation on the magnetic phase transition and magnetocaloric effect (MCE) in a doped Ce(Fe0.96Ru0.04)2, intermettalic. We show that this leads to suppression of MCE and a to a disordered ferromagnetic phase.

  20. Tritium Retention and Permeation in Ion- and Neutron-Irradiated Tungsten under US-Japan PHENIX Collaboration

    Science.gov (United States)

    Shimada, Masashi; Taylor, Chase N.; Kolasinski, Robert D.; Buchenauer, Dean A.; Chikada, Takumi; Oya, Yasuhisa; Hatano, Yuji

    2015-11-01

    A critical challenge for long-term operation of ITER and beyond to a FNSF, a DEMO and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to intense heat and neutral/ion particle fluxes under the extreme fusion nuclear environment, while minimizing in-vessel inventories and ex-vessel permeation of tritium. Recent work at Tritium Plasma Experiment demonstrated that tritium diffuses in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. US-Japan PHENIX collaboration (2013-2019) investigates irradiation response on tritium behavior in tungsten, and performs one-of-a-kind neutron-irradiation with Gd thermal neutron shield at High Flux Isotope Reactor, ORNL. This presentation describes the challenge in elucidating tritium behavior in neutron-irradiated PFCs, the PHENIX plans for neutron-irradiation and post irradiation examination, and the recent findings on tritium retention and permeation in 14MeV neutron-irradiated and Fe ion irradiated tungsten. This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.

  1. Antiradiation Vaccine: Technology Development Of Prophylaxis, Prevention And Treatment Of Biological Consequences And Complications After Neutron Irradiation.

    Science.gov (United States)

    Popov, Dmitri; Maliev, Slava; Jones, Jeffrey

    Introduction: Neutrons irradiation produce a unique biological effectiveness compare to different types of radiation because their ability to create a denser trail of ionized atoms in biological living tissues[Straume 1982; Latif et al.2010; Katz 1978; Bogatyrev 1982]. The efficacy of an Anti-Radiation Vaccine for the prophylaxis, prevention and therapy of acute radiation pathology was studied in a neutron exposure facility. The biological effects of fast neutrons include damage of central nervous system and cardiovascular system with development of Acute Cerebrovascular and Cardiovascular forms of acute radiation pathology. After irradiation by high doses of fast neutron, formation of neurotoxins, hematotoxins,cytotoxins forming from cell's or tissue structures. High doses of Neutron Irradiation generate general and specific toxicity, inflammation reactions. Current Acute Medical Management and Methods of Radiation Protection are not effective against moderate and high doses of neutron irradiation. Our experiments demonstrate that Antiradiation Vaccine is the most effective radioprotectant against high doses of neutron-radiation. Radiation Toxins(biological substances with radio-mimetic properties) isolated from central lymph of gamma-irradiated animals could be working substance with specific antigenic properties for vaccination against neutron irradiation. Methods: Antiradiation Vaccine preparation standard - mixture of a toxoid form of Radiation Toxins - include Cerebrovascular RT Neurotoxin, Cardiovascular RT Neurotoxin, Gastrointestinal RT Neurotoxin, Hematopoietic RT Hematotoxin. Radiation Toxins were isolated from the central lymph of gamma-irradiated animals with different forms of Acute Radiation Syndromes - Cerebrovascular, Cardiovascular, Gastrointestinal, Hematopoietic forms. Devices for Y-radiation were "Panorama","Puma". Neutron exposure was accomplished at the Department of Research Institute of Nuclear Physics, Dubna, Russia. The neutrons

  2. Monte Carlo Calculations for Neutron and Gamma Radiation Fields on a Fast Neutron Irradiation Device

    Science.gov (United States)

    Vieira, A.; Ramalho, A.; Gonçalves, I. C.; Fernandes, A.; Barradas, N.; Marques, J. G.; Prata, J.; Chaussy, Ch.

    We used the Monte Carlo program MCNP to calculate the neutron and gamma fluxes on a fast neutron irradiation facility being installed on the Portuguese Research Reactor (RPI). The purpose of this facility is to provide a fast neutron beam for irradiation of electronic circuits. The gamma dose should be minimized. This is achieved by placing a lead shield preceded by a thin layer of boral. A fast neutron flux of the order of 109 n/cm2s is expected at the exit of the tube, while the gamma radiation is kept below 20 Gy/h. We will present results of the neutron and gamma doses for several locations along the tube and different thickness of the lead shield. We found that the neutron beam is very collimated at the end of the tube with a dominant component on the fast region.

  3. The effect of neutron irradiation on the mechanical properties of C/SiC composites

    Energy Technology Data Exchange (ETDEWEB)

    Shih, Chunghao [ORNL; Katoh, Yutai [ORNL; Snead, Lance Lewis [ORNL; Steinbeck, John [ORNL

    2013-01-01

    The effects of neutron irradiation to 3.5 and 9.5 dpa at 730 C on a 2D plain woven carbon fiber reinforced polymer derived SiC matrix composite are presented. For both fluences, the irradiation caused in-plane contraction and trans-plane expansion. Irradiation also caused substantial reduction in composite flexural strength (54%) and increase in flexural tangent modulus (+85%). The extents of dimensional/ mechanical property changes were greater for the higher fluence irradiated samples. Those changes suggest the instability of the polymer derived SiC matrix following irradiation. The nature of the mechanical property changes suggest increased clamping stress between the fiber and the matrix. The composite property changes are explained in terms of irradiation effects on composite constituents and are compared with carbon fiber reinforced carbon matrix composite as a reference material.

  4. The effect of neutron irradiation on the mechanical properties of C/SiC composites

    Energy Technology Data Exchange (ETDEWEB)

    Shih, Chunghao, E-mail: shihc@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory (United States); Katoh, Yutai, E-mail: katohy@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory (United States); Snead, Lance L., E-mail: sneadll@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory (United States); Steinbeck, John, E-mail: jws@psicorp.com [Physical Science Inc., Andover MA (United States)

    2013-08-15

    The effects of neutron irradiation to 3.5 and 9.5 dpa at 730 °C on a 2D plain woven carbon fiber reinforced polymer derived SiC matrix composite are presented. For both fluences, the irradiation caused in-plane contraction and trans-plane expansion. Irradiation also caused substantial reduction in composite flexural strength (−54%) and increase in flexural tangent modulus (+85%). The extents of dimensional/mechanical property changes were greater for the higher fluence irradiated samples. Those changes suggest the instability of the polymer derived SiC matrix following irradiation. The nature of the mechanical property changes suggest increased clamping stress between the fiber and the matrix. The composite property changes are explained in terms of irradiation effects on composite constituents and are compared with carbon fiber reinforced carbon matrix composite as a reference material.

  5. The effect of neutron irradiation on the mechanical properties of C/SiC composites

    Science.gov (United States)

    Shih, Chunghao; Katoh, Yutai; Snead, Lance L.; Steinbeck, John

    2013-08-01

    The effects of neutron irradiation to 3.5 and 9.5 dpa at 730 °C on a 2D plain woven carbon fiber reinforced polymer derived SiC matrix composite are presented. For both fluences, the irradiation caused in-plane contraction and trans-plane expansion. Irradiation also caused substantial reduction in composite flexural strength (-54%) and increase in flexural tangent modulus (+85%). The extents of dimensional/mechanical property changes were greater for the higher fluence irradiated samples. Those changes suggest the instability of the polymer derived SiC matrix following irradiation. The nature of the mechanical property changes suggest increased clamping stress between the fiber and the matrix. The composite property changes are explained in terms of irradiation effects on composite constituents and are compared with carbon fiber reinforced carbon matrix composite as a reference material.

  6. Temperature dependence of the deformation behavior of 316 stainless steel after low temperature neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Pawel-Robertson, J.E.; Rowcliffe, A.F.; Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)] [and others

    1996-10-01

    The effects of low temperature neutron irradiation on the tensile behavior of 316 stainless steel have been investigated. A single heat of solution annealed 316 was irradiated to 7 and 18 dpa at 60, 200, 330, and 400{degrees}C. The tensile properties as a function of dose and as a function of temperature were examined. Large changes in yield strength, deformation mode, strain to necking, and strain hardening capacity were seen in this irradiation experiment. The magnitudes of the changes are dependent on both irradiation temperature and neutron dose. Irradiation can more than triple the yield strength over the unirradiated value and decrease the strain to necking (STN) to less than 0.5% under certain conditions. A maximum increase in yield strength and a minimum in the STN occur after irradiation at 330{degrees}C but the failure mode remains ductile.

  7. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lamadrid B, A.I.; Garcia L, O. [CPHR, Calle 20 No. 4113 e/41 y 47, Playa, La Habana 11300 (Cuba); Delbos, M.; Voisin, P.; Roy, L. [Institut de Radioprotection et de Surete Nucleaire, BP 17, 92262 Fontenay-aux-Roses (France)]. e-mail: ana@cphr.edu.cu

    2006-07-01

    Dose-effect curves for dose assessment in Gamma and neutron overexposures to high doses are presented in this paper for the first time in literature. The relationships were obtained by plotting the Premature Chromosome Condensation -rings (PCC-R) frequencies in PCC Iymphocytes obtained by chemical induction with Calyculin A in vitro, with radiation doses between 5 to 25 Gy. For the elaboration of these curves 9 676 PCC cells in Gl G2 and M stages were analyzed. The results were fitted to a lineal quadratic model in Gamma irradiation. For neutron irradiation the data was fitted to a lineal quadratic model up to 10 Gy and then a markedly cell cycle arrest and saturation was observed. These curves are of particular interest for victims exposed to doses exceeding 5 Gy where it is always very difficult to estimate a dose using the conventional technique. (Author)

  8. Polarised SANS study of microstructural evolution under neutron irradiation in a martensitic steel for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coppola, R.; Dewhurst, C.D.; Lindau, R.; May, R.P.; Moeslang, A.; Valli, M

    2004-03-01

    This work presents the results of polarised small-angle neutron scattering (SANS) measurements of modified martensitic steel DIN1.4914, originally developed for application in future fusion reactors (MANET steel). SANS measurements were made using the D22 instrument at the ILL Grenoble using an ad hoc polarised beam set-up. The investigated MANET samples were neutron irradiated and subsequently post-irradiation tempered to reproduce as much as possible the expected service conditions. The results, based on the analysis of the nuclear-magnetic interference, are discussed taking into account both the occurrence of Cr redistribution phenomena with correlated changes in the composition of the precipitate phases, and the growth of non-magnetic defects (He-bubbles or microvoids)

  9. Glass-like, low-energy excitations in neutron-irradiated quartz

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, John William [Univ. of Illinois, Urbana-Champaign, IL (United States). Dept. of Physics

    1980-01-01

    The specific heat and thermal conductivity of neutron-irradiated crystalline quartz have been measured for temperatures ≈ 0.1 to 5 K. Four types of low-energy excitations are observed in the irradiated samples, two of which can be removed selectively by heat treatment. One set of remaining excitations gives rise to low-temperature thermal behavior characteristic of glassy (amorphous) solids. The density of these glass-like excitations can be 50% the density observed in vitreous silica, yet the sample still retains long-range atomic order. In a less-irradiated sample, glass-like excitations may be present with a density only ≈ 2.5% that observed in vitreous silica and possess a similar broad energy spectrum over 0.1 to 1 K.

  10. Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Saleh, Tarik A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Eftink, Benjamin P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, the synergies between ' and fine-scale and moderate-scale cavity formation is investigated.

  11. Structure and Spatial Distribution of Ge Nanocrystals Subjected to Fast Neutron Irradiation

    Directory of Open Access Journals (Sweden)

    Alexander N. Ionov

    2011-07-01

    Full Text Available The influence of fast neutron irradiation on the structure and spatial distribution of Ge nanocrystals (NC embedded in an amorphous SiO2 matrix has been studied. The investigation was conducted by means of laser Raman Scattering (RS, High Resolution Transmission Electron Microscopy (HR-TEM and X-ray photoelectron spectroscopy (XPS. The irradiation of Ge- NC samples by a high dose of fast neutrons lead to a partial destruction of the nanocrystals. Full reconstruction of crystallinity was achieved after annealing the radiation damage at 8000C, which resulted in full restoration of the RS spectrum. HR-TEM images show, however, that the spatial distributions of Ge-NC changed as a result of irradiation and annealing. A sharp decrease in NC distribution towards the SiO2 surface has been observed. This was accompanied by XPS detection of Ge oxides and elemental Ge within both the surface and subsurface region.

  12. Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

    Science.gov (United States)

    Koyanagi, Takaaki; Kumar, N. A. P. Kiran; Hwang, Taehyun; Garrison, Lauren M.; Hu, Xunxiang; Snead, Lance L.; Katoh, Yutai

    2017-07-01

    Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90-800 °C to 0.03-4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

  13. ATLAS MDT chamber behaviour after neutron irradiation and in a high rate background

    Energy Technology Data Exchange (ETDEWEB)

    Branchini, Paolo; Di Luise, Silvestro; Graziani, Enrico [Dipartimento di Fisica, Universita di Rome Tre and INFN Sezione di Roma Tre, Rome (Italy); Mazzotta, Concetta; Meoni, Evelin; Morello, Gianfranco [Dipartimento di Fisica, Universita della Calabria and INFN Gruppo Collegato di Cosenza, Cosenza (Italy); Passeri, Antonio; Petrucci, Fabrizio [Dipartimento di Fisica, Universita di Rome Tre and INFN Sezione di Roma Tre, Rome (Italy); Policicchio, Antonio [Dipartimento di Fisica, Universita della Calabria and INFN Gruppo Collegato di Cosenza, Cosenza (Italy)], E-mail: antonio.policicchio@cern.ch; Salvatore, Daniela; Schioppa, Marco [Dipartimento di Fisica, Universita della Calabria and INFN Gruppo Collegato di Cosenza, Cosenza (Italy)

    2007-10-21

    Many of the physics processes of interest at the Large Hadron Collider (LHC) will involve muon production in the final state. The Monitored Drift Tube (MDT) chambers, the precision tracking elements of the ATLAS muon spectrometer, are the main tools for the muon identification and measurement. They will operate in the harsh LHC background environment, mainly due to low energy photons and neutrons which will dominate the counting rate in most areas of the spectrometer, where an overall maximum counting rate of 500Hz/cm{sup 2} is expected. The upgrade to Super-LHC will involve fluxes ten times higher. To study the behaviour of MDT chambers under massive neutron irradiation at the level of Super-LHC, a test was performed at the 'Tapiro' Neutron Facility of the ENEA 'La Casaccia' Research Center.

  14. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-15

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  15. Thermal stability of neutron irradiation effects on KU1 fused silica

    Energy Technology Data Exchange (ETDEWEB)

    Leon, M. [Materiales Para Fusion, CIEMAT, Avenida Complutense 22, Madrid (Spain)], E-mail: monica.leon@ciemat.es; Martin, P. [Materiales Para Fusion, CIEMAT, Avenida Complutense 22, Madrid (Spain); Bravo, D.; Lopez, F.J. [Departamento Fisica de Materiales, Universidad Autonoma, Cantoblanco, Madrid (Spain); Ibarra, A. [Materiales Para Fusion, CIEMAT, Avenida Complutense 22, Madrid (Spain); Rascon, A. [Metrologia Radiaciones Ionizantes, CIEMAT, Avenida Complutense 22, Madrid (Spain); Mota, F. [Instituto de Fusion Nuclear, DENIM, Universidad Politecnica, Madrid (Spain)

    2008-03-15

    Optical absorption spectra of neutron irradiated (10{sup 21} n/m{sup 2} and 10{sup 22} n/m{sup 2}) KU1 quartz glass samples have been measured. The effects of post-irradiation isochronal thermal annealing, up to 850 deg. C, have been investigated. The general effect of the isochronal annealing is a decrease in the optical absorption bands as the temperature increases. Optical absorption bands have been identified with known defects from the literature, and their concentration temperature dependence has been analyzed. While the annealing curves of the E' and non-bridging oxygen hole centres (NBOHC) are similar, that corresponding to oxygen deficiency centres (ODC(II)) is quite different suggesting that the recombination of E' and NBOHC is part of the same process whereas the recombination of ODC is controlled by the presence of another undetected defect.

  16. Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

    Science.gov (United States)

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Katoh, Yutai; Snead, Lance L.; Wirth, Brian D.

    2016-03-01

    The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (∼90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S-W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage, providing insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

  17. Influência do teor de Mo na microestrutura de ligas Fe-9Cr-xMo Effect of the content of molybdenum in the microstructure of Fe-9Cr-xMo alloy

    Directory of Open Access Journals (Sweden)

    Rodrigo Freitas Guimarães

    2010-12-01

    Full Text Available Aços Cr-Mo são usados na indústria do petróleo em aplicações com óleos crus ricos em compostos sulfurosos. Aços comerciais como 2.5Cr1Mo ou 9Cr1Mo têm se mostrado ineficientes em consequência de altos índices de corrosão naftênica. Uma estratégia para resolver este problema é o aumento do teor de molibdênio destes aços. Neste trabalho foi estudado o efeito do aumento do teor de molibdênio na microestrutura de ligas Fe-9Cr-xMo, solubilizadas e soldadas. Foram levantados os diagramas de fases com auxílio de um programa comercial para verificar as possíveis fases a serem formadas e identificar os problemas de soldagem. A microestrutura das ligas solubilizadas foi analisada por microscopia óptica e EBSD, além da medição da dureza. Foram realizadas soldagens autógenas para verificar o efeito do aporte térmico na microestrutura e na dureza das ligas. O aumento do teor de molibdênio resultou no aumento da dureza das ligas. A análise microestrutural das ligas soldadas apresentou uma particularidade para a liga com menor teor de molibdênio, a presença de martensita. Já as ligas com maior teor de molibdênio apresentaram uma microestrutura completamente ferrítica. A formação de martensita pode ser um problema na solda da liga com menor teor de molibdênio, uma vez que a mesma pode causar perdas nas propriedades mecânicas comprometendo sua aplicação.Cr-Mo steels are used in the petroleum industry in applications with crude oils rich in sulfur compounds. 2.5Cr1Mo or 9Cr1Mo do not resist to operating conditions when in contact with crude oils. The increasing of molybdenum content can improve the corrosion resistance of these alloys. This paper studied the effect of increased concentration of molybdenum in the microstructure of Fe-9Cr-xMo alloys, annealed and welded. Phase diagrams were built with the aid of commercial program to check the possible phases to be formed and to identify the problems of welding. Analyses were

  18. The Chemically-Specific Structure of an Amorphous Molybdenum Germanium Alloy by Anomalous X-ray Scattering

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, H. A.

    2002-06-11

    Since its inception in the late 1970s, anomalous x-ray scattering (AXS) has been employed for chemically-specific structure determination in a wide variety of noncrystalline materials. These studies have successfully produced differential distribution functions (DDFs) which provide information about the compositionally-averaged environment of a specific atomic species in the sample. Despite the wide success in obtaining DDFs, there are very few examples of successful extraction of the fully-chemically-specific partial pair distribution functions (PPDFs), the most detailed description of an amorphous sample possible by x-ray scattering. Extracting the PPDFs is notoriously difficult since the matrix equation involved is ill-conditioned and thus extremely sensitive to errors present in the experimental quantities that enter the equation. Instead of addressing this sensitivity by modifying the data through mathematical methods, sources of error have been removed experimentally: A focusing analyzer crystal was combined with a position-sensitive linear detector to experimentally eliminate unwanted inelastic scattering intensity over most of the reciprocal space range probed. This instrumentation has been used in data collection for the extraction of PPDFs from amorphous (a)-MoGe{sub 3}. This composition arises as a phase separation endpoint in the Ge-rich region of the vapor-deposited Mo-Ge amorphous alloy system but is not present at equilibrium. Since the first Ge-rich compound in the Mo-Ge equilibrium system is MoGe{sub 2}, previous workers have speculated that perhaps a unique MoGe{sub 3} compound exists in the amorphous system. Rather than indicating a distinct MoGe{sub 3} compound with definitive local structure, however, the coordination results are more consistent with a densely-packed alloy having a wide range of solid solubility. Significant improvement in the quality and reliability of experimental PPDFs from a-MoGe{sub 3} by AXS has been achieved solely

  19. Study on preparation and properties of molybdenum alloys reinforced by nano-sized ZrO{sub 2} particles

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Chaopeng; Gao, Yimin; Zhou, Yucheng [Xi' an Jiaotong University, State Key Laboratory for Mechanical Behavior of Materials, Xi' an, Shaanxi Province (China); Wei, Shizhong [Henan University of Science and Technology, School of Materials Science and Engineering, Luoyang (China); Henan University of Science and Technology, Engineering Research Center of Tribology and Materials Protection, Ministry of Education, Luoyang (China); Zhang, Guoshang; Zhu, Xiangwei; Guo, Songliang [Henan University of Science and Technology, School of Materials Science and Engineering, Luoyang (China)

    2016-03-15

    The nano-sized ZrO{sub 2}-reinforced Mo alloy was prepared by a hydrothermal method and a subsequent powder metallurgy process. During the hydrothermal process, the nano-sized ZrO{sub 2} particles were added into the Mo powder via the hydrothermal synthesis. The grain size of Mo powder decreases obviously with the addition of ZrO{sub 2} particles, and the fine-grain sintered structure is obtained correspondingly due to hereditation. In addition to a few of nano-sized ZrO{sub 2} particles in grain boundaries or sub-boundaries, most are dispersed in grains. The tensile strength and yield strength have been increased by 32.33 and 53.76 %. (orig.)

  20. Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, Juergen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Aaron, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bell, Gary L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burgess, Thomas W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lessard, Timothy L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ohriner, Evan Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Perkins, Dale E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Varma, Venugopal Koikal [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-20

    Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady

  1. Post-irradiation experiments on physical thermal and microstructural properties of neutron-irradiated ceramics. 2

    Energy Technology Data Exchange (ETDEWEB)

    Yano, Toyohiko [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1999-03-01

    Succeeding to the report on the post-irradiation experiments conducted in the previous year, this is a summary report on the post-irradiation experiments of physical, thermal and microstructural properties of neutron-irradiated various ceramics, which are expected to be applied to the in-core materials of an Advanced Fast Breeder Reactor in near future. Four candidate ceramics, Al{sub 2}O{sub 3}, AlN, SiC and Si{sub 3}N{sub 4} were fast-neutron-irradiated up to a fluence of 3.9x10{sup 26} n/m{sup 2}, different irradiation conditions from the previous report specimens, in the CMIR-4 rig in the JOYO experimental fast reactor in JNC. The following observations were performed: (1) Microstructural observation by means of transmission electron microscopy, (2) Measurement of swelling, (3) Measurement of thermal diffusivity by a laser-flash method, (4) Recovery of swelling by isochronal annealing, and (5) Recovery of thermal diffusivity by isochronal annealing. Obtained main results are summarized as follows. Macroscopic length changes by neutron irradiation of Al{sub 2}O{sub 3} and AlN were measured to be 1.8-2.0% and these of SiC and Si{sub 3}N{sub 4} to be 0.2-0.4%, respectively. Thermal diffusivities of all irradiated materials degraded to 0.03-0.05 cm{sup 2}/s, irrespective of materials which had large difference before irradiation. Microstructural observation of irradiated materials by TEM revealed that Al{sub 2}O{sub 3} contained high-density loops, microvoids in grains, and microcracking along grain boundaries, AlN contained high-density loops and microcracking along grain boundaries, SiC contained high-density loops, and Si{sub 3}N{sub 4} contained loops lying on the planes parallel to the c-axis, respectively. Macroscopic length of Al{sub 2}O{sub 3} and AlN started to recover at around 800deg or 1100degC, respectively, irrespective of irradiation temperature, and reduced quickly. Macroscopic length of SiC recovered gradually from near the irradiation temperature

  2. Design and Demonstration of a Material-Plasma Exposure Target Station for Neutron Irradiated Samples

    Energy Technology Data Exchange (ETDEWEB)

    Rapp, Juergen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Aaron, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bell, Gary L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burgess, Thomas W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lessard, Timothy L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ohriner, Evan Keith [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Perkins, Dale E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Varma, Venugopal Koikal [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-20

    Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady

  3. Investigation of the agglomeration and amorphous transformation effects of neutron irradiation on the nanocrystalline silicon carbide (3C-SiC) using TEM and SEM methods

    Energy Technology Data Exchange (ETDEWEB)

    Huseynov, Elchin M., E-mail: elchin.h@yahoo.com [Department of Nanotechnology and Radiation Material Science, National Nuclear Research Center, Inshaatchilar pr. 4, AZ 1073 Baku (Azerbaijan); Institute of Radiation Problems of Azerbaijan National Academy of Sciences, B.Vahabzade 9, AZ 1143 Baku (Azerbaijan)

    2017-04-01

    Nanocrystalline 3C-SiC particles irradiated by neutron flux during 20 h in TRIGA Mark II light water pool type research reactor. Silicon carbide nanoparticles were analyzed by Scanning Electron Microscope (SEM) and Transmission Electron Microscopy (TEM) devices before and after neutron irradiation. The agglomeration of nanoparticles was studied comparatively before and after neutron irradiation. After neutron irradiation the amorphous layer surrounding the nanoparticles was analyzed in TEM device. Neutron irradiation defects in the 3C-SiC nanoparticles and other effects investigated by TEM device. The effect of irradiation on the crystal structure of the nanomaterial was studied by selected area electron diffraction (SAED) and electron diffraction patterns (EDP) analysis.

  4. Effect of Neutron Irradiation on Properties of Pb(Mg(1/3)Nb(2/3))O3-PbTiO3.

    Science.gov (United States)

    Kim, Yong-Il; Choi, Namkyoung; Kim, Geunwoo; Lee, Yun-Hee; Baek, Kwang-Sae; Kim, Ki-Bok

    2015-11-01

    The effect of neutron irradiation on the electrical and piezoelectric properties of a PMN-PT [(Pb(Mg(1/3)Nb(2/3))O3-PbTiO3)] single crystal such as permittivity, electrical impedance and piezoelectric constant d33 has been investigated at 1 kHz. The changes of d33 and permittivity depending on the dose of neutron irradiation for all samples of PMN-PT single crystal were found. In all samples, the permittivity, and piezoelectric constant d33 decreased with the increase of irradiation dose. Changes of XRD patterns depending on the dose of neutron irradiation for all samples were found. From the results of XRDs for analyzing the formation of the PMN-PT single crystals in single phase, the neutron irradiation will affect the crystallinity of PMN-PT single crystals.

  5. Investigation of the agglomeration and amorphous transformation effects of neutron irradiation on the nanocrystalline silicon carbide (3C-SiC) using TEM and SEM methods

    Science.gov (United States)

    Huseynov, Elchin M.

    2017-04-01

    Nanocrystalline 3C-SiC particles irradiated by neutron flux during 20 h in TRIGA Mark II light water pool type research reactor. Silicon carbide nanoparticles were analyzed by Scanning Electron Microscope (SEM) and Transmission Electron Microscopy (TEM) devices before and after neutron irradiation. The agglomeration of nanoparticles was studied comparatively before and after neutron irradiation. After neutron irradiation the amorphous layer surrounding the nanoparticles was analyzed in TEM device. Neutron irradiation defects in the 3C-SiC nanoparticles and other effects investigated by TEM device. The effect of irradiation on the crystal structure of the nanomaterial was studied by selected area electron diffraction (SAED) and electron diffraction patterns (EDP) analysis.

  6. Simultaneous impact of neutron irradiation and sputtering on the surface structure of self–damaged ITER–grade tungsten

    Directory of Open Access Journals (Sweden)

    A. I. Belyaeva

    2014-07-01

    Full Text Available Simultaneous effects of neutron irradiation and long–term sputtering on the surface relief of ITER–grade tungsten were studied. The effects of neutron–induced displacement damage have been simulated by irradiation of tungsten target with W6 + ions of 20 MeV energy. Ar+ ions with energy 600 eV were used as imitation of charge exchange atoms in ITER. The surface relief was studied after each sputtering act. The singularity in the WJ–IG surface relief was ascertained experimentally at the first time, which determines the law of roughness extension under sputtering. As follows from the experimental data, the neutron irradiation has not to make a decisive additional contribution in the processes developing under impact of charge exchange atoms only.

  7. On the onset of void swelling in pure tungsten under neutron irradiation: An object kinetic Monte Carlo approach

    Science.gov (United States)

    Castin, N.; Bakaev, A.; Bonny, G.; Sand, A. E.; Malerba, L.; Terentyev, D.

    2017-09-01

    We propose an object kinetic Monte Carlo (OKMC) model for describing the microstructural evolution in pure tungsten under neutron irradiation. We here focus on low doses (under 1 dpa), and we neglect transmutation in first approximation. The emphasis is mainly centred on an adequate description of neutron irradiation, the subsequent introduction of primary defects, and their thermal diffusion properties. Besides grain boundaries and the dislocation network, our model includes the contribution of carbon impurities, which are shown to have a strong influence on the onset of void swelling. Our parametric study analyses the quality of our model in detail, and confronts its predictions with experimental microstructural observations with satisfactory agreement. We highlight the importance for an accurate determination of the dissolved carbon content in the tungsten matrix, and we advocate for an accurate description of atomic collision cascades, in light of the sensitivity of our results with respect to correlated recombination.

  8. Separation of Protactinium from Neutron Irradiated Thorium Oxide; Separacion de Protactinio de Oxido de Torio Irradiado con Neutrones

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez, G.; Gutierrez, L.; Ropero, M.

    1983-07-01

    The chemical separation of thorium and protactinium can be carried out by leaching most of the last one, about 95%, with aqueous HF from neutron irradiated thorium oxide. This leaching reaction la highly favored by the transformation reaction of the ThO{sub 2} material into ThF{sub 4}. For both reactions, leaching and transformation, the reagents concentration, agitation speed and temperature influences were studied and the activation energies were found. (Author) 18 refs.

  9. SANS response of VVER440-type weld material after neutron irradiation, post-irradiation annealing and reirradiation

    OpenAIRE

    Ulbricht, Andreas; Bergner, Frank; Boehmert, Juergen; Valo, Matti; Mathon, Marie-Helene; Heinemann, Andre

    2007-01-01

    Abstract It is well accepted that the reirradiation behaviour of reactor pressure vessel (RPV) steel after annealing can be different from the original irradiation behaviour. We present the first small-angle neutron scattering (SANS) study of neutron irradiated, annealed and reirradiated VVER440-type RPV weld material. The SANS results are analysed both in terms of the size distribution of irradiation-induced defect/solute atom clusters and in terms of the ratio of total and nuclea...

  10. Prenatal exposure to gamma/neutron irradiation: Sensorimotor alterations and paradoxical effects on learning

    Energy Technology Data Exchange (ETDEWEB)

    Di Cicco, D.; Antal, S.; Ammassari-Teule, M. (Istituto di Psicobiologia e Psicofarmacologia del CNR, Rome (Italy))

    1991-01-01

    The effects of prenatal exposure on gamma/neutron radiations (0.5 Gy at about the 18th day of fetal life) were studied in a hybrid strain of mice (DBA/Cne males x C57BL/Cne females). During ontogeny, measurements of sensorimotor reflexes revealed in prenatally irradiated mice (1) a delay in sensorial development, (2) deficits in tests involving body motor control, and (3) a reduction of both motility and locomotor activity scores. In adulthood, the behaviour of prenatally irradiated and control mice was examined in the open field test and in reactivity to novelty. Moreover, their learning performance was compared in several situations. The results show that, in the open field test, only rearings were more frequent in irradiated mice. In the presence of a novel object, significant sex x treatment interactions were observed since ambulation and leaning against the novel object increased in irradiated females but decreased in irradiated males. Finally, when submitted to different learning tasks, irradiated mice were impaired in the radial maze, but paradoxically exhibited higher avoidance scores than control mice, possibly because of their low pain thresholds. Taken together, these observations indicate that late prenatal gamma/neutron irradiation induces long lasting alterations at the sensorimotor level which, in turn, can influence learning abilities of adult mice.

  11. PGNAA system preliminary design and measurement of In-Hospital Neutron Irradiator for boron concentration measurement.

    Science.gov (United States)

    Zhang, Zizhu; Chong, Yizheng; Chen, Xinru; Jin, Congjun; Yang, Lijun; Liu, Tong

    2015-12-01

    A prompt gamma neutron activation analysis (PGNAA) system has been recently developed at the 30-kW research reactor In-Hospital Neutron Irradiator (IHNI) in Beijing. Neutrons from the specially designed thermal neutron beam were used. The thermal flux of this beam is 3.08×10(6) cm(-2) s(-1) at a full reactor power of 30 kW. The PGNAA system consists of an n-type high-purity germanium (HPGe) detector of 40% efficiency, a digital spectrometer, and a shielding part. For both the detector shielding part and the neutron beam shielding part, the inner layer is composed of (6)Li2CO3 powder and the outer layer lead. The boron-10 sensitivity of the PGNAA system is approximately 2.5 cps/ppm. Two calibration curves were produced for the 1-10 ppm and 10-50 ppm samples. The measurement results of the control samples were in accordance with the inductively coupled plasma atomic emission spectroscopy (ICP-AES) results.

  12. 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.

  13. Effect and suppression of parasitic surface damage in neutron irradiated CMOS Monolithic Active Pixel Sensors

    CERN Document Server

    Deveaux, M; Scharrer, P; Stroth, J

    2016-01-01

    CMOS Monolithic Active Pixel Sensors (MAPS) were chosen as sensor technology for the vertex detectors of STAR, CBM and the upgraded ALICE-ITS. They also constitute a valuable option for tracking devices at future e+e- colliders. Those applications require a substantial tolerance to both, ionizing and non-ionizing radiation. To allow for a focused optimization of the radiation tolerance, prototypes are tested by irradiating the devices either with purely ionizing radiation (e.g. soft X-rays) or the most pure sources of non-ionizing radiation available (e.g. reactor neutrons). In the second case, it is typically assumed that the impact of the parasitic $\\gamma$-rays found in the neutron beams is negligible. We checked this assumption by irradiating MAPS with $\\gamma$-rays and comparing the radiation damage generated with the one in neutron irradiated sensors. We conclude that the parasitic radiation doses may cause non-negligible radiation damage. Based on the results we propose a procedure to recognize and to ...

  14. The neutron irradiation effect on the mechanical properties and structure of beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Fabritsiev, S.A. [D.V. Efremov Scientific Research Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S.; Bagautdinov, R.M. [Scientific Research Inst. of Atomic Reactors, Dimitrovgrad (Russian Federation)

    1999-10-01

    The neutron irradiation effect on the mechanical properties and structure of beryllium are presented. Irradiation was performed in the BOR-60 reactor up to doses of 0.7--1.1 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) at irradiation temperatures of 350 C, 400 C, 520 C, 780 C. Two modifications of RF beryllium, i.e., DShG-200 and TShG-56, were chosen for investigation. For irradiation at temperatures of 350--400 C Be hardening due to the accumulation of radiation defect complexes. Hardening is accompanied with a sharp drop in plasticity at T{sub test} {le} 300 C. The fracture of samples is of brittle, mainly transcrystallite, type. High-temperature irradiation (T{sub irr} = 780 C) gives rise to large helium pores over the grain boundaries and smaller pores in the grain body. Fracture is brittle and intercrystalline at T{sub test} {ge} 600 C. Helium embrittlement is also accompanied with a drop in the Be mechanical properties. The conclusion is made that the irradiation temperature range, where irradiated beryllium has a satisfactory level of properties, is rather narrow: 300 C {le} T {le} 500 C.

  15. Investigation of reactor neutron irradiation induced dark signals increase in COTS array CCDs

    Directory of Open Access Journals (Sweden)

    Zujun Wang

    2014-09-01

    Full Text Available The experiments of reactor neutron irradiation which induce dark signal increase in COTS array CCDs are presented. The flux of the reactor neutron beams was about 1.33 × 108 n/cm2s. The three samples were exposed to 1MeV neutron-equivalent fluences of 1 × 1011, 5 × 1011, and 1 × 1012 n/cm2, respectively. The mean dark signal (KD, dark signal non-uniformity (DSNU, and dark signal spikes (hot pixels versus neutron fluence are investigated. The degradation mechanisms of the dark signal in CCDs are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from the CCDs irradiated by neutrons are presented to investigate the generation of dark signal spike. The 1D and 2D figures which show the output signal voltage of pixels in dark images irradiated by different neutron beam fluences, are presented to compare the degradation of KD, DSNU, and dark signal spike.

  16. Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

    Directory of Open Access Journals (Sweden)

    Zujun Wang

    2014-07-01

    Full Text Available The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 μm CMOS technology. The flux of neutron beams was about 1.33 × 108 n/cm2s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 × 1011, 5 × 1011, and 1 × 1012 n/cm2, respectively. The mean dark signal (KD, dark signal spike, dark signal non-uniformity (DSNU, noise (VN, saturation output signal voltage (VS, and dynamic range (DR versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike.

  17. Views of TAGSI on effects of neutron irradiation on ductile tearing in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Metallurgy and Materials, University of Birmingham, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Technical and Assurance Services, Walton House, 404 Faraday Street, Birchwood Park, Warrington WA3 6GA (United Kingdom)], E-mail: david.lidbury@serco.com

    2009-07-15

    The paper reviews information pertaining to effects of neutron irradiation on 'upper-shelf' Charpy impact behaviour and on elastic/plastic fracture mechanics characterising parameters, again for 'upper shelf' conditions, in which the initiation and early growth of a crack involve ductile tearing. The hardening and associated reduction in strain-hardening capacity induced by irradiation gives rise to a decrease in Charpy upper shelf energy. Effects on J-based parameters are more complicated. The material resistance parameters tend to increase for low dose, but decrease at high dose, when the decrease in crack-tip ductility outweighs the effect of hardening. High doses can produce 'fast shear' fracture, which propagates rapidly and is therefore more likely to induce brittle cleavage fracture. The situation is exacerbated if the irradiation also promotes inter-granular segregation and fracture, hence reducing the local brittle fracture stress. For the levels of irradiation experienced by the types of UK civil reactors in operation, no fracture instability is expected to arise as a result of ductile fracture mechanisms alone.

  18. Localised vibrational mode spectroscopy studies of self-interstitial clusters in neutron irradiated silicon

    Energy Technology Data Exchange (ETDEWEB)

    Londos, C. A.; Antonaras, G. [University of Athens, Solid State Physics Section, Panepistimiopolis Zografos, Athens 157 84 (Greece); Chroneos, A. [Materials Engineering, The Open University, Milton Keynes MK7 6AA (United Kingdom); Department of Materials, Imperial College, London SW7 2AZ (United Kingdom)

    2013-07-28

    The evolution of self-interstitial clusters in silicon (Si), produced by fast neutron irradiation of silicon crystals followed by anneals up to 750 °C, is investigated using localised vibrational mode spectroscopy. A band at 582 cm{sup −1} appears after irradiation and is stable up to 550 °C was attributed to small self-interstitial clusters (I{sub n}, n ≤ 4), with the most probable candidate the I{sub 4} structure. Two bands at 713 and 758 cm{sup −1} arising in the spectra upon annealing of the 582 cm{sup −1} band and surviving up to ∼750 °C were correlated with larger interstitial clusters (I{sub n}, 5 ≤ n ≤ 8), with the most probable candidate the I{sub 8} structure or/and with chainlike defects which are precursors of the (311) extended defects. The results illustrate the presence of different interstitial clusters I{sub n}, at the various temperature intervals of the material, in the course of an isochronal anneal sequence. As the annealing temperature increases, they evolve from first-order structures with a small number of self-interstitials (I{sub n}, n ≤ 4) for the temperatures 50 < T < 550 °C, to second order structures (I{sub n}, 5 ≤ n ≤ 8) with a larger number of interstitials, for the temperatures 550 < T < 750 °C.

  19. Neutron flux assessment of a neutron irradiation facility based on inertial electrostatic confinement fusion.

    Science.gov (United States)

    Sztejnberg Gonçalves-Carralves, M L; Miller, M E

    2015-12-01

    Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutron flux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutron flux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutron fluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutron fluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice.

  20. Flux dependence of cluster formation in neutron-irradiated weld material

    Science.gov (United States)

    Bergner, F.; Ulbricht, A.; Hein, H.; Kammel, M.

    2008-03-01

    The effect of neutron flux on the formation of irradiation-induced clusters in reactor pressure vessel (RPV) steels is an unresolved issue. Small-angle neutron scattering was measured for a neutron-irradiated RPV weld material containing 0.22 wt% impurity Cu. The experiment was focused on the influence of neutron flux on the formation of irradiation-induced clusters at fixed fluence. The aim was to separate and tentatively interpret the effect of flux on the characteristics of the cluster size distribution. We have observed a pronounced effect of neutron flux on cluster size, whereas the total volume fraction of irradiation-induced clusters is insensitive to the level of flux. The result is compatible with a rate theory model according to which the range of applied fluxes covers the transition from a flux-independent regime at lower fluxes to a regime of decelerating cluster growth. The results are confronted with measured irradiation-induced changes of mechanical properties. Despite the observed flux effect on cluster size, both yield stress increase and transition temperature shift turned out to be independent of flux. This is in agreement with the volume fraction of irradiation-induced clusters being insensitive to the level of flux.

  1. The effect of neutron irradiation on oxygen aggregation processes in Si material treated under hydrostatic pressure

    Energy Technology Data Exchange (ETDEWEB)

    Londos, Charalampos A.; Andrianakis, Andreas [Solid State Section, Physics Department, University of Athens, Panepistimiopolis, Zografos, 157 84 Athens (Greece); Misiuk, Andrzej [Institute of Electron Technology, Al. Lotnikow 46, 02-668 Warsaw (Poland)

    2011-03-15

    Silicon is the dominant material in electronic industry. Its use for various applications requires processing stages, important among them those involving thermal treatments. Such treatments in Si trigger the mechanisms of oxygen aggregation resulting in the formation of oxygen precipitates which have important influence on the quality of the material. In the present work, we have investigated the effect of thermal treatments, with or without the application of high hydrostatic pressure, on the development of oxygen precipitates. We have particularly studied the effect of neutron irradiation on the formation of the various oxygen agglomerates in the course of the above treatments. To this end, Si samples initially irradiated by neutrons were subjected to high temperature or/and high temperature-high pressure treatments at 1000 and 1130 C. Afterwards, infrared (IR) measurements were undertaken to study various precipitate morphologies, in particular those giving rise to an IR band around 1080 cm{sup -1} related to octahedral-shaped precipitates and an IR band at 1225 cm{sup -1} attributed to platelet-shaped precipitates. The obtained results were found to be consistent with reports cited in the literature. It was confirmed that the application of pressure during treatments as well as the irradiation with neutrons before these treatments enhance substantially the oxygen aggregation process. Comparisons of the results between treatments at 1000 and 1130 C are presented and discussed. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. a Study of Stress Relaxation Rate in Un-Irradiated and Neutron-Irradiated Stainless Steel

    Science.gov (United States)

    Ghauri, I. M.; Afzal, Naveed; Zyrek, N. A.

    Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2-12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

  3. Fracture behavior of neutron-irradiated high-manganese austenitic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1991-03-01

    The instrumented Charpy impact test was applied to study the fracture behavior of high-manganese austenitic steels before and after neutron irradiations. Quarter-size specimens of a commercial high-manganese steel (18% Mn-5% Ni-16% Cr), three reference steels (21% Mn-1% Ni-9% Cr, 20% Mn-1% Ni-11% Cr, 15% Mn-1% Ni-13% Cr) and two model steels (17% Mn-4.5% Si-6.5% Cr, 22% Mn-4.5% Si-6.5% Cr-0.2% N) were used for the impact tests at temperatures between 77 and 523 K. The load-deflection curves showed typical features corresponding to characteristics of the fracture properties. The temperature dependences of fracture energy and failure deflection obtained from the curves clearly demonstrate only small effects up to 2 × 10 23 n/m 2 ( E > 0.1 MeV) and brittleness at room temperature in 17% Mn-Si-Cr steel at 1.6 × 10 25 n/m 2 ( E > 0.1 MeV), while ductility still remains in 22%Mn-Si-Cr steel.

  4. Fracture behaviors of neutron-irradiated ferritic steels studied by the instrumented charpy impact test

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1989-12-01

    The instrumented Charpy impact test for quarter-size specimens was developed and applied to study fracture behavior of ferritic steels and a ferritic-martensitic steel (JFMS) before and after neutron irradiation. The load-deflection curves obtained for U- and V-notched specimens showed typical characteristics of fracture properties of these steels. The temperature dependence of the fracture energy ( Ef) and the failure deflection ( Df) clearly indicates ductile-brittle transition and the DBTT can be determined from the Ef and Df versus temperature curves. The V-notched specimens showed sharper transition at higher temperatures for the JFMS than the U-notched ones, where the former were sensitive to brittle fracture and the latter well demonstrated the behavior of crack propagation. For the ferritic steels the DBTTs showed low values at compositions containing approximate 8-10% Cr and the increase of the DBTT (Δ DBTT) due to irradiation also showed a similar tendency. The Δ DBTT appeared to be relatively larger for the JFMS than the ferritic steels.

  5. Effect of neutron irradiation on charge collection efficiency in 4H-SiC Schottky diode

    Science.gov (United States)

    Wu, Jian; Jiang, Yong; Lei, Jiarong; Fan, Xiaoqiang; Chen, Yu; Li, Meng; Zou, Dehui; Liu, Bo

    2014-01-01

    The charge collection efficiency (CCE) in 4H-SiC Schottky diode is studied as a function of neutron fluence. The 4H-SiC diode was irradiated with fast neutrons of a critical assembly in Nuclear Physics and Chemistry Institute and CCE for 3.5 MeV alpha particles was then measured as a function of the applied reverse bias. It was found from our experiment that an increase of neutron fluence led to a decrease of CCE. In particular, CCE of the diode was less than 1.3% at zero bias after an irradiation at 8.26×1014 n/cm2. A generalized Hecht's equation was employed to analyze CCE in neutron irradiated 4H-SiC diode. The calculations nicely fit the CCE of 4H-SiC diode irradiated at different neutron fluences. According to the calculated results, the extracted electron μτ product (μτ)e and hole μτ product (μτ)h of the irradiated 4H-SiC diode are found to decrease by increasing the neutron fluence.

  6. Neutron-irradiation creep of silicon carbide materials beyond the initial transient

    Science.gov (United States)

    Koyanagi, Takaaki; Katoh, Yutai; Ozawa, Kazumi; Shimoda, Kazuya; Hinoki, Tatsuya; Snead, Lance L.

    2016-09-01

    Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. The materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber-reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ∼1 GPa for the fibers and ∼300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ∼1 × 10-7 [dpa-1 MPa-1] at 430-750 °C for the range of 1-30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures-such as grain boundary, crystal orientation, and secondary phases-increase with increasing irradiation temperature.

  7. Effect of the carbide phase on the tribological properties of high-manganese antiferromagnetic austenitic steels alloyed with vanadium and molybdenum

    Science.gov (United States)

    Korshunov, L. G.; Kositsina, I. I.; Sagaradze, V. V.; Chernenko, N. L.

    2011-07-01

    Effect of special carbides (VC, M 6C, Mo2C) on the wear resistance and friction coefficient of austenitic stable ( M s below -196°C) antiferromagnetic ( T N = 40-60°C) steels 80G20F2, 80G20M2, and 80G20F2M2 has been studied. The structure and the effective strength (microhardness H surf, shear resistance τ) of the surface layer of these steels have been studied using optical and electron microscopy. It has been shown that the presence of coarse particles of primary special carbides in the steels 80G20F2, 80G20M2, and 80G20F2M2 quenched from 1150°C decreases the effective strength and the resistance to adhesive and abrasive wear of these materials. This is caused by the negative effect of carbide particles on the toughness of steels and by a decrease in the carbon content in austenite due to a partial binding of carbon into the above-mentioned carbides. The aging of quenched steels under conditions providing the maximum hardness (650°C for 10 h) exerts a substantial positive effect on the parameters of the effective strength ( H surf, τ) of the surface layer and, correspondingly, on the resistance of steels to various types of wear (abrasive, adhesive, and caused by the boundary friction). The maximum positive effect of aging on the wear resistance is observed upon adhesive wear of the steels under consideration. Upon friction with enhanced sliding velocities (to 4 m/s) under conditions of intense (to 500-600°C) friction-induced heating, the 80G20F2, 80G20M2, and, especially, 80G20F2M2 steels subjected to quenching and aging substantially exceed the 110G13 (Hadfield) steel in their tribological properties. This is due to the presence in these steels of a favorable combination of high effective strength and friction heat resistance of the surface layer, which result from the presence of a large amount of special carbides in these steels and from a high degree of alloying of the matrix of these steels by vanadium and molybdenum. In the process of friction

  8. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de

    2008-07-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature

  9. Effects of as-cast and wrought Cobalt-Chrome-Molybdenum and Titanium-Aluminium-Vanadium alloys on cytokine gene expression and protein secretion in J774A.1 macrophages

    DEFF Research Database (Denmark)

    Jakobsen, Stig Storgaard; Larsen, Agnete; Stoltenberg, Meredin

    2007-01-01

    to the metal implant and wear-products. The aim of the present study was to compare surfaces of as-cast and wrought Cobalt-Chrome-Molybdenum (CoCrMo) alloys and Titanium-Aluminium-Vanadium (TiAlV) alloy when incubated with mouse macrophage J774A.1 cell cultures. Changes in pro- and anti-inflammatory cytokines...... the cell viability. Surface properties of the discs were characterised with a profilometer and with energy dispersive X-ray spectroscopy. We here report, for the first time, that the prosthetic material surface (non-phagocytable) of as-cast high carbon CoCrMo reduces the pro-inflammatory cytokine IL-6...... transcription, the chemokine MCP-1 secretion, and M-CSF secretion by 77%, 36%, and 62%, respectively. Furthermore, we found that reducing surface roughness did not affect this reduction. The results suggest that as-cast CoCrMo alloy is more inert than wrought CoCrMo and wrought TiAlV alloys and could prove...

  10. Molybdenum: Industrial applications

    Energy Technology Data Exchange (ETDEWEB)

    Bulla, W.; Fairhurst, W.; King, P.

    1982-12-01

    Molybdenum availability and molybdenum consumption are reviewed. About 9% of the total amount of molybdenum consumed is used in the production of nonferrous metals. The authors list applications for which molybdenum materials, owing to their physical and high-temperature characteristics, are particularly well suited.

  11. Life span, testis damage and immune cell populations of spleen in C57BL mice with neutron irradiation by lying flat pose

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ki Jung; kim, Myung Sup; Kyung, Yoo Bo [KAERI, Taejon (Korea)

    2003-10-01

    This study deals with the biological effects of black mouse (C57BL) irradiated with neutron irradiation by using Boron Neutron Capture Therapy facility in HANARO reactor. These include mortality, body wt., hair color, testis volume, sperm count and immune cell populations in mouse spleen after 80 days later by thermal neutron irradiation. Six week old C57BL male mice were irradiated with neutron irradiation for 1 hr or 2 hrs (flux : 1.036739E +09). These irradiat ion doses estimated 15Gy and 30Gy, respectively. Survival days and hair color in mice was checked. On day 80 after irradiation, testis were taken for volume and sperm count. Also spleen was taken for FACS and spleen cells were isolatd and discarded RBC by treating with lysising solution. These cells were placed on ice and immunofluorescence staining was performed. Phycoerythrin (PE )-anti-CD3e, fluorescein isothiocyanate (FITC)-anti-CD4, and FITC-anti-CD8 were added, then the immunostaining cells were incubated on ice for 40 min. The resulting cells were washed with a PBS buffer 3 times and analyzed using a Flow cytometer. All experimental animals survived over 90 days but in case of 30 Gy neutron irradiation, black mice hair were changed white color on the center of the back. Neutron irradiation of black mice show similar in damage of spleen immune cells by subpopulation of T helper and T cytotoxic cells compared to the control non - irradiated group. These results show that treatment of neutron irradiation without boron compounds for 2 hrs in mice can survive over 90 days with hair color change from black to white. Damaged spleen cells recover after long time by irradiation but testis volume and no. of sperm are not recover compared to the normal group in response to neutron irradiation.

  12. Study of the VMM1 read-out chip in a neutron irradiation environment

    Science.gov (United States)

    Alexopoulos, T.; Fanourakis, G.; Geralis, T.; Kokkoris, M.; Kourkoumeli-Charalampidi, A.; Papageorgiou, K.; Tsipolitis, G.

    2016-05-01

    Within 2015, the LHC operated close to the design energy of √s = 13-14 TeV delivering instantaneous luminosities up to Script L = 5 × 1033 cm-2s-1. The ATLAS Phase-I upgrade in 2018/19 will introduce the MicroMEGAS detectors in the area of the small wheel at the end caps. Accompanying new electronics are designed and built such as the VMM front end ASIC, which provides energy, timing and triggering information and allows fast data read-out. The first VMM version (VMM1) has been widely produced and tested in various test beams, whilst the second version (VMM2) is currently being tested. This paper focuses on the VMM1 single event upset studies and more specifically on the response of the configuration registers under harsh radiation environments. Similar conditions are expected at Run III with Script L = 2 × 1034 cm-2s-1 and a mean of 55 interactions per bunch crossing. Two VMM1s were exposed in a neutron irradiation environment using the TANDEM Van Der Graaff accelerator at NSCR Demokritos, Athens, Greece. The results showed a rate of SEU occurrences at a measured cross section of (4.1±0.8)×10-14 cm2/bit for each VMM. Consequently, when extrapolating this value to the luminosity expected in Run III, the occurrence is roughly 6 SEUs/min in all the read-out system comprising 40,000 VMMs installed during the Phase-I upgrade.

  13. The influence of low dose neutron irradiation on the thermal conductivity of Allcomp carbon foam

    Energy Technology Data Exchange (ETDEWEB)

    Burchell, Timothy D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Porter, Wallace D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-03-01

    Oak Ridge National Laboratory was contracted via a Work for Others Agreement with Allcomp Inc. (NFE-14-05011-MSOF: Carbon Foam for Beam Stop Applications ) to determine the influence of low irradiation dose on the thermal conductivity of Allcomp Carbon Foam. Samples (6 mm dia. x 5 mm thick) were successfully irradiated in a rabbit capsule in a hydraulic tube in the target region of the High Flux Isotope Reactor at the Oak Ridge National Laboratory. The specimens were irradiated at Tirr = 747.5 C to a neutron damage dose of 0.2 dpa. There is a small dimensional and volume shrinkage and the mass and density appear reduced (we would expect density to increase as volume reduces at constant mass). The small changes in density, dimensions or volume are not of concern. At 0.2 dpa the irradiation shrinkage rate difference between the glassy carbon skeleton and the CVD coating was not sufficient to cause a large enough irradiation-induced strain to create any mechanical degradation. Similarly differential thermal expansion was not a problem. It appears that only the thermal conductivity was affected by 0.2 dpa. For the intended application conditions, i.e. @ 400 C and 0 DPA (start- up) the foam thermal conductivity is about 57 W/m.K and at 700 C and 0.2 DPA (end of life) the foam thermal conductivity is approx. 30.7 W/m.K. The room temp thermal conductivity drops from 100-120 W/m.K to approximately 30 W/m.K after 0.2 dpa of neutron irradiation.

  14. Irradiation temperature dependence of production efficiency of lattice defects in some neutron-irradiated oxides

    Energy Technology Data Exchange (ETDEWEB)

    Okada, Moritami [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 5900494 (Japan)]. E-mail: okada@rri.kyoto-u.ac.jp; Atobe, Kozo [Faculty of Science, Naruto University of Education, Naruto, Tokushima 7728502 (Japan); Nakagawa, Masuo [Faculty of Education, Kagawa University, Takamatsu, Kagawa 7608522 (Japan)

    2004-11-01

    Temperature dependence of production efficiency of irradiation-induced defects in neutron-irradiated oxides has been investigated. Some oxide single crystals, MgO, {alpha}-Al{sub 2}O{sub 3} (sapphire) and TiO{sub 2} (rutile), were irradiated at several controlled temperatures, 10, 20, 50, 100, 150 and 200 K, using the low-temperature irradiation facility of Kyoto University Reactor (KUR-LTL), and at ambient temperature ({approx}370 K) in the same facility. Irradiation temperature dependence of production efficiency of a 1 {mu}m band in TiO{sub 2} differs greatly from that of anion vacancy (F-type centers) in MgO and {alpha}-Al{sub 2}O{sub 3}. Results for MgO and {alpha}-Al{sub 2}O{sub 3} show steep negative gradients from 10 to 370 K, whereas that for TiO{sub 2} includes a valley between 40 and 60 K and a hump at about 130 K, and then disappear at about 200 K. In MgO and {alpha}-Al{sub 2}O{sub 3}, this behavior can be explained by the recombination of Frenkel pairs, which is activated at higher temperature. In TiO{sub 2}, in addition to the recombination mechanism, a covalent bonding property is thought to be exerted strong influence, and it is suggested that a disappearance of the 1 {mu}m band at above 200 K is due to the recombination process of Frenkel pairs which is caused by the irradiation-induced crystallization.

  15. Modeling of displacement damage in silicon carbide detectors resulting from neutron irradiation

    Science.gov (United States)

    Khorsandi, Behrooz

    There is considerable interest in developing a power monitor system for Generation IV reactors (for instance GT-MHR). A new type of semiconductor radiation detector is under development based on silicon carbide (SiC) technology for these reactors. SiC has been selected as the semiconductor material due to its superior thermal-electrical-neutronic properties. Compared to Si, SiC is a radiation hard material; however, like Si, the properties of SiC are changed by irradiation by a large fluence of energetic neutrons, as a consequence of displacement damage, and that irradiation decreases the life-time of detectors. Predictions of displacement damage and the concomitant radiation effects are important for deciding where the SiC detectors should be placed. The purpose of this dissertation is to develop computer simulation methods to estimate the number of various defects created in SiC detectors, because of neutron irradiation, and predict at what positions of a reactor, SiC detectors could monitor the neutron flux with high reliability. The simulation modeling includes several well-known---and commercial---codes (MCNP5, TRIM, MARLOWE and VASP), and two kinetic Monte Carlo codes written by the author (MCASIC and DCRSIC). My dissertation will highlight the displacement damage that may happen in SiC detectors located in available positions in the OSURR, GT-MHR and IRIS. As extra modeling output data, the count rates of SiC for the specified locations are calculated. A conclusion of this thesis is SiC detectors that are placed in the thermal neutron region of a graphite moderator-reflector reactor have a chance to survive at least one reactor refueling cycle, while their count rates are acceptably high.

  16. Enhancement of flux pinning properties in nanosized MgO added Bi-2212 superconductor through neutron irradiation

    Science.gov (United States)

    Mohiju, Zaahidah'Atiqah; Hamid, Nasri A.; Abdullah, Yusof

    2017-01-01

    For superconducting material to maintain high critical current density, Jc in any applications, effective flux pinning centers are needed. The addition of small size MgO particles in bulk Bi2Sr2CaCu2O8 (Bi-2212) superconductor has been proven to enhance the effective flux pinning centers in the superconducting material by creating a desired microstructure with appropriate defects. To further enhance the pinning properties, radiation is one of the convenient ways to improve the microstructure of the material that has correlation with basic properties of superconductors. Neutron irradiation is one of the niche techniques that can be used to perform the task. Defects with larger radius have dimension comparable to the coherence length of the material and thus improved its superconducting properties. In this paper, a small amount of nanosized MgO particles was used to create defects in the Bi-2212 superconducting material. The Bi-2212/MgO compounds were heat treated, followed by partial melting and slow cooling. Part of the samples was subjected to neutron irradiation using the TRIGA-MARK-II research reactor at the Malaysian Nuclear Agency. Characterization of non-irradiated and irradiated samples was performed via the temperature dependence on electrical resistance measurements, X-ray Diffraction Patterns (XRD), and Scanning Electron Microscopy (SEM) with Energy Dispersive X-ray (EDX) analysis. From the analysis, there was changed in the critical current density and transition temperature of samples subjected to neutron irradiation due to formation of point defects in the microstructure. Higher critical current density indicates better flux pinning properties in the Bi-2212/MgO compounds.

  17. Magnetization measurements on HoNi{sub 2}B{sub 2}C single crystals before and after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Fuger, R. [Atomic Institute of the Austrian Universities, 1020 Vienna (Austria)], E-mail: rfuger@ati.ac.at; Krutzler, C. [Atomic Institute of the Austrian Universities, 1020 Vienna (Austria); Fuchs, G.; Behr, G. [Institut fuer Festkoerper- und Werkstofforschung Dresden, 14109 Dresden (Germany); Weber, H.W. [Atomic Institute of the Austrian Universities, 1020 Vienna (Austria)

    2007-09-01

    A single crystal of HoNi{sub 2}B{sub 2}C was fully characterised by magnetization measurements. The magnetic and the superconducting phase diagram were determined in different crystal directions and the superconductive properties evaluated as a function of temperature and applied field. The critical current density was calculated from magnetization loops using the Bean model. The results on the critical current density reveal bulk pinning in those regions of the phase diagram, where superconductivity is not suppressed by metamagnetically ordered structures of the Ho 4f-moments. These measurements were repeated after neutron irradiation of the sample.

  18. Tritium release from neutron irradiated beryllium: Kinetics, long-time annealing and effect or crack formation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe, (Germany)

    1995-09-01

    Since beryllium is considered as one of the best neutron multiplier materials in the blanket of the next generation fusion reactors, several studies have been started to evaluate its behaviour under irradiation during both operating and accidental conditions. Based on safety considerations, tritium produced in beryllium during neutron irradiation represents one important issue, therefore it is necessary to investigate tritium transport processes by using a comprehensive mathematical model and comparing its predictions with well characterized experimental tests. Because of the difficulties in extrapolating the short-time tritium release tests to a longer time scale, also long-time annealing experiments with beryllium samples from the SIBELIUS irradiation. have been carried out at the Forschungszentrum Karlsruhe. Samples were annealed up to 12 months at temperatures up to 650{degrees}C. The inventory after annealing was determined by heating the samples up to 1050{degrees}C with a He+0.1 vo1% H{sub 2} purge gas. Furthermore, in order to investigate the likely effects of cracks formation eventually causing a faster tritium release from beryllium, the behaviour of samples irradiated at low temperature (40-50{degrees}C) but up to very high fast neutron fluences (0.8-3.9{center_dot}10{sup 22} cm{sup -2}, E{sub n}{ge}1 MeV) in the BR2 reactor has been investigated. Tritium was released by heating the beryllium samples up to 1050{degrees}C and purging them with He+0.1 vo1% H{sub 2}. Tritium release from high-irradiated beryllium samples showed a much faster kinetics than from the low-irradiated ones, probably because of crack formation caused by thermal stresses in the brittle material and/or by helium bubbles migration. The obtained experimental data have been compared with predictions of the code ANFIBE with the goal to better understand the physical mechanisms governing tritium behaviour in beryllium and to assess the prediction capabilities of the code.

  19. Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.K.

    2001-01-30

    A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

  20. Enhancement of critical current density in fast neutron irradiated melt-textured YBa 2Cu 3O 7- x

    Science.gov (United States)

    Puźniak, R.; Wiśniewski, A.; Baran, M.; Szymczak, H.; Pingxiang, Zhang; Jingrong, Wang; Lian, Zhou; Pytel, K.; Pytel, B.

    The critical current density in melt-textured samples obtained by the powder melting process was determined from magnetization measurements. Linear dependence between the width of the hysteresis loop and sample size was observed for both unirradiated and irradiated samples. This indicates that the critical current is circulating through the whole sample and is not disconnected by weak links, even when a magnetic field is applied in the irradiated sample. After fast neutron irradiation with fluences from 5 × 10 16 to 6 × 10 17 n cm -2 ( E > 0.5 MeV), significant enhancement of the critical current density, jc, was observed. Critical current density, determined from magnetization measurements, for magnetic field perpendicular to the a-b plane, jcab, reaches - 10 5 A cm 42 at 77 K in 1 T. For H parallel to the a-b plane, jcc along the c-axis reaches 5 × 10 3 A cm -2. An increase in the anisotropy of the critical current was observed after fast neutron irradiation in the temperature range 60 - 80 K.

  1. Optical studies of defects generated in neutron-irradiated Cz-Si during HP-HT treatment

    Energy Technology Data Exchange (ETDEWEB)

    Surma, B.; Wnuk, A. [Institute of Electronic Materials Technology, Wolczynska 133, 01-919 Warsaw (Poland); Misiuk, A. [Institute of Electron Technology, Al. Lotnikow 32/46, 02-668 Warsaw (Poland); Londos, C.A. [Department of Physics, Panepistimiopolis, GR-15784 Zografos, Athens (Greece); Bukowski, A. [Institute of Electronic Materials Technology, Wolczynska 133, 01-919 Warsaw (Poland); Silicon CEMAT, Wolczynska 133, 01-919 Warsaw (Poland)

    2005-04-01

    Neutron-irradiated Czochralski grown silicon subjected to heat treatment (HT) at 350 C and 1000 C under enhanced hydrostatic pressure (HP) was studied in this work. It has been shown that external hydrostatic pressure enhances the creation of VO{sub 2} defects in neutron irradiated silicon subjected to the HP-HT treatment at 350 C. Enhanced formation of platelet-like oxygen precipitates was found in the samples treated at 1000 C under 1.1 GPa. This effect was more pronounced in the samples with VO{sub 2} defects. Presented results seem to suggest that probably HP helps to transform VO{sub 2} to some kind of defects or change alone VO{sub 2} defects in the form that can act as an additional nucleus for an additional oxygen precipitation at 1000 C. No correlation between the plate-like oxygen precipitates related absorption at 1225 cm{sup -1} and dislocation-related emission has been confirmed. (copyright 2005 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  2. Effects of neutron irradiation on pinning force scaling in state-of-the-art Nb3Sn wires

    CERN Document Server

    Baumgartner, T; Weber, H W; Flükiger, R; Scheuerlein, C; Bottura, L

    2014-01-01

    We present an extensive irradiation study involving five state-of-the-art Nb3Sn wires which were subjected to sequential neutron irradiation up to a fast neutron fluence of 1.6 x 10(22) m(-2) (E > 0.1 MeV). The volume pinning force of short wire samples was assessed in the temperature range from 4.2 to 15 K in applied fields of up to 7 T by means of SQUID magnetometry in the unirradiated state and after each irradiation step. Pinning force scaling computations revealed that the exponents in the pinning force function differ significantly from those expected for pure grain boundary pinning, and that fast neutron irradiation causes a substantial change in the functional dependence of the volume pinning force. A model is presented, which describes the pinning force function of irradiated wires using a two-component ansatz involving a point-pinning contribution stemming from radiation induced pinning centers. The dependence of this point-pinning contribution on fast neutron fluence appears to be a universal funct...

  3. Study of the relative dose-response of BANG-3® polymer gel dosimeters in epithermal neutron irradiation

    Science.gov (United States)

    Uusi-Simola, J.; Savolainen, S.; Kangasmäki, A.; Heikkinen, S.; Perkiö, J.; Abo Ramadan, U.; Seppälä, T.; Karila, J.; Serén, T.; Kotiluoto, P.; Sorvari, P.; Auterinen, I.

    2003-09-01

    Polymer gels have been reported as a new, potential tool for dosimetry in mixed neutron-gamma radiation fields. In this work, BANG-3 (MGS Research Inc.) gel vials from three production batches were irradiated with 6 MV photons of a Varian Clinac 2100 C linear accelerator and with the epithermal neutron beam of the Finnish boron neutron capture therapy (BNCT) facility at the FiR 1 nuclear reactor. The gel is tissue equivalent in main elemental composition and density and its T2 relaxation time is dependent on the absorbed dose. The T2 relaxation time map of the irradiated gel vials was measured with a 1.5 T magnetic resonance (MR) scanner using spin echo sequence. The absorbed doses of neutron irradiation were calculated using DORT computer code, and the accuracy of the calculational model was verified by measuring gamma ray dose rate with thermoluminescent dosimeters and 55Mn(n,gamma) activation reaction rate with activation detectors. The response of the BANG-3 gel dosimeter for total absorbed dose in the neutron irradiation was linear, and the magnitude of the response relative to the response in the photon irradiation was observed to vary between different gel batches. The results support the potential of polymer gels in BNCT dosimetry, especially for the verification of two- or three-dimensional dose distributions.

  4. The effect of neutron irradiation on the AlGaN/GaN high electron mobility transistors

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Wenping; Hao, Yue; Yang, Lin' an; Duan, Chao; Duan, Huantao; Zhang, Jincheng; Ma, Xiaohua [School of Microelectronics, Xidian University, China Key Lab of Ministry of Education for Wide Band-Gap Semiconductor Materials and Devices, Xi' an 710071 (China)

    2010-07-15

    The SiN-passivated AlGaN/GaN high electron mobility transistors were investigated by 1MeV neutron irradiation at fluences up to 10{sup 15} cm{sup -2}, yielding a significant degradation for the transconductance near the knee voltage and the reverse gate leakage current at fluences ranging from 10{sup 14} to 10{sup 15} cm{sup -2} which could be attributed to the irradiation induced mobility shift and the defects in SiN passivation layers respectively since no any recovery was found after 20 hour annealing at room temperature. Meanwhile the negligible degradation of the saturation drain current, the maximal transconductance and the threshold voltage gave the fact that the effectiveness of SiN layers in passivated surface states in the source-gate spacer and gate-drain spacer was undiminished by neutron irradiation. Moreover the ohmic contact was robustness to neutron since the sheet resistance of ohmic contact region hardly shifted, but the schottky characteristics degraded obviously. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  5. Improvement of switching speed of a 600-V nonpunch through insulated gate bipolar transistor using fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Ha Ni; Sun, Gwang Min; Kim, Ji Suck; Hoang, Sy Minh Tuan; Jin, Mi Eun; Ahn, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-02-15

    Fast neutron irradiation was used to improve the switching speed of a 600-V nonpunch-through insulated gate bipolar transistor. Fast neutron irradiation was carried out at 30-MeV energy in doses of 1 × 10{sup 8} n/cm{sup 2}, 1 × 10{sup 9} n/cm{sup 2}, 1 × 10{sup 10} n/cm{sup 2}, and 1 × 10{sup 11} n/cm{sup 2}. Electrical characteristics such as current–voltage, forward on-state voltage drop, and switching speed of the device were analyzed and compared with those prior to irradiation. The on-state voltage drop of the initial devices prior to irradiation was 2.08 V, which increased to 2.10 V, 2.20 V, 2.3 V, and 2.4 V, respectively, depending on the irradiation dose. This effect arises because of the lattice defects generated by the fast neutrons. In particular, the turnoff delay time was reduced to 92 nanoseconds, 45% of that prior to irradiation, which means there is a substantial improvement in the switching speed of the device.

  6. Neutron irradiation and frequency effects on the electrical conductivity of nanocrystalline silicon carbide (3C-SiC)

    Science.gov (United States)

    Huseynov, Elchin

    2016-09-01

    In this present work nanocrystalline silicon carbide (3C-SiC) has been irradiated with neutron flux (∼ 2 ×1013 ncm-2s-1) up to 20 hours at different periods. Electrical conductivity of nanocrystalline 3C-SiC particles (∼18 nm) is comparatively analyzed before and after neutron irradiation. The frequency dependencies of electrical conductivity of 3C-SiC nanoparticles is reviewed at 100 K-400 K temperature range before and after irradiation. The measurements were carried out at 0.1 Hz-2.5 MHz frequency ranges and at different temperatures. Radiation-induced conductivity (RIC) was observed in the nanocrystalline 3C-SiC particles after neutron irradiation and this conductivity study as a function of frequency are presented. The type of conductivity has been defined based on the interdependence between real and imaginary parts of electrical conductivity function. Based on the obtained results the mechanism behind the electrical conductivity of nanocrystalline 3C-SiC particles is explained in detail.

  7. 稀土氧化物粒子对钼合金粉末冶金过程及力学性能的影响%EFFECTS OF RARE-EARTH PARTICLES ON MOLYBDENUM ALLOY POWDER METALLURGY PROCESSING AND MECHANICAL PROPERTIES

    Institute of Scientific and Technical Information of China (English)

    赵虎; 杨秦莉; 冯鹏发; 刘仁智; 付静波

    2013-01-01

    试验研究了掺杂LaO3、Y2 O3、CeO2稀土氧化物颗粒对钼合金的粉末物性、烧结进程、制品的烧结致密度及压力加工丝材的室温力学性能的影响规律.试验结果表明,掺杂稀土氧化物粒子细化了钼粉的粒度,降低了松装密度和粒度分布范围,同时导致粉末团聚现象增多;稀土氧化物粒子延迟了钼合金的烧结进程,降低了烧结制品的致密度,同时细化了烧结体晶粒尺寸.稀土氧化物粒子以弥散强化和细晶强化的形式,提高了钼合金丝的室温强度.CeO2显著提高了钼合金丝的室温韧性,La203、Y2O3则降低了钼合金丝的室温韧性.%The effects on the molybdenum alloy powder properties,sintering proceeding and sintering products density and mechanical properties of molybdenum alloy as-worked wire which was doped with rare-earth oxides particles such as La2O3 、Y2O3 、CeO2 were studied.The results showed that the particles size of the molybdenum alloy powder were refined,loose density and particle size distribution were reduced by doping rare-earth oxides particles,though at the same time molybdenum alloy powder aggregation were increased.Sintering proceeding was delayed,sintering products density was reduced and the grain size was refined by the rare-earth oxides particles during sintering process.The room temperature strength of the molybdenum alloy wire as-worked was increased by the grain refinement strengthening and dispersion strengthening of the doped rare-earth oxides particles.Room temperature toughness of molybdenum alloy wire as-worked was greatly improved as doped CeO2 particles but reduced as doped La2O3 and Y2O3 particles.

  8. Effects of AS-cast and wrought cobalt-chrome-molybdenum and titanium-aluminium-vanadium alloys on cytokine gene expression and protein secretion in J774A.1 macrophages

    Directory of Open Access Journals (Sweden)

    S S Jakobsen

    2007-09-01

    Full Text Available Insertion of metal implants is associated with a possible change in the delicate balance between pro- and anti-inflammatory proteins, probably leading to an unfavourable predominantly pro-inflammatory milieu. The most likely cause is an inappropriate activation of macrophages in close relation to the metal implant and wear-products. The aim of the present study was to compare surfaces of as-cast and wrought Cobalt-Chrome-Molybdenum (CoCrMo alloys and Titanium-Aluminium-Vanadium (TiAlV alloy when incubated with mouse macrophage J774A.1 cell cultures. Changes in pro- and anti-inflammatory cytokines [TNF-alpha, IL-6, IL-alpha, IL-1beta, IL-10] and proteins known to induce proliferation [M-CSF], chemotaxis [MCP-1] and osteogenesis [TGF-beta, OPG] were determined by ELISA and Real Time reverse transcriptase - PCR (Real Time rt-PCR. Lactate dehydrogenase (LDH was measured in the medium to asses the cell viability. Surface properties of the discs were characterised with a profilometer and with energy dispersive X-ray spectroscopy. We here report, for the first time, that the prosthetic material surface (non-phagocytable of as-cast high carbon CoCrMo reduces the pro-inflammatory cytokine IL-6 transcription, the chemokine MCP-1 secretion, and M-CSF secretion by 77 %, 36 %, and 62 %, respectively. Furthermore, we found that reducing surface roughness did not affect this reduction. The results suggest that as-cast CoCrMo alloy is more inert than wrought CoCrMo and wrought TiAlV alloys and could prove to be a superior implant material generating less inflammation which might result in less osteolysis.

  9. Effects of crystal defects on stress-corrosion susceptibility in aluminum alloy 7075

    Science.gov (United States)

    Bentle, G. G.; Jacobs, A. J.

    1970-01-01

    Point defects were introduced into specimens of three heat-treated tempers of alloy 7075 by neutron irradiation. Continuous ultrasonic monitoring allowed crack growth to be observed. Effects on stress-corrosion susceptibility, elongation, hardness, and yield strength are noted and compared for the three tempers.

  10. Contribution To Degradation Study, Behavior Of Unsaturated Polyester Resin Under Neutron Irradiation

    Science.gov (United States)

    Abellache, D.; Lounis, A.; Taïbi, K.

    2010-01-01

    Applications of unsaturated polyester thermosetting resins are numerous in construction sector, in transport, electric spare parts manufactures, consumer goods, and anticorrosive materials. This survey reports the effect of thermosetting polymer degradation (unsaturated polyester): degradation by neutrons irradiation. In order to evaluate the deterioration of our material, some comparative characterizations have been done between standard samples and damaged ones. Scanning electron microscopy (SEM), ultrasonic scanning, hardness test (Shore D) are the techniques which have been used. The exposure to a neutrons flux is carried out in the column of the nuclear research reactor of Draria (Algiers-Algeria). The energetic profile of the incidental fluxes is constituted of fast neutrons (ΦR = 3.1012n.cm-2.s-1, E = 2 Mev) of thermal neutrons (ΦTH = 1013n.cm-2.s-1; E = 0.025 ev) and epithermal neutrons (Φepi = 7.1011 n.cm-2.s-1; E>4,9 ev). The received dose flow is 0,4 Kgy. We notice only a few scientific investigations can be found in this field. In comparison with the standard sample (no exposed) it is shown that the damage degree is an increasing process with the exposure. Concerning the description of irradiation effects on polymers, we can advance that several reactions are in competition : reticulation, chain break, and oxidation by radical mechanism. In our case the incidental particle of high energy fast neutrons whose energy is greater or equal to 2 Mev, is braked by the target with a nuclear shock during which the incidental particle transmits a part of its energy to an atom. If the energy transfer is sufficient, the nuclear shock permits to drive out an atom of its site the latter will return positioning interstitially, the energy that we used oversteps probably the energy threshold (displacement energy). This fast neutrons collision with target cores proceeds to an indirect ionization by the preliminary creation of excited secondary species that will

  11. Radiation Effects in Refractory Alloys

    Science.gov (United States)

    Zinkle, Steven J.; Wiffen, F. W.

    2004-02-01

    In order to achieve the required low reactor mass per unit electrical power for space reactors, refractory alloys are essential due to their high operating temperature capability that in turn enables high thermal conversion efficiencies. One of the key issues associated with refractory alloys is their performance in a neutron irradiation environment. The available radiation effects data are reviewed for alloys based on Mo, W, Re, Nb and Ta. The largest database is associated with Mo alloys, whereas Re, W and Ta alloys have the least available information. Particular attention is focused on Nb-1Zr, which is a proposed cladding and structural material for the reactor in the Jupiter Icy Moons Orbiter (JIMO) project. All of the refractory alloys exhibit qualitatively similar temperature-dependent behavior. At low temperatures up to ~0.3TM, where TM is the melting temperature, the dominant effect of radiation is to produce pronounced radiation hardening and concomitant loss of ductility. The radiation hardening also causes a dramatic decrease in the fracture toughness of the refractory alloys. These low temperature radiation effects occur at relatively low damage levels of ~0.1 displacement per atom, dpa (~2×1024 n/m2, E>0.1 MeV). As a consequence, operation at low temperatures in the presence of neutron irradiation must be avoided for all refractory alloys. At intermediate temperatures (0.3 to 0.6 TM), void swelling and irradiation creep are the dominant effects of irradiation. The amount of volumetric swelling associated with void formation in refractory alloys is generally within engineering design limits (>10 dpa). Very little experimental data exist on irradiation creep of refractory alloys, but data for other body centered cubic alloys suggest that the irradiation creep will produce negligible deformation for near-term space reactor applications.

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

    Science.gov (United States)

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

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

  13. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Leng, B. [University of Wisconsin-Madison, Madison, WI 53706 (United States); Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Shanghai, 201800 (China); Rooyen, I.J. van, E-mail: Isabella.vanrooyen@inl.gov [Fuel Design and Development Department, Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Wu, Y.Q. [Department of Materials Science and Engineering, Boise State University, Boise, ID 83725-2090 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Szlufarska, I.; Sridharan, K. [University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2016-07-15

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd{sub 2}Si{sub 2}U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously. - Highlights: • First research data in neutron irradiated TRISO coated particles showing a Ag-Pd nano-sized precipitate inside a SiC grain. • Intragranular Ag Pd

  14. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

  15. Molybdenum, molybdenum oxides, and their electrochemistry.

    Science.gov (United States)

    Saji, Viswanathan S; Lee, Chi-Woo

    2012-07-01

    The electrochemical behaviors of molybdenum and its oxides, both in bulk and thin film dimensions, are critical because of their widespread applications in steels, electrocatalysts, electrochromic materials, batteries, sensors, and solar cells. An important area of current interest is electrodeposited CIGS-based solar cells where a molybdenum/glass electrode forms the back contact. Surprisingly, the basic electrochemistry of molybdenum and its oxides has not been reviewed with due attention. In this Review, we assess the scattered information. The potential and pH dependent active, passive, and transpassive behaviors of molybdenum in aqueous media are explained. The major surface oxide species observed, reversible redox transitions of the surface oxides, pseudocapacitance and catalytic reduction are discussed along with carefully conducted experimental results on a typical molybdenum glass back contact employed in CIGS-based solar cells. The applications of molybdenum oxides and the electrodeposition of molybdenum are briefly reviewed. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Inverse magnetocaloric effect in Ce(Fe{sub 0.96}Ru{sub 0.04}){sub 2}: Effect of fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dube, V.; Mishra, P. K.; Prajapat, C. L.; Singh, M. R.; Ravikumar, G. [Technical Physics Division, Bhabha Atomic Research Centre, Mumbai-4000085 (India); Rajarajan, A. K.; Sastry, P. U. [Solid State Physics Division, Bhabha Atomic Research Centre, Mumbai-4000085 (India); Thakare, S. V. [Radio Pharmaceuticals Division, Bhabha Atomic Research Centre, Mumbai-4000085 (India)

    2013-02-05

    We have shown the effect of fast neutron irradiation on the magnetic phase transition and magnetocaloric effect (MCE) in a doped Ce(Fe{sub 0.96}Ru{sub 0.04}){sub 2}, intermettalic. We show that this leads to suppression of MCE and a to a disordered ferromagnetic phase.

  17. Defects annihilation behavior of neutron-irradiated SiC ceramics densified by liquid-phase-assisted method after post-irradiation annealing

    National Research Council Canada - National Science Library

    Idris, Mohd Idzat; Yamazaki, Saishun; Yoshida, Katsumi; Yano, Toyohiko

    2016-01-01

    Numerous studies on the recovery behavior of neutron-irradiated high-purity SiC have shown that most of the defects present in it are annihilated by post-irradiation annealing, if the neutron fluence is less than 1×1026 n/m2 (>0.1MeV...

  18. Inorganic polarography in organic solvents-II: polarographic examination of the molybdenum(V) thiocyanate complex in diethyl ether.

    Science.gov (United States)

    Afghan, B K; Dagnall, R M

    1967-02-01

    A procedure involving the solvent extraction of molybdenum(V) thiocyanate into diethyl ether followed by a direct polarographic examination of the organic phase offers a selective method for the determination of molybdenum down to 0.5 ppm. Only molybdenum, amongst 21 elements examined, is observed to give a reduction wave under the recommended conditions. The method is evaluated with respect to various experimental factors and is applied to the determination of molybdenum in mild and alloy steels.

  19. A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-06-15

    Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

  20. Improvement of Measurement of Silicon Content in Zinc and Zinc Alloy by Molybdenum Blue Spectrophotometric Method%钼蓝分光光度法测锌及锌合金中硅量方法改进

    Institute of Scientific and Technical Information of China (English)

    刘梦影

    2015-01-01

    The paper analyzes molybdenum blue spectrophotometric method that is used to measure silicon content in zinc and zinc alloy, meanwhile, researches some shortcomings and gives some supplements and advices. The new method uses matrix matching to correct standard curve and the blank by eliminating the matrix interference. Color range pH5.6 ~ 7.6 is used on nitrophenol to adjust pH. Experiments show that silicon and ammonium molybdate react to generate silicon molybdenum heteropoly acid in sulfuric acid medium by using ascorbic acid as reducing agent. The concentration of silicon standard solution within 0 ~ 1.0 μg/mL conforms to beer's law, standard recovery is 98% ~ 101%, which is better than the national standard method and broaden the scope of determination of silicon content.%对国标中用于测定锌及锌合金中硅含量的钼蓝光度法进行了分析,研究了其中的不足,并作出了几点补充和建议.新方法采用基体匹配方式校正标准曲线及空白,消除基体干扰;使用变色范围pH5.6~7.6的对硝基苯酚调pH.实验表明:在硫酸介质中,硅与钼酸铵生成硅钼杂多酸,用抗坏血酸作还原剂,硅标准溶液浓度在0~1.0 μg/mL范围内符合比尔定律,标准回收率为98%~101%,较国标方法更理想,拓宽了硅量测定范围.

  1. Effect of 16.3 dpa neutron irradiation on fatigue lifetime of the RAFM steel EUROFER97

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A.; Rolli, R.; Schneider, H.-C. [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-15

    Low cycle fatigue specimens of the reduced-activation martensitic/ferritic steel EUROFER97 were neutron irradiated at 250 deg. C up to an accumulated dose of 16.3 dpa. After irradiation, the specimens were push-pull fatigue tested under strain-controlled conditions at 250 deg. C to determine the impact of irradiation on lifetime, fracture behavior, and microstructure. The typical cyclic softening of martensitic/ferritic steels was observed. Furthermore, a considerable increase of lifetime after irradiation and subsequent cycling at lower strain amplitudes was remarkable. This behavior was attributed to the homogeneous distribution of stable irradiation-induced dislocation loops and small precipitates acting as barriers for the cyclic motion of dislocations, thereby influencing substantially crack initiation and crack network formation. While in the un-irradiated material push-pull fatigue sweeps the dislocations to the boundaries, a significant fraction of dislocations was fixed at irradiation-induced defects after irradiation and fatigue testing.

  2. Effects of neutron irradiation on dimensional stability and on mechanical properties of SiC/SiC composites

    Energy Technology Data Exchange (ETDEWEB)

    Youngblood, G.E.; Henager, C.H. Jr.; Senor, J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1995-04-01

    The objective of this work is to assess the development and the performance of continuous fiber SiC{sub f}/SiC composites as a structural material for advanced fusion reactor application. The dimensional stability and some mechanical properties of two similar 2D 0-90{degree} weave SiC{sub f}/SiC composites made with Nacalon{trademark} ceramic-grade fiber were characterized and compared after neutron irradiation to those properties for {beta}-SiC. The major difference between these two composites was that one had a thin (150 nm) and the other a thick (1000 nm) graphite interface layer. The irradiation conditions consisted of relatively high doses (4.3 to 26 dpa-SiC) at high temperature (430-1200{degree}C).

  3. High-dose neutron irradiation of Hi-Nicalon Type S silicon carbide composites. Part 2: Mechanical and physical properties

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai, E-mail: katohy@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Nozawa, Takashi [Japan Atomic Energy Agency, Rokkasho, Aomori-ken (Japan); Shih, Chunghao [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Ozawa, Kazumi [Japan Atomic Energy Agency, Rokkasho, Aomori-ken (Japan); Koyanagi, Takaaki; Porter, Wally; Snead, Lance L. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    2015-07-15

    Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. The material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. The observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

  4. EPR study of gamma and neutron irradiation effects on KU1, KS-4V and Infrasil 301 silica glasses

    Energy Technology Data Exchange (ETDEWEB)

    Lagomacini, Juan C., E-mail: jc.lagomacini@uam.es [Dept. Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain); Bravo, David [Dept. Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain); Leon, Monica; Martin, Piedad; Ibarra, Angel [Materiales para Fusion, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain); Martin, Agustin [Dept. Fisica e Instalaciones, ETS Arquitectura UPM, E-28040 Madrid (Spain); Lopez, Fernando J. [Dept. Fisica de Materiales, Universidad Autonoma de Madrid, E-28049 Madrid (Spain)

    2011-10-01

    Electron paramagnetic resonance (EPR) studies have been carried out on KU1 and KS-4V high purity quartz glasses and commercial silica Infrasil 301, irradiated with gamma rays up to a dose of 11.6 MGy and neutron fluences of 10{sup 21} and 10{sup 22} n/m{sup 2}. Gamma irradiations produce a much higher concentration of defect centres (mainly E', POR and NBOHC) for KU1 and I301 than for KS-4V silica. In contrast, neutron irradiation at the highest fluence produces similar concentrations in all silica types. These results agree to a good extent with those obtained in previous optical absorption measurements. Moreover, oxygen-related centres (POR and NBOHC) have been well characterized by means of electron paramagnetic resonance.

  5. Scoping of material response under DEMO neutron irradiation: comparison with fission and influence of nuclear library selection

    CERN Document Server

    Gilbert, M R

    2016-01-01

    Predictions of material activation inventories will be a key input to virtually all aspects of the operation, safety and environmental assessment of future fusion nuclear plants. Additionally, the neutron-induced transmutation (change) of material composition (inventory) with time, and the creation and evolution of configurational damage from atomic displacements, require precise quantification because they can lead to significant changes in material properties, and thus influence reactor-component lifetime. A comprehensive scoping study has been performed to quantify the activation, transmutation (depletion and build-up) and immediate damage response under neutron irradiation for all naturally occurring elements from hydrogen to bismuth. The resulting database provides a global picture of the response of a material, covering the majority of nuclear technological space, but focussing specifically on typical conditions expected for a demonstration fusion power plant (DEMO). Results from fusion are compared aga...

  6. Tritium and helium release from beryllium pebbles neutron-irradiated up to 230appm tritium and 3000appm helium

    Directory of Open Access Journals (Sweden)

    Vladimir Chakin

    2016-12-01

    Full Text Available Study of tritium and helium release from beryllium pebbles with diameters of 0.5 and 1mm after high-dose neutron irradiation at temperatures of 686–968K was performed. The release rate always has a single peak, and the peak temperatures at heating rates of 0.017K/s and 0.117K/s lie in the range of 1100–1350K for both tritium and helium release. The total tritium release from 1mm pebbles decreases considerably by increasing the irradiation temperature. The total tritium release from 0.5mm pebbles is less than that from 1mm pebbles and remains constant regardless of the irradiation temperature. At high irradiation temperatures, open channels are formed which contribute to the enhanced tritium release.

  7. Subtask 12F2: Microstructural evolution of V-4Cr-4Ti during neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Gazda, J.; Loomis, B.A. [Argonne National Lab., IL (United States)

    1995-03-01

    The objective of this work is to characterize the microstructural evolution of V-4Cr-4Ti alloy during irradiation by fast neutrons, and thereby to provide a better understanding of long-term performance of the alloy under fusion conditions. Microstructural evolution of V-4Cr-4Ti, an alloy recently shown to exhibit excellent tensile and creep properties, virtual immunity to irradiation embrittlement, and good resistance to swelling, was characterized after irradiation in a lithium environment in the Fast Flux Test Facility (FFTF) (a sodium-cooled fast reactor located in Richland, Washington) at 420, 520, and 600{degrees}C to 24-34 dpa. The primary feature of microstructural evolution during irradiation at 520 and 600{degrees}C was high-density formation of ultrafine Ti{sub 5}Si{sub 3} precipitates and short dislocations. For irradiation at 420{degrees}C, precipitation of Ti{sub 5}Si{sub 3} was negligible, and {open_quotes}black-dot{close_quotes} defects and dislocations were observed in significantly higher densities. In spite of their extremely high densities, neither the {open_quotes}black-dot{close_quotes} defects nor Ti{sub 5}Si{sub 3} precipitates are overly detrimental to ductility and toughness of the alloy, yet they very effectively suppress irradiation-induced swelling. Therefore, these features, normally observed in V-base alloys containing Ti and Si, are considered stable. Unstable microstructural modifications that are likely to degrade mechanical properties significantly were not observed, e.g., irradiation-induced formation of fine oxides, carbides, nitrides, or Cr-rich clusters. 18 refs., 4 figs., 1 tab.

  8. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Science.gov (United States)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  9. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  10. Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Masashi Shimada; Y. Hatano; Y. Oya; T. Oda; M. Hara; G. Cao; M. Kobayashi; M. Sokolov; H. Watanabe; B. Tyburska; Y. Ueda; P. Calderoni

    2011-09-01

    Plasma-facing components (PFCs) will be exposed to 14 MeV neutrons from deuterium-tritium (D-T) fusion reactions, and tungsten, a candidate PFC for the divertor in ITER, is expected to receive a neutron dose of 0.7 displacement per atom (dpa) by the end of operation in ITER. The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in PFCs, it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high PKA energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influence the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron irradiation. Therefore, the effort to correlate among high-energy ions, fission neutrons, and fusion neutrons is crucial for accurately estimating tritium retention under a neutron-irradiation environment. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at. % purity from A.L.M.T. Co.) were irradiated by neutron in the High Flux Isotope Reactor (HFIR), ORNL, at 50 and 300C to 0.025, 0.3, and 1.2 dpa, and the investigation of deuterium retention in neutron-irradiation was performed in the INL Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe2+, 20 MeV W2+, and 700 keV H-) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

  11. Interaction processes between vacancies and dislocations in molybdenum in the temperature range around 0.3 of the melting temperature

    Energy Technology Data Exchange (ETDEWEB)

    Zelada-Lambri, G.I. [Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avenida Pellegrini 250, 2000 Rosario (Argentina); Lambri, O.A. [Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Laboratorio de Materiales, Escuela de Ingenieria Electrica, Avenida Pellegrini 250, 2000 Rosario (Argentina); Instituto de Fisica Rosario, Member of the CONICET' s Research Staff (Argentina)], E-mail: olambri@fceia.unr.edu.ar; Bozzano, P.B. [Laboratorio de Microscopia Electronica, Unidad de Actividad Materiales, Centro Atomico Constituyentes, Comision Nacional de Energia Atomica, Avenida General Paz 1499, 1650 San Martin (Argentina); Garcia, J.A. [Departamento de Fisica Aplicada II, Facultad de Ciencias y Tecnologia, Universidad del Pais Vasco, Apartado 644, 48080 Bilbao, Pais Vasco (Spain); Celauro, C.A. [Reactor Nuclear RA-4, Facultad de Ciencias Exactas, Ingenieria y Agrimensura, Universidad Nacional de Rosario, Riobamba y Berruti, 2000 Rosario (Argentina)

    2008-10-15

    Mechanical spectroscopy, electrical resistivity and transmission electron microscopy studies have been performed on pre-strained neutron irradiated single crystalline molybdenum in order to check the interaction processes between vacancies and dislocations in the temperature range between room temperature and 1273 K. The anelastic relaxation in molybdenum which appears between 800 K and 1273 K has been separated in two different physical mechanisms depending on the temperature of appearance of the relaxation peak. The physical mechanism which controls the damping peak appearing at around 800 K was related with the dragging of jogs by the dislocation under movement assisted by vacancy diffusion. The damping peak which appears at higher temperatures of about 1000 K was more consistent with the formation and diffusion of vacancies assisted by the dislocation movement.

  12. Final report on characterization of physical and mechanical properties of copper and copper alloys before and after irradiation

    DEFF Research Database (Denmark)

    Singh, B.N.; Tähtinen, S.

    2002-01-01

    The present report summarizes and highlights the main results of the work carried out during the last 5-6 years on effects of neutron irradiation on physical and mechanical properties of copper and copper alloys. The work was an European contribution toITER Research and Development programme...... amount of further effort is needed to find a realistic and optimum solution....

  13. Final Report on investigations of the influence of helium concentration and implantation rate on cavity nucleation and growth during neutron irradiation of Fe and EUROFER 97

    Energy Technology Data Exchange (ETDEWEB)

    Eldrup, M.; Singh, B.N. (Risoe DTU, Materials Research Div., Roskilde (Denmark)); Golubov, S. (Materials Science and Technology Div., Oak Ridge National Lab., Oak Ridge (United States))

    2010-09-15

    This report presents results of investigations of damage accumulation during neutron irradiation of pure iron and EUROFER 97 steel with or without prior helium implantation. The defect microstructure, in particular the cavities, was characterized using Positron Annihilation Spectroscopy (PAS) and Transmission Electron Microscopy (TEM). The PAS investigations revealed a clear difference between the He implantation effects in Fe and EUROFER 97 at 623 K. For both materials the mean positron lifetimes increased with He dose in the range 1-100 appm, although the increase was stronger for Fe than for EUROFER 97 and for both materials smaller for implantation at 623 K than at 323 K. This lifetime increase is due primarily to the formation of He bubbles. For He doses of 10-100 appm cavity sizes and densities in Fe were estimated to be 1.7-2.8 nm and 4-14 x 10{sup 21} m{sup -3}, respectively. Neutron irradiation after He implantation in general leads to an increase of both cavity sizes and densities. Estimates of cavity sizes and densities in EUROFER 97 after neutron irradiation with or without prior helium implantation are rather uncertain, but lead to values of the same order as for iron. TEM cannot resolve any cavities in Fe or EUROFER 97 after implantation of 100 appm He neither at 323 K nor at 623 K. However, neutron irradiation at 623 K to a dose level of 0.23 dpa in Fe is observed to lead to cavities with sizes of about 4 nm and densities of about 1.5 x 10{sup 21} m{sup -3}. He implantation (100 appm) prior to neutron irradiation results in a cavity density increase to {approx} 1 x 10{sup 22} m{sup -3}. In EUROFER 97 a very inhomogeneous cavity distribution, formed at dislocations and interfaces, is observed after He implantation with subsequent neutron irradiation. In addition, a very low density of very large voids have been observed in Fe (without He) neutron irradiated at 323 K, already at a dose level of 0.036 dpa. Detailed numerical calculations within the

  14. [Radioactivity of phosphorus implanted TiNi alloy].

    Science.gov (United States)

    Zhao, Xingke; Cai, Wei; Zhao, Liancheng

    2003-09-01

    Exposed to neutron flow, the phosphorus implanted TiNi alloy gets radioactive. This radioactive material is used in vascular stent for prevention and cure of restenosis. Phosphorus implantation is carried out in a plasma immerged ion implantation system, and the dose of phosphorus implantation is in the range of 2-10 x 10(17) cm-2. After ion implantation, the alloy is exposed to the slow neutron flow in a nuclear reactor, the dose of the slow neutron is 1.39-5.88 x 10(19) n/cm2. The radioactivity of the TiNi alloy was measured by liquid scintillation spectrometry and radio-chromic-film dosimetry. The result shows that whether the phosphorus is implanted or not, the TiNi alloy comes to be radioactive after exposure to neutron flow. Just after neutron irradiation, the radiation dose of phosphorus implanted TiNi alloy is about one hundred times higher than that of un-phosphorus implanted TiNi alloy. The radiation difference between phosphorus and un-phosphorus implanted alloy decreases as time elapses. Within three months after neutron irradiation, the average half-decay period of phosphorus implanted TiNi alloy is about 62 days. The radiation ray penetration of phosphorus implanted TiNi alloy is deeper than that of pure 32P; this is of benefit to making radiation uniformity between stent struts and reducing radiation grads beyond the edge of stent.

  15. Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    Science.gov (United States)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

  16. Positron Annihilation Lifetime Spectroscopy Study of Neutron Irradiated High Temperature Superconductors YBa2Cu3O7-δ for Application in Fusion Facilities

    Science.gov (United States)

    Veterníková, J.; Chudý, M.; Slugeň, V.; Eisterer, M.; Weber, H. W.; Sojak, S.; Petriska, M.; Hinca, R.; Degmová, J.; Sabelová, V.

    2012-02-01

    This study focuses on the crystallographic defects introduced by neutron irradiation and the resulting changes of the superconducting properties in the high temperature superconductor YBa2Cu3O7-δ. This material is considered to be most promising for magnet systems in future fusion reactors. Two different bulk samples, pure non-doped YBa2Cu3O7-δ (YBCO) and multi-seed YBa2Cu3O7-δ doped by platinum (MS2F) were studied prior to and after irradiation in the TRIGA MARK II reactor in Vienna. Neutron irradiation is responsible for a significant enhancement of the critical current densities as well as for a reduction in critical temperature. The accumulation of small open volume defects (treatment.

  17. Study on neutron irradiation effect of superconductors and installation of 15.5 T magnet in hot laboratory at IMR in Tohoku University

    Science.gov (United States)

    Nishimura, Arata; Takeuchi, Takao; Nishijima, Shigehiro; Ochiai, Kentaro; Nishijima, Gen; Watanabe, Kazuo; Narui, Minoru; Kurishita, Hiroaki; Shikama, Tatsuo

    2011-10-01

    A fusion reactor will yield a lot of high energy neutrons, and some of them will stream out of a plasma vacuum vessel and penetrate a blanket system and reach superconducting magnets which provide high magnetic field to confine high energy ionized particles. Under the neutron irradiation, the magnet materials will be activated and the properties will change. In this study, a research network on study of the irradiation effects is introduced and some data recently obtained are presented. By neutron irradiation, the critical current of the Nb 3Sn wire increased and the critical field did not change. The organic insulation materials were degraded at neutron fluence of 1 × 10 22 n/m 2. In addition, the outline of an installation plan of 15.5 T superconducting magnet into radiation control area is described and study issues are explained. The project is undergoing as a program of Nuclear Basic Infrastructure Strategic Study Initiative.

  18. Observation of the crossover from two-gap to single-gap superconductivity through specific heat measurements in neutron-irradiated MgB2.

    Science.gov (United States)

    Putti, M; Affronte, M; Ferdeghini, C; Manfrinetti, P; Tarantini, C; Lehmann, E

    2006-02-24

    We report specific heat measurements on neutron-irradiated MgB2 samples, for which the critical temperature is lowered to 8.7 K, but the superconducting transition remains extremely sharp, indicative of a defect structure extremely homogeneous. Our results evidence the presence of two superconducting gaps in the temperature range above 21 K, while single-gap superconductivity is well established as a bulk property, not associated with local disorder fluctuations, when Tc decreases to 11 K.

  19. Study of boron carbide evolution under neutron irradiation; Contribution a l'etude de l'evolution du carbure de bore sous irradiation neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Simeone, D. [CEA/Saclay, Dept. de Mecanique et de Technologie (DMT), 91 - Gif-sur-Yvette (France)]|[Universite Blaise Pascal, Clermont-Ferrand II, (CNRS), 63 - Aubiere (France)

    1999-07-01

    Owing to its high neutron efficiency, boron carbide (B{sub 4}C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the {sup 10}B(n,{alpha}){sup 7}Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B{sub 4}C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the {sup 7}Li(p,p'{gamma}){sup 7}Li reaction, allows to measure lithium profile in B{sub 4}C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B{sub 4}C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B{sub 4}C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B{sub 4}C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B{sub 4}C. (author)

  20. RECYCLING TECHNOLOGY INTO INDUSTRIAL TURNOVER OF BISMUTH AND MOLYBDENUM FROM DEAD CATALYST

    Directory of Open Access Journals (Sweden)

    O. S. Komarov

    2013-01-01

    Full Text Available The technology of separate extraction of bismuth and molybdenum from spent catalyst was presented and information on the effectiveness of its use in a composition of comprehensive modifier in the iron-carbon alloy was given.

  1. Research and development on vanadium alloys for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Matsui, H.; Abe, K. [Tohoku Univ. (Japan); Smith, D.L. [Argonne National Lab., IL (United States); Osch, E. van [NERF, Petten (Netherlands); Kazakov, V.A. [RIAR, Dimitrovgrad (Russian Federation)

    1998-03-01

    The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

  2. A facility for fast-neutron irradiations at Jyväskylä and its use for nuclide cross-section measurements in fission

    Science.gov (United States)

    Lhersonneau, G.; Malkiewicz, T.; Jones, P.; Karvonen, P.; Ketelhut, S.; Bajeat, O.; Fadil, M.; Gaudu, S.; Saint-Laurent, M. G.; Trzaska, W. H.

    2013-01-01

    An efficient and reliable transport system for fast-neutron irradiations has been built at the Physics Department, Jyväskylä, Finland. It is constructed from commercial bicycle components and is driven by a computer-controlled stepping motor. It can be operated in single or cyclic mode. The neutron irradiated targets are moved within 1.2 s (full stop to full stop) to a well-shielded position 3 m away where they can be removed or directly investigated by γ spectroscopy. The system has been built with the aim to experimentally verify the calculated production rates of neutron-rich nuclei in the SPIRAL2 uranium target. However, the facility can be used for various kinds of fast-neutron irradiations, with a neutron spectrum up to 60 MeV produced by stopping a deuteron beam of several μA in a thick target. Examples of applications are activation and integral cross-section measurements, evaluation of damages in materials and biological cells.

  3. Ionizing/displacement synergistic effects induced by gamma and neutron irradiation in gate-controlled lateral PNP bipolar transistors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chenhui, E-mail: wangchenhui@nint.ac.cn [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O. Box 69-10, Xi’an 710024 (China); Chen, Wei; Yao, Zhibin; Jin, Xiaoming; Liu, Yan; Yang, Shanchao [State Key Laboratory of Intense Pulsed Irradiation Simulation and Effect, Northwest Institute of Nuclear Technology, P.O. Box 69-10, Xi’an 710024 (China); Wang, Zhikuan [State Key Laboratory of Analog Integrated Circuit, Chongqing 400060 (China)

    2016-09-21

    A kind of gate-controlled lateral PNP bipolar transistor has been specially designed to do experimental validations and studies on the ionizing/displacement synergistic effects in the lateral PNP bipolar transistor. The individual and mixed irradiation experiments of gamma rays and neutrons are accomplished on the transistors. The common emitter current gain, gate sweep characteristics and sub-threshold sweep characteristics are measured after each exposure. The results indicate that under the sequential irradiation of gamma rays and neutrons, the response of the gate-controlled lateral PNP bipolar transistor does exhibit ionizing/displacement synergistic effects and base current degradation is more severe than the simple artificial sum of those under the individual gamma and neutron irradiation. Enough attention should be paid to this phenomenon in radiation damage evaluation. - Highlights: • A kind of gate-controlled lateral PNP bipolar transistor has been specially designed to facilitate the analysis of ionizing/displacement synergistic effects induced by the mixed irradiation of gamma and neutron. • The difference between ionizing/displacement synergistic effects and the simple sum of TID and displacement effects is analyzed. • The physical mechanisms of synergistic effects are explained.

  4. Modification of chemical, optical and structural properties of Bayfol CR-6-2 using gamma and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shehata, Mohamed M.; Radwan, Samh I.; Hassan, Amin [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Centre; Waly, Sayed A. [Atomic Energy Authority, Cairo (Egypt). Second Research Reactor; Badawy, Zaynab M. [Atomic Energy Authority, Cairo (Egypt). Experimental Nuclear Physics Dept.

    2016-08-01

    The effects of gamma and neutron irradiations on the chemical, optical and structural properties of Bayfol CR-6-2 were investigated. The samples were irradiated by γ-rays from a {sup 60}Co source at various doses ranging between 16 and 900 kGy at room temperature in atmospheric air. For neutrons, an Am-Be neutron facility was used for the sample irradiation in thermal mode which had an activity of 185 GBq. Samples were irradiated with different doses of neutrons ranging from 15.7 to 564.2 mGy. The changes induced were analyzed using UV-Vis and Fourier transform infrared (FTIR) spectrometry. The results demonstrated an occurrence of oxidative degradation, resulting in the formation of carbonyl groups at 1700 cm{sup -1}. Simultaneous thermo-gravimetric investigation (TGA) has been performed on the samples of 0.3 mm thickness. The results obtained indicate that cross-linking predominates at small neutron doses and main chain scission happens at higher doses.

  5. Irradiation-induced precipitates in a neutron irradiated 304 stainless steel studied by three-dimensional atom probe

    Energy Technology Data Exchange (ETDEWEB)

    Toyama, T., E-mail: ttoyama@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Nozawa, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Van Renterghem, W. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Matsukawa, Y.; Hatakeyama, M.; Nagai, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Al Mazouzi, A. [EDF R and D, Avenue des Renardieres Ecuelles, 77818 Moret sur Loing Cedex (France); Van Dyck, S. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium)

    2011-11-15

    Highlights: > Irradiation-induced precipitates in a 304 stainless steel were investigated by three-dimensional atom probe. > The precipitates were found to be {gamma}' precipitates (Ni{sub 3}Si). > Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening. - Abstract: Irradiation-induced precipitates in a 304 stainless steel, neutron-irradiated to a dose of 24 dpa at 300 deg. C in the fuel wrapper plates of a commercial pressurized water reactor, were investigated by laser-assisted three-dimensional atom probe. A high number density of 4 x 10{sup 23} m{sup -3} of Ni-Si rich precipitates was observed, which is one order of magnitude higher than that of Frank loops. The average diameter was {approx}10 nm and the average chemical composition was 40% Ni, 14% Si, 11% Cr and 32% Fe in atomic percent. Over a range of Si concentrations, the ratio of Ni to Si was {approx}3, close to that of {gamma}' precipitate (Ni{sub 3}Si). In some precipitates, Mn enrichment inside the precipitate and P segregation at the interface were observed. Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening.

  6. Development of indigenous insulation material for superconducting magnets and study of its characteristics under influence of intense neutron irradiation

    Science.gov (United States)

    Sharma, Rajiv; Tanna, V. L.; Rao, C. V. S.; Abhangi, Mitul; Vala, Sudhirsinh; Sundaravel; Varatharajan, S.; Sivakumar, S.; Sasi, K.; Pradhan, S.

    2017-02-01

    Epoxy based glass fiber reinforced composites are the main insulation system for the superconducting magnets of fusion machines. 14MeV neutrons are generated during the DT fusion process, however the energy spectra and flux gets modified to a great extent when they reach the superconducting magnets. Mechanical properties of the GFRP insulation material is reported to degrade up to 30%. As a part of R & D activity, a joint collaboration with IGCAR, Kalpakkam has been established. The indigenous insulation material is subjected to fast neutron fluence of 1014 - 1019 n/m2 (E>0.1 MeV) in FBTR and KAMINI Reactor, India. TRIM software has been used to simulate similar kind of damage produced by neutrons by ion irradiation with 5 MeV Al ions and 3 MeV protons. Fluence of the ions was adjusted to get the same dpa. We present the test experiment of neutron irradiation of the composite material (E-glass, S-glass fiber boron free and DGEBA epoxy). The test results of tensile, inter laminar shear and electrical breakdown strength as per ASTM standards, assessment of micro-structure surface degradation before and after irradiation will be presented. MCNP simulations are carried out for neutron flux, dose and damages produced in the insulation material.

  7. Spectrum evaluation at the filter-modified neutron irradiation field for neutron capture therapy in Kyoto University Research Reactor

    Science.gov (United States)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2004-10-01

    The Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor (KUR-HWNIF) was updated in March 1996, mainly to improve the facility for neutron capture therapy (NCT). In this facility, neutron beams with various energy spectra, from almost pure thermal to epithermal, are available. The evaluation of the neutron energy spectra by multi-activation-foil method was performed as a series of the facility characterization. The spectra at the normal irradiation position were evaluated for the combinations of heavy-water thickness of the spectrum shifter and the open-close condition of the cadmium and boral filters. The initial spectra were made mainly using a two-dimensional transport code, and the final spectra were obtained using an adjusting code. For the verification of the evaluated spectra, simulation calculations using a phantom were performed on the assumption of NCT-clinical-irradiation conditions. It resulted that the calculated data for the depth neutron-flux distributions were in good agreement with the experimental ones.

  8. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT.

    Science.gov (United States)

    Evans, J F; Blue, T E

    1996-11-01

    Protecting the facility personnel and the general public from radiation exposure is a primary safety concern of an accelerator-based epithermal neutron irradiation facility. This work makes an attempt at answering the questions "How much?" and "What kind?" of shielding will meet the occupational limits of such a facility. Shielding effectiveness is compared for ordinary and barytes concretes in combination with and without borated polyethylene. A calculational model was developed of a treatment room , patient "scatterer," and the epithermal neutron beam. The Monte Carlo code, MCNP, was used to compute the total effective dose equivalent rates at specific points of interest outside of the treatment room. A conservative occupational effective dose rate limit of 0.01 mSv h-1 was the guideline for this study. Conservative Monte Carlo calculations show that constructing the treatment room walls with 1.5 m of ordinary concrete, 1.2 m of barytes concrete, 1.0 m of ordinary concrete preceded by 10 cm of 5% boron-polyethylene, or 0.8 m of barytes concrete preceded by 10 cm of 5% boron-polyethylene will adequately protect facility personnel.

  9. Pixel pitch and particle energy influence on the dark current distribution of neutron irradiated CMOS image sensors.

    Science.gov (United States)

    Belloir, Jean-Marc; Goiffon, Vincent; Virmontois, Cédric; Raine, Mélanie; Paillet, Philippe; Duhamel, Olivier; Gaillardin, Marc; Molina, Romain; Magnan, Pierre; Gilard, Olivier

    2016-02-22

    The dark current produced by neutron irradiation in CMOS Image Sensors (CIS) is investigated. Several CIS with different photodiode types and pixel pitches are irradiated with various neutron energies and fluences to study the influence of each of these optical detector and irradiation parameters on the dark current distribution. An empirical model is tested on the experimental data and validated on all the irradiated optical imagers. This model is able to describe all the presented dark current distributions with no parameter variation for neutron energies of 14 MeV or higher, regardless of the optical detector and irradiation characteristics. For energies below 1 MeV, it is shown that a single parameter has to be adjusted because of the lower mean damage energy per nuclear interaction. This model and these conclusions can be transposed to any silicon based solid-state optical imagers such as CIS or Charged Coupled Devices (CCD). This work can also be used when designing an optical imager instrument, to anticipate the dark current increase or to choose a mitigation technique.

  10. RBE values for colo-rectal injury after caesium 137 gamma-ray and neutron irradiation. 1. Single doses

    Energy Technology Data Exchange (ETDEWEB)

    Terry, N.H.A.; Denekamp, J.; Maughan, R.L. (Mount Vernon Hospital, Northwood (UK). Gray Lab.)

    1983-04-01

    Colo-rectal damage in mice has been assessed after caesium ..gamma.. irradiation and 3 MeV neutrons given as single doses. Several assays were used, including body weight changes, faecal deformity and lethality. Dose response curves were constructed for each assay at times ranging from 10 days to 16 months after irradiation. An initial loss of weight at 10-20 days was presumably related to epithelial denudation, but a dose-dependent weight reduction (compared with controls) persisted over the animals' life span. Mice died progressively after localised pelvic ..gamma.. irradiation; there was no sharp demarcation between an early and late phase of lethal injury. Death resulted from intestinal stricture or stenosis. The time course for lethality was qualitatively different after neutron irradiation, with little progression of damage between 5 and 11 months. Faecal deformity was detectable as a higher proportion of small pellets when the rectum became constricted by fibrosis. No significant faecal deformity was observed before 6 months after which time dose response curves could be obtained. The RBE for early damage (assessed at 1-3 months) was 2.2-2.7, falling to 1.7-1.9 for late damage (determined at 10-15 months) over the range of neutron doses of 7.5-12 Gy.

  11. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  12. Effects of dose and dose protraction on embryotoxicity of 14.1 MeV neutron irradiation in rats

    Energy Technology Data Exchange (ETDEWEB)

    Beckman, D.A.; Buck, S.J. [Alfred I. duPont Institute, Wilmington, DE (United States)]|[Thomas Jefferson Univ., Philadelphia, PA (United States); Solomon, H.M. [SmithKline and Beecham Pharmaceuticals, King of Prussia, PA (United States); Gorson, R.O. [Thomas Jefferson Univ., Philadelphia, PA (United States); Mills, R.E. [Brookhaven National Lab., Upton, NY (United States); Brent, R.L. [Alfred I. duPont Institute, Wilmington, DE (United States)]|[Thomas Jefferson Univ., Philadelphia, PA (United States)

    1994-06-01

    The embryotoxic effects of neutron radiation on rodent embryos are documented, but there is disagreement about the dose-response relationship and the impact of protracting the dose. Pregnant rats were exposed to total absorbed doses of 0.15 to 1.50 Gy 14.1 MeV neutrons on day 9.5 after conception, coincident with the most sensitive stage of embryonic development for the induction of major congenital malformations. In general terms, the incidence of embryotoxic effects increased with increasing total absorbed dose. However, the dose-response relationship differed depending on the parameter of embryotoxicity chosen, namely, intrauterine death, malformations or very low body weight. In a second study, embryos were exposed to a single embryotoxic absorbed dose (0.75 Gy) administered at a range of dose rates, from 0.10 to 0.50 Gy/h. The results offer no evidence that protraction of this selected dose significantly increased or decreased the incidence or pattern of embryotoxicity of the neutron exposure used in this study. The results do not support the hypothesis of a linear dose-response relationship for the effects of prenatal neutron irradiation that contribute to embryotoxicity for total absorbed doses of 0.15 to 1.50 Gy. 23 refs., 8 tabs.

  13. Analysis of niobium alloys.

    Science.gov (United States)

    Ferraro, T A

    1968-09-01

    An ion-exchange method was applied to the analysis of synthetic mixtures representing various niobium-base alloys. The alloying elements which were separated and determined include vanadium, zirconium, hafnium, titanium, molybdenum, tungsten and tantalum. Mixtures containing zirconium or hafnium, tungsten, tantalum and niobium were separated by means of a single short column. Coupled columns were employed for the resolution of mixtures containing vanadium, zirconium or titanium, molybdenum, tungsten and niobium. The separation procedures and the methods employed for the determination of the alloying elements in their separate fractions are described.

  14. Effects of neutron irradiation on the microstructure of alpha-annealed zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Bajaj, R.; Kammenzind, B.F. [Bettis Atomic Power Lab., West Mifflin, PA (United States); Farkas, D.M. [General Electric Co., Pleasanton, CA (United States). Vallecitos Nuclear Center

    2002-07-01

    Analytical electron microscopy (AEM) was used to study the separate effects of the irradiation parameters on the evolution of the microstructure in recrystallized alpha-annealed Zircaloy-4 under controlled irradiation conditions. The effects of fast neutron flux from {approx} 4 x 10{sup 13} n/cm{sup 2}-s to {approx} 1.5 x 10{sup 14} n/cm{sup 2}-s (E > 1 MeV){sup 3} neutron fluence in the range of {approx} 15 x 10{sup 20} n/cm{sup 2} to {approx} 50 x 10{sup 20} n/cm{sup 2} and temperature from {approx} 270 to {approx} 330 deg C were studied. The completeness of the test matrix and the exposure in the controlled environment of the advanced test reactor permitted the separate effects of fast neutron flux, fluence, and irradiation temperature to be delineated for the first time. It was found that an increase in the neutron flux increases the degree of amorphization of the second-phase precipitates but retards the redistribution of iron out of the amorphous region (neutron fluence and irradiation temperature remaining the same), whereas increasing temperature (neutron flux and neutron fluence remaining the same) has a reverse effect. Overall, the rate of amorphization of the second-phase precipitates observed in this work was larger than that predicted by many existing literature models. Finally, neither segregation of alloying elements to grain boundaries nor precipitation of any new phases were encountered. (authors)

  15. Organometallic Chemistry of Molybdenum.

    Science.gov (United States)

    Lucas, C. Robert; Walsh, Kelly A.

    1987-01-01

    Suggests ways to avoid some of the problems students have learning the principles of organometallic chemistry. Provides a description of an experiment used in a third-year college chemistry laboratory on molybdenum. (TW)

  16. Numerical Simulation for Temperature Field of Laser Gas Alloying on Molybdenum Surface%脉冲激光钼表面氮化处理温度场的数值模拟

    Institute of Scientific and Technical Information of China (English)

    高昕; 苏增立

    2001-01-01

    本文用有限差分法对金属钼表面脉冲激光生成氮化钼薄膜过程的温度场进行了三维数值模拟计算。计算模型在能量平衡方程的基础上,将入射的脉冲激光在时间与空间上的分布以Gauss分布考虑,同时考虑工件尺寸、工件材料热物理性质及对流辐射造成的表面热损失等对温度场的影响。此外还从理论上计算了激光脉冲在脉冲宽度加宽后的温度场变化,分析了利用长脉冲激光进行材料表面相变硬化和激光重熔的可行性。%In this paper, numerical simulation for temperature field of Mo2N films generated on molybdenum surface using laser gas alloying(LGA) method is performed by finite-difference method. The numerical model used takes into account of Gauss distribution ‘ in space and time' of laser spot, the finite size of sample, the temperature dependence of themophysical properties, and the surface heat losses due to convection and radiation. The changes of temperature field are calculated accordingly. The feasibility of using pulse laser for laser surface transformation hardening and laser surface remelting is also analyzed.

  17. Promising Cu-Ni-Cr-Si alloy for first wall ITER applications

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, A. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Abramov, V. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rodin, M. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation)

    1996-10-01

    Precipitation-hardened Cu-Ni-Cr-Si alloy, a promising material for ITER applications, is considered. Available commercial products, chemical composition, physical and mechanical properties are presented. Embrittlement of Cu-Ni-Cr-Si alloy at 250-300 C is observed. Mechanical properties of Cu-Ni-Cr-Si alloy neutron irradiated to a dose of {proportional_to}0.2 dpa at 293 C are investigated. Embrittlement of Cu-Ni-Cr-Si alloy can be avoided by annealing. (orig.).

  18. Promising CuNi&.sbnd;CrSi alloy for first wall ITER applications

    Science.gov (United States)

    Ivanov, A.; Abramov, V.; Rodin, M.

    1996-10-01

    Precipitation-hardened CuNiCrSi alloy, a promising material for ITER applications, is considered. Available commercial products, chemical composition, physical and mechanical properties are presented. Embrittlement of CuNiCrSi alloy at 250-300°C is observed. Mechanical properties of CuNiCrSi alloy neutron irradiated to a dose of ˜0.2 dpa at 293°C are investigated. Embrittlement of CuNiCrSi alloy can be avoided by annealing.

  19. Alloy

    Science.gov (United States)

    Cabeza, Sandra; Garcés, Gerardo; Pérez, Pablo; Adeva, Paloma

    2014-07-01

    The Mg98.5Gd1Zn0.5 alloy produced by a powder metallurgy route was studied and compared with the same alloy produced by extrusion of ingots. Atomized powders were cold compacted and extruded at 623 K and 673 K (350 °C and 400 °C). The microstructure of extruded materials was characterized by α-Mg grains, and Mg3Gd and 14H-LPSO particles located at grain boundaries. Grain size decreased from 6.8 μm in the extruded ingot, down to 1.6 μm for powders extruded at 623 K (350 °C). Grain refinement resulted in an increase in mechanical properties at room and high temperatures. Moreover, at high temperatures the PM alloy showed superplasticity at high strain rates, with elongations to failure up to 700 pct.

  20. Study of damages by neutron irradiation in lithium aluminates; Estudio de danos por irradiacion neutronica en aluminatos de litio

    Energy Technology Data Exchange (ETDEWEB)

    Palacios G, O

    1999-06-01

    Lithium aluminates proposed to the production of tritium in fusion nuclear reactors, due to the thermal stability that they present as well as the behavior of the aluminium to the irradiation. As a neutron flux with profile ({approx_equal} 14 Mev) of a fusion reactor is not available. A irradiation experiment was designed in order to know the micro and nano structure damages produced by fast and thermal neutrons in two irradiation positions of the fusion nuclear reactor Triga Mark III: CT (Thermal Column) and SIFCA (System of Irradiation Fixed of Capsules). In this work samples of lithium aluminate were characterized by XRD (X-Ray Diffraction), TEM (Transmission Electron Microscopy) and SEM (Scanning Electron Microscopy). Two samples were prepared by two methods: a) coalition method and b) peroxide method. This characterization comprised original and irradiated samples. The irradiated sample amounted to 4 in total: one for each preparation method and one for each irradiation position. The object of this analysis was to correlate with the received neutron dose the damages suffered by the samples with the neutron irradiation during long periods (440 H), in their micro and nano structure aspects; in order to understand the changes as a function of the irradiation zone (with thermal and fast neutron flux) and the preparation methods of the samples and having as an antecedent the irradiation in SIFCA position by short times (2h). The obtained results are referred to the stability of {gamma} -aluminate phase, under given conditions of irradiation and defined nano structure arrangement. They also refer to the proposals of growth mechanism and nucleation of new phases. The error associated with the measurement of neutron dose is also discussed. (Author)

  1. In vivo skin leptin modulation after 14 MeV neutron irradiation: a molecular and FT-IR spectroscopic study

    Energy Technology Data Exchange (ETDEWEB)

    Cestelli Guidi, M.; Mirri, C.; Marcelli, A. [Laboratori Nazionali di Frascati - INFN, Frascati, Rome (Italy); Fratini, E.; Amendola, R. [ENEA, UT BIORAD-RAB, Rome (Italy); Licursi, V.; Negri, R. [Universita La Sapienza, Dip. Biologia e Biotecnologie ' ' Charles Darwin' ' , Rome (Italy)

    2012-09-15

    This paper discusses gene expression changes in the skin of mice treated by monoenergetic 14 MeV neutron irradiation and the possibility of monitoring the resultant lipid depletion (cross-validated by functional genomic analysis) as a marker of radiation exposure by high-resolution FT-IR (Fourier transform infrared) imaging spectroscopy. The irradiation was performed at the ENEA Frascati Neutron Generator (FNG), which is specifically dedicated to biological samples. FNG is a linear electrostatic accelerator that produces up to 1.0 x 10{sup 11} 14-MeV neutrons per second via the D-T nuclear reaction. The functional genomic approach was applied to four animals for each experimental condition (unirradiated, 0.2 Gy irradiation, or 1 Gy irradiation) 6 hours or 24 hours after exposure. Coregulation of a subclass of keratin and keratin-associated protein genes that are physically clustered in the mouse genome and functionally related to skin and hair follicle proliferation and differentiation was observed. Most of these genes are transiently upregulated at 6 h after the delivery of the lower dose delivered, and drastically downregulated at 24 h after the delivery of the dose of 1 Gy. In contrast, the gene coding for the leptin protein was consistently upregulated upon irradiation with both doses. Leptin is a key protein that regulates lipid accumulation in tissues, and its absence provokes obesity. The tissue analysis was performed by monitoring the accumulation and the distribution of skin lipids using FT-IR imaging spectroscopy. The overall picture indicates the differential modulation of key genes during epidermis homeostasis that leads to the activation of a self-renewal process at low doses of irradiation. (orig.)

  2. Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

    Science.gov (United States)

    Koyanagi, Takaaki; Katoh, Yutai

    2017-10-01

    Silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230-340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  3. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Krishna, R. [Dalton Cumbrian Facility, Dalton Nuclear Institute, The University of Manchester, Westlakes Science & Technology Park, Moor Row, Whitehaven, Cumbria, CA24 3HA (United Kingdom); Jones, A.N., E-mail: Abbie.Jones@manchester.ac.uk [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom); McDermott, L.; Marsden, B.J. [Nuclear Graphite Research Group, School of MACE, The University of Manchester, Manchester, M13 9PL (United Kingdom)

    2015-12-15

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated ‘D’peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of ‘G’ and ‘D’ in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure. - Highlights: • Irradiated graphite

  4. Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

    Science.gov (United States)

    Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

    2015-12-01

    Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

  5. Evaluation of ductile-brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Chul [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)]. E-mail: mckim@kaeri.re.kr; Oh, Yong Jun [Hanbat National University, Deogmyeong-dong, Yuseong-gu, Daejeon 305-719 (Korea, Republic of); Lee, Bong Sang [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2005-08-01

    Small punch (SP) tests were performed to evaluate the ductile-brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm x 10 mm x 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 deg C. Small punch tests were performed in the temperature range of RT to -196 deg C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T {sub SP}), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T {sub 41J}. T {sub SP} from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test ({delta}T {sub 41J}) shows a better correlation with the transition temperature shift ({delta}T {sub SP(E)}) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T {sub SP} also had a correlation with the reference temperature (T {sub 0}) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  6. Molybdenum oxide nanowires: synthesis & properties

    Directory of Open Access Journals (Sweden)

    Liqiang Mai

    2011-07-01

    Full Text Available Molybdenum oxide nanowires have been found to show promise in a diverse range of applications, ranging from electronics to energy storage and micromechanics. This review focuses on recent research on molybdenum oxide nanowires: from synthesis and device assembly to fundamental properties. The synthesis of molybdenum oxide nanowires will be reviewed, followed by a discussion of recent progress on molybdenum oxide nanowire based devices and an examination of their properties. Finally, we conclude by considering future developments.

  7. Determination of the Silicon, Vanadium, Iron, Aluminum, Nickel, Molybdenum and Chromium in Titanium Alloy by ICP-AES%ICP-AES测定钛合金中硅钒铁铝镍钼铬

    Institute of Scientific and Technical Information of China (English)

    成勇

    2012-01-01

    This paper has built an analysis method of inductively coupled plasma atomic emission spectrometry (ICP-AES) for direct and simultaneous determination of alloying elements or trace impurities of silicon, vanadium, iron, aluminum, nickel, molybdenum and chromium in the titanium alloy. The titanium alloy samples were digested completely by hydrofluoric acid and nitric acid mixed reagents and the heating conditions of the digestion reaction were controlled at room temperature or 70 ℃ water bath to avoid the volatilization loss of the element and ensure that the hydrolysis reaction of high-concentration titanium in the low acidity medium did not occur. The effect of titanium matrix and coexisting elements on the determination of the spectral interference was tested. The internal standard correction method using the yttrium as internal standard element was employed, and elemental analysis of spectral lines, internal calibration spectrum, the synchronous background correction positions and ICP spectrometer working conditions were selected preferably to effectively eliminate the physical interference resulting from titanium substrate and improve the detection precision and detection limit level. The test results of the practical application show that the detection limit is 10~27 μg/L, the background equivalent density is 5~38 μg/L, the correlation coefficient r≥0.9992, the recovery rate is 95.0%~105.0% and the RSD≤2.27%.%建立了电感耦合等离子体原子发射光谱法(ICP-AES)直接同时测定钛合金中合金元素或微量杂质硅钒铁铝镍钼铬的分析方法.采用氢氟酸和硝酸混合试剂并且在室温或70℃水浴控制加热条件下消解样品,从而避免了待测元素的挥发损失以及确保了高浓度钛基体在低酸度介质中也不会发生水解反应.试验了钛基体和共存元素对测定的光谱干扰影响,采取以钇作为内标元素的内标校正法,并且优选了待测元素分析谱线、内标校正谱

  8. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  9. Final Report on Investigations of the influence of Helium concentration and implantation rate on Cavity Nucleation and Growth during neutron irradiation of Fe and EUROFER 97

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain; Golubov, S.

    This report presents results of investigations of damage accumulation during neutron irradiation of pure iron and EUROFER 97 steel with or without prior helium implantation. The defect microstructure, in particular the cavities, was characterized using Positron Annihilation Spectroscopy (PAS......) and Transmission Electron Microscopy (TEM). The PAS investigations revealed a clear difference between the He implantation effects in Fe and EUROFER 97 at 623 K. For both materials the mean positron lifetimes increased with He dose in the range 1 – 100 appm, although the increase was stronger for Fe than...

  10. Differential RBE values obtained for mammary adenocarcinoma tumor cell subpopulations after 14. 8-MeV neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    DeWyngaert, J.K.; Leith, J.T.; Peck, R.A.; Bliven, S.F.

    1981-10-01

    For tumor cell subpopulations which were isolated from a single mouse mammary adenocarcinoma were examined for their relative sensitivities to 250-kVp x irradiation and 14.8-MeV neutron irradiation. The sublines are designated 66, 67, 4.10, and 68H and differ significantly in their biological characteristics. Exponentially growing cells were exposed at the Radiological Research Accelerator Facility (RARAF) at Brookhaven National Laboratories, Upton, NY. The purpose of these studies was to compare the response of these cell lines to ionizing radiation, for high-linear-energy-transfer radiation as well as for low. The interest of such an intercomparison lies in the fact that these different cell lines, while closely related, were biologically distinguishable. Survival curve parameters obtained by fitting the single dose-response curves to a linear-quadratic equation using linear least-squares regression analysis gave values for sublines 66, 67, 4.10, and 68H, respectively, of: ..cap alpha../sub n/ (G/sub 8//sup -1/) = 0.00. 0.150, 0.041, and 0.182; ..cap alpha../sub x/ (G/sub 8//sup -1/) = 0.672, 0.845, 0.787, and 0.709; ..beta../sub x/ (G/sub 8//sup -2/) = 0.0462, 0.0345, 0.0576, and 0.0503; and ..beta../sub n/ (G/sub 8//sup -2/) = 0.0253, 0.0000, 0.0156, and 0.0666. Different relative biological effectiveness (RBE) values were obtained for sublines 66, 67, 4.10, and 68H of 4.0, 3.6, 3.9, and 2.7 at the 50% level of survival and 2.4, 2.4, 2.2, and 2.0 at the 10% level. Sublines 67 and 68H show responses which suggest a constant RBE at low values of dose, while sublines 66 and 4.10 do not. It is felt that these data illustrate the need to consider biological information as well as microdosimetric considerations in attempts to relate celluar inactivation responses to radiation quality. Further implications of these data in relation to the dual-action model of radiation inactivation are discussed.

  11. Synchrotron VUV-UV and positron lifetime spectroscopy study of vacancy-type defects in reactor neutron-irradiated MgO·nAl2O3 (n = 2

    Directory of Open Access Journals (Sweden)

    Abu Zayed Mohammad Saliqur Rahman

    2016-12-01

    Full Text Available We investigated neutron-irradiation-induced point defects in spinel single crystals using a synchrotron VUV-UV source and positron lifetime spectroscopy. Photoexcitation (PE spectra near 230 nm and their corresponding photoluminescence (PL spectra at 475 nm were attributed to F-centers. With increasing irradiation temperature and fluence, PE efficiency and PL intensity decreased dramatically. Positron lifetimes (PLT of neutron-irradiated and non-irradiated samples were measured to identify the cation vacancies. A PLT measurement of 250 ps was obtained in a neutron-irradiated (20 K sample which is tentatively attributed to an aluminum monovacancy. Decreasing PLT with higher irradiation indicates the formation of oxygen-vacancy complex centers.

  12. Enhancement of critical current density in fast neutron irradiated melt-textured YBa[sub 2]Cu[sub 3]O[sub 7-x

    Energy Technology Data Exchange (ETDEWEB)

    Puzniak, R.; Wisniewski, A.; Baran, M.; Szymczak, H. (Polska Akademia Nauk, Warsaw (Poland). Inst. Fizyki); Zhang Pingxiang; Wang Jingrong; Zhou Lian (Northwest Inst. for Nonferrous Metal Research, Baoji, SN (China)); Pytel, K.; Pytel, B. (Institute of Atomic Energy, Otwock-Swierk (Poland))

    1993-01-01

    The critical current density in melt-textured samples of YBa[sub 2]Cu[sub 3]O[sub 7]-x obtained by the powder melting process was determined from magnetization measurements. A linear dependence between the width of the hysteresis loop and sample size was observed for both unirradiated and irradiated samples. This indicates that the critical current is circulating through the whole sample and is not disconnected by weak links, even when a magnetic field is applied in the irradiated sample. After fast neutron irradiation with fluences form 5 x 10[sup 16] to 6 x 10[sup 17] n cm[sup -2] (E > 0.5 MeV), significant enhancement of the critical current density, j[sub c], was observed. Critical current density, determined from magnetization measurements, for magnetic field perpendicular to the a-b plane j[sub c][sup ab], reaches [approx] 10[sup 5] A cm[sup -2] at 77 K in 1 T. For H parallel to the a-b plane, j[sub c][sup c] along the c-axis reaches 5 x 10[sup 3] A cm[sup -2]. An increase in the anisotropy of the critical current was observed after fast neutron irradiation in the temperature range 60-80 K. (Author).

  13. Quantifying interference of krypton produced from neutron irradiation of inclusion-hosted and lattice-coordinated bromine with 40Ar/39Ar geochronology

    Science.gov (United States)

    Rutte, Daniel; Becker, Tim A.; Renne, Paul R.

    2017-08-01

    Various interfering reactions producing Ar isotopes during neutron irradiation from Cl, Ar, K, and Ca have been previously detailed with the significant ones being routinely corrected for in 40Ar/39Ar geochronology. Interference of double charged 80Kr (80Kr++) with 40Ar has not yet been considered. Significant amounts of 80Kr are produced during neutron irradiation through the reaction 79Br(n,β-)80Kr. While previous workers reported a computed fission spectrum averaged cross section of ∼48 mb-compared to ∼113 mb for 39K(n,p)39Ar-we determined a ∼33-fold higher production rate of 80Kr from Br compared to 39Ar from K in the CLICIT facility of the OSU reactor. Low-K, high-Br phases, e.g., some amphiboles, and fluid or melt inclusion rich samples are prone to 80Kr++ interference biasing 40Ar/39Ar dates in the ‰ to % range. 80Kr++ resembles excess 40Ar in step-heating experiments. The interference can be corrected for by using the parallel reaction 81Br(n,β-)82Kr.

  14. The medical-irradiation characteristics for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

    Science.gov (United States)

    Sakurai, Yoshinori; Kobayashi, Tooru

    2002-10-01

    At the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor, the mix irradiation of thermal and epi-thermal neutrons, and the solo irradiation of epi-thermal neutrons are available additionally to the thermal neutron irradiation, and then the neutron capture therapy (NCT) at this facility became more flexible, after the update in 1996. The estimation of the depth dose distributions in NCT clinical irradiation, were performed for the standard irradiation modes of thermal, mixed and epi-thermal neutrons, from the both sides of experiment and calculation. On the assumption that the 10B concentration in tumor part was 40 ppm and the ratio of tumor to normal tissue was 3.5, the advantage depth were estimated to 5.4, 6.0, and 8.0, for the respective standard irradiation modes. It was confirmed that the various irradiation conditions can be selected according to the target-volume conditions, such as size, depth, etc. Besides, in the viewpoint of the radiation shielding for patient, it was confirmed that the whole-body exposure is effectively reduced by the new clinical collimators, compared with the old one.

  15. Effect of post-weld heat treatment and neutron irradiation on a dissimilar-metal joint between F82H steel and 316L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Haiying, E-mail: haigirl1983@gmail.com [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); Nagasaka, Takuya [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); National Institute for Fusion Science, Toki (Japan); Kometani, Nobuyuki [Nagoya University, Nagoya (Japan); Muroga, Takeo [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); National Institute for Fusion Science, Toki (Japan); Guan, Wenhai; Nogami, Shuhei; Yabuuchi, Kiyohiro; Iwata, Takuya; Hasegawa, Akira [Tohoku University, Sendai (Japan); Yamazaki, Masanori [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University (Japan); Kano, Sho; Satoh, Yuhki; Abe, Hiroaki [Institute for Materials Research, Tohoku University, Sendai (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho (Japan)

    2015-10-15

    Highlights: • Significant hardening after neutron irradiation at 300 °C for 0.1 dpa was found in the fine-grain HAZ of F82H for the dissimilar-metal joint between F82H and 316L. • The possible hardening mechanism was explained from the viewpoint of carbon behavior. • However, the significant hardening did not degrade the impact property significantly. - Abstract: A dissimilar-metal joint between F82H steel and 316L stainless steel was fabricated by using electron beam welding (EBW). By microstructural analysis and hardness test, the heat-affected zone (HAZ) of F82H was classified into interlayer area, fine-grain area, and coarse-carbide area. Post-weld heat treatment (PWHT) was applied to control the hardness of HAZ. After PWHT at 680 °C for 1 h, neutron irradiation at 300 °C with a dose of 0.1 dpa was carried out for the joint in Belgian Reactor II (BR-II). Compared to the base metals (BMs) and weld metal (WM), significant irradiation hardening up to 450HV was found in the fine-grain HAZ of F82H. However, the impact property of F82H-HAZ specimens, which was machined with the root of the V-notch at HAZ of F82H, was not deteriorated obviously in spite of the significant irradiation hardening.

  16. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Marini, B., E-mail: bernard.marini@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France); Averty, X. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SEMI (now DEN/DANS/DM2S/SEMT), F-91191 Gif-sur Yvette (France); Wident, P.; Forget, P.; Barcelo, F. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France)

    2015-10-15

    The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 °C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing.

  17. Development of advanced low alloy steel for nuclear RPV

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. C.; Shin, K. S.; Lee, S. H.; Lee, B. J. [Seoul National Univ., Seoul (Korea)

    2000-04-01

    Low carbon low alloy steels are used in nuclear power plants as pressure vessel, steam generator, etc. Nuclear pressure vessel material requires good combination of strength/ toughness, good weldability and high resistance to neutron irradiation and corrosion fatigue. For SA508III steels, most widely used in the production of nuclear power plant, attaining toughness is more difficult than strength. When taking into account the loss of toughness due to neutron irradiation, attaining as low transition temperature as possible prior to operation is a critical task in the production of nuclear pressure vessels. In the present study, we investigated detrimental microstructural features of SA508III steels to toughness, then alloy design directions to achieve improved mechanical properties were devised. The next step of alloy design was determined based on phase equilibrium thermodynamics and obtained results. Low carbon low alloy steels having low transition temperatures with enough strength and hardenability were developed. Microstructure and mechanical properties of HAZ of SA508III steels and alloy designed steels were investigated. 22 refs., 147 figs., 38 tabs. (Author)

  18. NICKEL-BASE ALLOY

    Science.gov (United States)

    Inouye, H.; Manly, W.D.; Roche, T.K.

    1960-01-19

    A nickel-base alloy was developed which is particularly useful for the containment of molten fluoride salts in reactors. The alloy is resistant to both salt corrosion and oxidation and may be used at temperatures as high as 1800 deg F. Basically, the alloy consists of 15 to 22 wt.% molybdenum, a small amount of carbon, and 6 to 8 wt.% chromium, the balance being nickel. Up to 4 wt.% of tungsten, tantalum, vanadium, or niobium may be added to strengthen the alloy.

  19. Separation of Radiocopper 64/67Cu from the Matrix of Neutron-Irradiated Natural Zinc Applicable for 64Cu Production

    Directory of Open Access Journals (Sweden)

    S. Soenarjo

    2012-04-01

    Full Text Available Radioisotope 64Cu is a promising radiometallic-isotope for molecular-targeted-radiopharmaceuticals. Having a half-life of 12.70 hours and emitting β+-radiation (E+ = 0.6531 MeV as well as β—ray (E = 0.5787 MeV, it is widely used in the form of biomedical-substrate-radiopharmaceutical for positron emission tomography (PET diagnosis and simultaneously for targeted radiotherapy of cancer. The potential needs on the availability of 64Cu-labeled pharmaceuticals for domestic nuclear medicine hospitals lead to a necessity for the local production of carrier-free 64Cu using BATAN’s G.A. Siwabessy reactor because of the technical and economical constraints in the production using BATAN’s cyclotron. The presented work is accordingly to study whether the radioisotope 64Cu can be produced and separated from the matrix of post-neutron-irradiated-natural zinc. This study is expected can be further improved and implemented in production technology of carrier-free 64Cu based on 64Zn (n,p 64Cu nuclear reaction exploiting the fast neutron fraction among the major thermal fraction due to unavailability of fast-neutron-irradiation facility in the BATAN’s G.A. Siwabessy reactor. The solution of post-neutron-irradiated-natural zinc in 1M acetic acid was loaded into Chelex-100 cation exchanger resin column to pass out the Zn/Zn* fraction whereas the Cu* fraction which remained in the column was then eluted out from the column by using 1.5 M HCl and loaded into the second column containing Dowex-1X8 anion exchanger resin. The second column was then eluted with 0.5 M HCl. The collected eluate was expected to be zinc-free Cu* fraction. It was observed from the half-life and the -spectrometric analysis that radioactive copper-64Cu containing 67Cu was produced by neutron activation on the natural Zn-foil target and can be separated from the target matrix by the presented two-steps-column-chromatographic separation technique. The radioactivity

  20. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)], E-mail: Sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Luetzenkirchen-Hecht, D.; Frahm, R. [Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)

    2009-03-31

    Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of {approx}5%. There are indications that the formation of Cu and Ni clusters differs significantly.

  2. Why neutron guides may end up breaking down? Some results on the macroscopic behaviour of alkali-borosilicate glass support plates under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Boffy, R.; Kreuz, M. [Institut Laue-Langevin, 71 avenue des Martyrs, CS 20156, F-38042 Grenoble Cedex 9 (France); Beaucour, J., E-mail: beaucour@ill.fr [Institut Laue-Langevin, 71 avenue des Martyrs, CS 20156, F-38042 Grenoble Cedex 9 (France); Köster, U. [Institut Laue-Langevin, 71 avenue des Martyrs, CS 20156, F-38042 Grenoble Cedex 9 (France); Bermejo, F.J. [Instituto de Estructura de la Materia, Consejo Superior de Investigaciones Cientificas, Serrano 123, E-20886 Madrid (Spain)

    2015-09-01

    In this paper we report on a first part of a study on the mechanisms leading to brittle fracture in neutron guides made of glass as structural element. Such devices are widely used to deliver thermal and cold neutron beams to experimental lines in most large neutron research facilities. We present results on macroscopic properties of samples of guide glass substrates which are subjected to neutron irradiation at relatively large fluences. The results show a striking dependence of some of the macroscopic properties such as density, shape or surface curvature upon the specific chemical composition of a given glass. The relevance of the present findings for the installation of either replacement guides at the existing facilities or for the deployment of instruments for ongoing projects such as the European Spallation Source is briefly discussed.

  3. Monte Carlo simulation of prompt gamma-ray spectra from depleted uranium under D-T neutron irradiation and electron recoil spectra in a liquid scintillator detector

    CERN Document Server

    Qin, Jianguo; Liu, Rong; Zhu, Tonghua; Zhang, Xinwei; Ye, Bangjiao

    2015-01-01

    To overcome the problem of inefficient computing time and unreliable results in MCNP5 calculation, a two-step method is adopted to calculate the energy deposition of prompt gamma-rays in detectors for depleted uranium spherical shells under D-T neutrons irradiation. In the first step, the gamma-ray spectrum for energy below 7 MeV is calculated by MCNP5 code; secondly, the electron recoil spectrum in a BC501A liquid scintillator detector is simulated based on EGSnrc Monte Carlo Code with the gamma-ray spectrum from the first step as input. The comparison of calculated results with experimental ones shows that the simulations agree well with experiment in the energy region 0.4-3 MeV for the prompt gamma-ray spectrum and below 4 MeVee for the electron recoil spectrum. The reliability of the two-step method in this work is validated.

  4. 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)

  5. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  6. Elaboration, physical and electrochemical characterizations of CO tolerant PEMFC anode materials. Study of platinum-molybdenum and platinum-tungsten alloys and composites; Elaborations et caracterisations electrochimiques et physiques de materiaux d'anode de PEMFC peu sensibles a l'empoisonnement par CO: etude d'alliages et de composites a base de platine-molybdene et de platine-tungstene

    Energy Technology Data Exchange (ETDEWEB)

    Peyrelade, E.

    2005-06-15

    PEMFC development is hindered by the CO poisoning ability of the anode platinum catalyst. It has been previously shown that the oxidation potential of carbon monoxide adsorbed on the platinum atoms can be lowered using specific Pt based catalysts, either metallic alloys or composites. The objective is then to realize a catalyst for which the CO oxidation is compatible with the working potential of a PEMFC anode. In our approach, to enhance the CO tolerance of platinum based catalyst supported on carbon, we studied platinum-tungsten and platinum-molybdenum alloys and platinum-metal oxide materials (Pt-WO{sub x} and Pt-MoO{sub x}). The platinum based alloys demonstrate a small effect of the second metal towards the oxidation of carbon monoxide. The platinum composites show a better tolerance to carbon monoxide. Electrochemical studies on both Pt-MoO{sub x} and Pt-WO{sub x} demonstrate the ability of the metal-oxides to promote the ability of Pt to oxidize CO at low potentials. However, chrono-amperometric tests reveal a bigger influence of the tungsten oxide. Complex chemistry reactions on the molybdenum oxide surface make it more difficult to observe. (author)

  7. Irradiation behavior of Ti 4Al 2V (ΠT-3B) alloy for ITER blanket modules flexible attachment

    Science.gov (United States)

    Rodchenkov, B. S.; Kozlov, A. V.; Kuznetsov, Yu. G.; Kalinin, G. M.; Strebkov, Yu. S.

    2007-08-01

    Titanium alloys are recommended as a material to manufacture flexible attachments of the shield blanket modules in the ITER reactor owing to their advantageous combination of properties, i.e., high resistance to impact loading, strength, density and low thermal expansion coefficient. An additional factor for selecting Ti alloys is their fast induced radioactivity decay. The (α + β)-Ti alloys have higher strength than (α)-Ti alloys but are less developed. The data base on the irradiation behavior of these materials is limited. Neutron irradiation of (α)-Ti-4Al-2V (ΠT-3B) alloy has been performed in the framework of the ITER R&D programme. Specimens from a forging of Ti-4Al-2V alloy were irradiated in the IVV-2M reactor to doses of (0.32-0.43) dpa at temperatures of (240-260) °C. This paper describes the results of tensile, low cycle fatigue and fracture toughness tests of alloy in the unirradiated and neutron irradiated conditions. The results obtained are compared with those of the (α + β)-Ti-6Al-4V alloy.

  8. Bioaccessibility of micron-sized powder particles of molybdenum metal, iron metal, molybdenum oxides and ferromolybdenum--Importance of surface oxides.

    Science.gov (United States)

    Mörsdorf, Alexander; Odnevall Wallinder, Inger; Hedberg, Yolanda

    2015-08-01

    The European chemical framework REACH requires that hazards and risks posed by chemicals, including alloys and metals, that are manufactured, imported or used in different products (substances or articles) are identified and proven safe for humans and the environment. Metals and alloys need hence to be investigated on their extent of released metals (bioaccessibility) in biologically relevant environments. Read-across from available studies may be used for similar materials. This study investigates the release of molybdenum and iron from powder particles of molybdenum metal (Mo), a ferromolybdenum alloy (FeMo), an iron metal powder (Fe), MoO2, and MoO3 in different synthetic body fluids of pH ranging from 1.5 to 7.4 and of different composition. Spectroscopic tools and cyclic voltammetry have been employed to characterize surface oxides, microscopy, light scattering and nitrogen absorption for particle characterization, and atomic absorption spectroscopy to quantify released amounts of metals. The release of molybdenum from the Mo powder generally increased with pH and was influenced by the fluid composition. The mixed iron and molybdenum surface oxide of the FeMo powder acted as a barrier both at acidic and weakly alkaline conditions. These findings underline the importance of the surface oxide characteristics for the bioaccessibility of metal alloys.

  9. Boron - Molybdenum - Tungsten

    Science.gov (United States)

    Bulanova, Marina; Heulens, Jeroen

    This document is part of Volume 11 `Ternary Alloy Systems: Phase Diagrams, Crystallographic and Thermodynamic Data', Subvolume E `Refractory Metal Systems', of Landolt-Börnstein - Group IV `Physical Chemistry'.

  10. PERSPECTIVES OF MOLIBDENUM CONTAINING MATERIALS APPLICATION FOR ALLOYING OF IRONCARBON ALLOYS DURING MANUFACTURING OF CRITICAL CASTINGS

    Directory of Open Access Journals (Sweden)

    A. G. Slutsky

    2015-01-01

    Full Text Available Motor is one of most important part of automobile determine its economical effectiveness of usage. On the other hand, sleeves, pistons and rings are crucible parts as they determine the service life of a motor. These parts are producing in big scale – dozens of millions pieces. Increase of cylinder sleeves physical-mechanical properties results in prolongation of motor service life and improvement of motor’s characteristics. Nowadays low alloyed cast irons with perlite structure are used to manufacture motor’s sleeves. For alloying purposes such traditional elements as Cr, Ni, Cu, and V are applied. But it is interesting to use molybdenum for cast iron alloying. It is known that alloying of alloys allows considerable increasing of consumption properties of castings. But in spite of advantages of alloys alloying the increase of molybdenum containing iron-carbon alloys production is restricted by economical reasons – high cost of alloying additions. Expenditures on alloying additions can be reduced by the application cheap secondary alloys in the charge. So, the present paper is devoted to investigation of alloying peculiarities during the treatment of ferrous alloys with molybdenum applying different initial materials.

  11. Molybdenum Tube Characterization report

    Energy Technology Data Exchange (ETDEWEB)

    Beaux II, Miles Frank [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Usov, Igor Olegovich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-07

    Chemical vapor deposition (CVD) techniques have been utilized to produce free-standing molybdenum tubes with the end goal of nuclear fuel clad applications. In order to produce tubes with properties desirable for this application, deposition rates were lowered requiring long deposition durations on the order of 50 hours. Standard CVD methods as well as fluidized-bed CVD (FBCVD) methods were applied towards these objectives. Characterization of the tubes produced in this manner revealed material suitable for fuel clad applications, but lacking necessary uniformity across the length of the tubes. The production of freestanding Mo tubes that possess the desired properties across their entire length represents an engineering challenge that can be overcome in a next iteration of the deposition system.

  12. Study of the behaviour under neutron irradiation of hafnium diboride; Etude du comportement sous irradiation neutronique du diborure d`hafnium

    Energy Technology Data Exchange (ETDEWEB)

    Cheminant-Coatanlem, P

    1998-12-31

    Owing to its good neutron cross section and to its high melting point, hafnium diboride is a potential candidate for a use as neutron absorbing material in control rod of pressurized water reactor of the next generation. The main causes of damage under neutron irradiation in this ceramic are due to the {sup 10}B(n,{alpha}){sup 7}Li reaction that introduces in the crystal structure new atoms and point defects. The materials under consideration are the stoichiometric HfB{sub 2} compound and the HfB{sub 2} + 10 vol. % Hf compound. They are been irradiated with neutrons at several fluences and temperatures. Electron irradiations, helium and lithium implantations have been carried out in order to simulate the creation of point defects and/or fission products. Transmission and scanning electron microscopy have been used to determine damage mechanisms in HfB{sub 2}. At a low temperature (<500 deg C), irradiation defects precipitate in dislocation loops of both nature, interstitial and vacancy. Those loops have a particular organisation in the HfB{sub 2} lattice: vacancy loops are lying in the basal plane and interstitial loops in planes perpendicular to basal planes. This induces anisotropic deformation of grains that originates internal stress development. These stresses are associated with the dislocation staking and consequently with the cavity formation at grain boundaries. At a higher temperature (>700 deg C), the same dislocation loops are observed. But, in addition, the irradiation defects diffuse to grain boundaries where helium bubbles are formed. The damage caused by this latter mechanism becomes predominant. The HfB{sub 2} + 10 vol. % Hf materials is more resistant under neutron irradiation than the HfB{sub 2} pellets that display a very damaged surface. This result is explained by the fact that, on the one band, the HfB{sub 2} + 10 vol. % Hf pellets have a higher toughness than the HfB{sub 2} pellets and, on the other hand, the HfB{sub 2} + 10 vol. % Hf

  13. Analysis of recovery process of low-dose neutron irradiation-induced defects in silicon nitride-based ceramics by thermal annealing

    Energy Technology Data Exchange (ETDEWEB)

    Rueanngoen, Areerak, E-mail: areerak_k@yahoo.com [Department of Nuclear Engineering, Graduate School of Science and Engineering, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Kanazawa, Koumei [Department of Nuclear Engineering, Graduate School of Science and Engineering, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Imai, Masamitsu; Yoshida, Katsumi; Yano, Toyohiko [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-12-15

    Two kinds of silicon nitride ceramics consisting of different polymorphs were neutron-irradiated up to 8.5 × 10{sup 24} n/m{sup 2} (E > 0.1 MeV) at 563 K, and their annealing behaviors were compared to those of previously reported SiAlON polymorphs subjected to the same irradiation condition. The macroscopic length change of α- and β-Si{sub 3}N{sub 4} and α- and β-SiAlON were 0.11%, 0.06%, 0.12% and 0.14%, respectively. Based on swelling data and microstructural observations, the low dose neutron irradiation-induced defects in silicon nitride-based ceramics were considered to be primarily point defects. In order to investigate the kinetics of defect recovery, these irradiated specimens were isothermally and isochronally annealed continuously up to 1473 K. Macroscopic length change decreased gradually with increasing annealing temperature. Recovery curves of isochronal annealing of α-Si{sub 3}N{sub 4} and α-SiAlON were similar, and those of β-Si{sub 3}N{sub 4} and β-SiAlON were also similar. The recombination rate constant as a first-order reaction increased with the increasing of the isothermal annealing temperature. A two-stage recovery process was considered between the irradiation temperature and 1473 K. The activation energies at higher temperatures were almost double those at lower temperatures in both Si{sub 3}N{sub 4} and SiAlON. At lower temperatures range the recovery should occur by annihilation of close-spaced Frenkel pairs. On the other hand, at higher temperatures, the recovery process may be governed by the annihilation of separated Frenkel pairs. In addition, the activation energies for defect recovery in Si{sub 3}N{sub 4} were larger than defects in SiAlON. Recovery characteristics of α- and β-phases were different in both crystals that are suggested to be due to differences in crystal structures.

  14. Castable nickel aluminide alloys for structural applications

    Science.gov (United States)

    Liu, Chain T.

    1992-01-01

    The specification discloses nickel aluminide alloys which include as a component from about 0.5 to about 4 at. % of one or more of the elements selected from the group consisting of molybdenum or niobium to substantially improve the mechanical properties of the alloys in the cast condition.

  15. A Simple Kinetic Model of Zircaloy Zr(Fe,Cr){sub 2} Precipitate Amorphization During Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, D.F.; Peters, H.R.; Yang, W.J.S.

    1999-07-01

    At neutron flux levels typical for Zircaloy fuel cladding in commercial power reactors, there is insufficient thermal energy below about 600 K to maintain long-range order in hexagonal close packed (hcp) Zr(Fe,Cr){sub 2} precipitates, and these Laves-phase intermetallics gradually become amorphous. The transformation is homogeneous with no change in composition at low temperatures, but above 500 K an amorphous zone containing only 10 at% Fe grows inward from the periphery as Fe moves outward to the adjacent alloy matrix. The shrinking central cores of Zr(Fe,Cr){sub 2} precipitates in Zircaloy-4 remain crystalline, while in Zircaloy-2 these precipitates quickly undergo partial transformation and the low-Fe amorphous front advances into a random mixture of amorphous and crystalline regions, each with the original composition. Above 600 K, the Zr(Fe,Cr){sub 2} precipitates tend to retain both their hcp structure and original chemical composition. These observations suggest that a dynamic competition between kinetic excitation to an amorphous state and thermal recrystallization makes some fraction of the Fe atoms available for flux-assisted diffusion to the alloy matrix by displacing them from hcp lattice positions into metastable, probably interstitial, sites. With one set of kinetic constants, a simple analytic representation of these processes accurately predicts precipitate amorphization as a function of neutron flux, temperature, and time for either Zircaloy-2 or -4. By implication, over the composition range of interest, hcp Zr(Fe,Cr){sub 2} is most stable thermodynamically with about 33 at% Fe, typical of Zircaloy-2, but amorphous Zr(Fe,Cr){sub 2} has the smallest activation energy for recrystallization with the slightly higher Fe content typical of Zircaloy-4.

  16. Biocompatibility of dental alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braemer, W. [Heraeus Kulzer GmbH and Co. KG, Hanau (Germany)

    2001-10-01

    Modern dental alloys have been used for 50 years to produce prosthetic dental restorations. Generally, the crowns and frames of a prosthesis are prepared in dental alloys, and then veneered by feldspar ceramics or composites. In use, the alloys are exposed to the corrosive influence of saliva and bacteria. Metallic dental materials can be classified as precious and non-precious alloys. Precious alloys consist of gold, platinum, and small amounts of non-precious components such as copper, tin, or zinc. The non-precious alloys are based on either nickel or cobalt, alloyed with chrome, molybdenum, manganese, etc. Titanium is used as Grade 2 quality for dental purposes. As well as the dental casting alloys, high purity electroplated gold (99.8 wt.-%) is used in dental technology. This review discusses the corrosion behavior of metallic dental materials with saliva in ''in vitro'' tests and the influence of alloy components on bacteria (Lactobacillus casei and Streptococcus mutans). The test results show that alloys with high gold content, cobalt-based alloys, titanium, and electroplated gold are suitable for use as dental materials. (orig.)

  17. Filler metal alloy for welding cast nickel aluminide alloys

    Science.gov (United States)

    Santella, M.L.; Sikka, V.K.

    1998-03-10

    A filler metal alloy used as a filler for welding cast nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and cast in copper chill molds. 3 figs.

  18. Filler metal alloy for welding cast nickel aluminide alloys

    Energy Technology Data Exchange (ETDEWEB)

    Santella, Michael L. (Knoxville, TN); Sikka, Vinod K. (Oak Ridge, TN)

    1998-01-01

    A filler metal alloy used as a filler for welding east nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and east in copper chill molds.

  19. J{sub c} increase of MPMG-processed YBa{sub 2}Cu{sub 3}O{sub 7-x} (Y-123) bulk due to fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Terai, Takayuki [University of Tokyo, 7-3-1 Hongo, Tokyo 113-8656 (Japan)], E-mail: tera@n.t.u-tokyo.ac.jp; Nagamoto, Yoshifumi; Kubo, Toshiharu [University of Tokyo, 7-3-1 Hongo, Tokyo 113-8656 (Japan); Chikumoto, Noriko [SRL-ISTEC, 1-10-13 Shinonome, Tokyo 135-0062 (Japan); Sawa, Kazuhiro [Japan Atomic Energy Agency, 3607 Narita-cho, Oarai 311-1394 (Japan)

    2007-09-01

    The change in pinning properties of YBa{sub 2}Cu{sub 3}O{sub 7-x} (Y-123) prepared by MPMG process due to fast neutron irradiation and thermal annealing treatment was investigated in order to confirm that fast neutron irradiation is effective for the J{sub c} increase in practical material as well as single crystal. In this study, MPMG-processed Y-123 specimens (1.5 mm x 1.0 mm x 0.5 mm) with 0.5 wt% of Pt were irradiated with fast neutrons from 2.9 x 10{sup 15} cm{sup -2} to 8.0 x 10{sup 18} cm{sup -2} below 313 K. Before and after the fast neutron irradiation, J{sub c} was evaluated from the magnetic hysteresis loop at 20 K, 40 K and 60 K using the extended Bean's model, and pinning potential was obtained from the magnetic relaxation curve at 20 K. In addition, T{sub c} was evaluated by AC magnetization measurement. The maximum values of J{sub c} were 1200 kA cm{sup -2} at 20 K and 7 T, and 400 kA cm{sup -2} at 40 K and 7 T. They exceeded the target value for its practical superconducting magnet uses e.g. nuclear fusion reactor, magnetic levitation, etc. Furthermore, T{sub c} was recovered sufficiently by the thermal annealing treatment. It turns out that the pinning centers introduced by neutron irradiation work effectively also in the practical material which has already included the strong pinning centers of dispersed Y-211 phase.

  20. Iron binary and ternary coatings with molybdenum and tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Yar-Mukhamedova, Gulmira, E-mail: gulmira-alma-ata@mail.ru [Institute Experimental and Theoretical Physics Al-Farabi Kazakh National University, 050038, Al-Farabi av., 71, Almaty (Kazakhstan); Ved, Maryna; Sakhnenko, Nikolay; Karakurkchi, Anna; Yermolenko, Iryna [National Technical University “Kharkov Polytechnic Institute”, Kharkov (Ukraine)

    2016-10-15

    Highlights: • High quality coatings of double Fe-Mo and ternary Fe-Mo-W electrolytic alloys can be produced both in a dc and a pulsed mode. • Application of unipolar pulsed current allows receiving an increased content of the alloying components and their more uniform distribution over the surface. • It is established that Fe-Mo and Fe-Mo-W coatings have an amorphous structure and exhibit improved corrosion resistance and microhardness as compared with the steel substrate due to the inclusion molybdenum and tungsten. - Abstract: Electrodeposition of Fe-Mo-W and Fe-Mo layers from a citrate solution containing iron(III) on steel and iron substrates is compared. The utilization of iron(III) compounds significantly improved the electrolyte stability eliminating side anodic redox reactions. The influence of concentration ratios and electrodeposition mode on quality, chemical composition, and functional properties of the alloys is determined. It has been found that alloys deposited in pulse mode have more uniform surface morphology and chemical composition and contain less impurities. Improvement in physical and mechanical properties as well as corrosion resistance of Fe-Mo and Fe-Mo-W deposits when compared with main alloy forming metals is driven by alloying components chemical passivity as well as by alloys amorphous structure. Indicated deposits can be considered promising materials in surface hardening technologies and repair of worn out items.

  1. Accumulation of dislocation loops in the α phase of Zr Excel alloy under heavy ion irradiation

    Science.gov (United States)

    Yu, Hongbing; Yao, Zhongwen; Idrees, Yasir; Zhang, He K.; Kirk, Mark A.; Daymond, Mark R.

    2017-08-01

    In-situ heavy ion irradiations were performed on the high Sn content Zr alloy 'Excel', measuring type dislocation loop accumulation up to irradiation damage doses of 10 dpa at a range of temperatures. The high content of Sn, which diffuses slowly, and the thin foil geometry of the sample provide a unique opportunity to study an extreme case where displacement cascades dominate the loop formation and evolution. The dynamic observation of dislocation loop evolution under irradiation at 200 °C reveals that type dislocation loops can form at very low dose (0.0025 dpa). The size of the dislocation loops increases slightly with irradiation damage dose. The mechanism controlling loop growth in this study is different from that in neutron irradiation; in this study, larger dislocation loops can condense directly from the interaction of displacement cascades and the high concentration of point defects in the matrix. The size of the dislocation loop is dependent on the point defect concentration in the matrix. A negative correlation between the irradiation temperature and the dislocation loop size was observed. A comparison between cascade dominated loop evolution (this study), diffusion dominated loop evolution (electron irradiation) and neutron irradiation suggests that heavy ion irradiation alone may not be enough to accurately reproduce neutron irradiation induced loop structures. An alternative method is proposed in this paper. The effects of Sn on the displacement cascades, defect yield, and the diffusion behavior of point defects are established.

  2. Effect on fast neutron irradiation to 4 dpa at 400{degrees}C on the properties of V-(4-5)Cr-(4-5)Ti alloys

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J.; Alexander, D.J.; Robertson, J.P. [Oak Ridge National Lab., TN (United States)] [and others

    1997-04-01

    Tensile, Charpy impact and electrical resistivity measurements have been performed at ORNL on V-4Cr-4Ti and V-5Cr-5Ti specimens that were prepared at ANL and irradiated in the lithium-bonded X530 experiment in the EBR-II fast reactor. All of the specimens were irradiated to a damage level of about 4 dpa at a temperature of {approximately}400{degrees}C. A significant amount of radiation hardening was evident in both the tensile and Charpy impact tests. The irradiated V-4Cr-4Ti yield strength measured at {approximately}390{degrees}C was >800 MPa, which is more than three times as high as the unirradiated value. The uniform elongations of the irradiated tensile specimens were typically {approximately}1%, with corresponding total elongations of 4-6%. The ductile to brittle transition temperature of the irradiated specimens was less than the unirradiated resistivity, which suggests that hardening associated with interstitial solute pickup was minimal.

  3. The changes of the structural, magnetic, and mechanical properties in a reactor pressure vessel steel neutron-irradiated at 70 .deg. C

    CERN Document Server

    Park, D G; Jang, K S; Jung, M M; Kim, G M

    1999-01-01

    The irradiation embrittlement of reactor-pressure-vessel steel has been one of the main safety concerns in nuclear power plants. In the present study, an SA508-3 RPV steel was irradiated by neutrons with various fluences up to 10 sup 1 sup 8 n/cm sup 2 (E>=1MeV) at a temperature of approximately 70 .deg. C. The irradiation responses of the structural, the magnetic, and the mechanical properties of the steel were investigated by means of X-ray diffraction, Moessbauer spectroscopy, magnetic Barkhausen noise, and micro-Vickers hardness measurements. The transitions of all of these parameters occurred above a neutron does of 10 sup 1 sup 6 n/cm sup 2. The results of the X-ray and the Moessbauer experiments revealed that neutron irradiation led to the possibility of partial amorphization in the investigated RPV steel. The changes of the physical and the mechanical properties were discussed in terms of irradiation-induced cascade damage of crystalline materials.

  4. The influence of fast neutron irradiation on the intra- and intergrain properties of the polycrystalline BiPbSrCaCuO system

    Science.gov (United States)

    Wiśniewski, A.; Baran, M.; Kozioł, Z.; Przysłupski, P.; Piechota, J.; Puźniak, R.; Pajaçzkowska, A.; Pȩkała, M.; Pytel, B.; Pytel, K.

    1990-09-01

    The influence of irradiation by fast neutrons with fluences from 3.3 x 10 16n/ cm2 up to 3 x 10 18n/ cm2 on the physical properties of polycrystalline Bi0.7Pb0.3SrCaCu1.8Ox was examined. Studies of DC magnetization, AC susceptibility, transport and thermoelectric power were performed. The irradiation caused a decrease of Tc, determined from the onset of diamagnetism, by as much as 31 K for a fluence of 3 x 10 18n/ cm2. A strong influence of neutron irradiation on both intra- and intergranular properties was observed. The defects within the superconducting grains created by neutrons caused an increase of the pinning forces which enhanced the critical magnetization current. A gradual decoupling of Josephson weak links with increasing neutron fluence was observed in transport and low field magnetization measurements. From the AC susceptibility measurements the irreversibility lines between the flux-creep and flux-flow regions were determined. An increase of the absolute values of thermoelectric power with rising fluence was noticed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Stergar, E., E-mail: estergar@sckcen.be [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Eremin, S.G. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation); Gavrilov, S.; Lambrecht, M. [SCK-CEN, Belgian Nuclear Research Centre, Boeretang 200, 2400 Mol (Belgium); Makarov, O.; Iakovlev, V. [RIAR, Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

    2016-05-15

    The LEXUR–II–LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

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

    Science.gov (United States)

    Stergar, E.; Eremin, S. G.; Gavrilov, S.; Lambrecht, M.; Makarov, O.; Iakovlev, V.

    2016-05-01

    The LEXUR-II-LBE irradiation campaign was conducted from 2011 to 2012 and was aimed to investigate the combined influence of irradiation and LBE environment. In this irradiation campaign tensile test samples, pressurized tubes and corrosion samples were irradiated in LBE filled capsules. To separate the effect of exposure to LBE and neutron irradiation a parallel furnace experiment where the samples were exposed to LBE at the irradiation temperature for the corresponding time was conducted. Here we report results of the first extracted capsule which was irradiated about 6 months and dismantled after a cooling phase to decrease activity. The results of SSRT tests for irradiated T91 show that the exposure to LBE at 350 °C for a long time leads to the appearance of liquid metal embrittlement without any pre-treatment which is usually necessary to promote LME. Irradiation increases the effect of LME on the ductility of T91. In contrast to the findings for T91 the gained results also show that tensile tests on irradiated austenitic stainless steel 316L show no influence of LBE environment on the tensile properties.

  7. A study of the neutron irradiation effects on the susceptibility to embrittlement of A316L and T91 steels in lead-bismuth eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Sapundjiev, D. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium)]. E-mail: danislav.sapundjiev@sckcen.be; Al Mazouzi, A. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium); Van Dyck, S. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium)

    2006-09-15

    The effects of neutron irradiation on the susceptibility to liquid metal embrittlement of two primary selected materials for MYRRHA project an accelerator driven system (ADS), was investigated by means of slow strain rate tests (SSRT). The latter were carried out at 200 deg. C in nitrogen and in liquid Pb-Bi at a strain rate of 5 x 10{sup -6} s{sup -1}. The small tensile specimens were irradiated at the BR-2 reactor in the MISTRAL irradiation rig at 200 deg. C for 3 reactor cycles to reach a dose of about 1.50 dpa. The SSR tests were carried out under poor and under dissolved oxygen conditions ({approx}1.5 x 10{sup -12} wt% dissolved oxygen) which at this temperature will favour formation of iron and chromium oxides. Although both materials differ in structure (fcc for A316L against bcc for T91), their flow behaviour in contact with liquid lead bismuth eutectic before and after irradiation is very similar. Under these testing conditions none of them was found susceptible to liquid metal embrittlement (LME)

  8. Development and characteristics of the HANARO ex-core neutron irradiation facility for applications in the boron neutron capture therapy field

    CERN Document Server

    Kim, M S; Jun, B J; Kim, H; Lee, B C; Hwang, Sung-Yul; Jun, Byung-Jin; Kim, Heonil; Kim, Myong-Seop; Lee, Byung-Chul

    2006-01-01

    The HANARO ex-core neutron irradiation facility was developed for various applications in the boron neutron capture therapy (BNCT) field, and its characteristics have been investigated. In order to obtain a sufficient thermal neutron flux with a low level contamination of fast neutrons and gamma-rays, a radiation filtering method is adopted. The radiation filter has been designed by using a silicon single crystal cooled by liquid nitrogen and a bismuth crystal. The installation of the main components of the irradiation facility and the irradiation room are finished. Experimental measurements of the neutron beam characteristics have been performed by using bare and cadmium covered gold foils and wires. The in-phantom neutron flux distribution was measured for a flux mapping inside the phantom. The gamma-ray dose was determined by using TLD-700 thermoluminescence dosimeters. The thermal and fast neutron fluxes and the gamma-ray dose were calculated by using the MCNP code, and they were compared with experimenta...

  9. Post-irradiation annealing of Ni–Mn–Si-enriched clusters in a neutron-irradiated RPV steel weld using Atom Probe Tomography

    Energy Technology Data Exchange (ETDEWEB)

    Styman, P.D., E-mail: paul.styman@materials.ox.ac.uk [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Hyde, J.M. [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Parfitt, D.; Wilford, K. [Rolls-Royce, PO BOX 2000, Raynesway, Derby DE21 7XX (United Kingdom); Burke, M.G. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); English, C.A. [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Efsing, P. [Vattenfall Ringhals AB, Väröbacka (Sweden)

    2015-04-15

    Highlights: • Characterisation of high Ni neutron irradiated RPV surveillance samples at high fluence. • Post-irradiation annealing performed to give insight into the formation mechanisms of Ni–Mn–Si precipitates. • Dissolution of Ni–Mn–Si clusters appears to be lead by the removal of Mn. - Abstract: Atom Probe Tomography has been performed on as-irradiated and post-irradiation annealed surveillance weld samples from Ringhals Unit 3. The weld contains low Cu (0.07 at.%) and high Ni (1.5 at.%). A high number density (∼4 × 10{sup 23} m{sup −3}) of Ni–Mn–Si-enriched clusters was observed in the as-irradiated material. The onset of recovery was observed during the annealing for 30 min at 450 °C. Much more significant dissolution of clusters occurred during the 10 min 500 °C anneal, resulting in a reduction in mean cluster size and a halving of their volume fraction. Detailed analyses of the changes in microstructure demonstrate that the dissolution process is driven by migration of Mn atoms from the clusters. This may indicate a strong correlation between Mn and point defects. Dissolution of the clusters is shown to correlate with recovery of mechanical properties in this material.

  10. Evolution of Helium with Temperature in Neutron-Irradiated 10B-Doped Aluminum by Small-Angle X-Ray Scattering

    Directory of Open Access Journals (Sweden)

    Chaoqiang Huang

    2014-01-01

    Full Text Available Helium status is the primary effect of material properties under radiation. 10B-doped aluminum samples were prepared via arc melting technique and rapidly cooled with liquid nitrogen to increase the boron concentration during the formation of compounds. An accumulated helium concentration of ~6.2 × 1025 m−3 was obtained via reactor neutron irradiation with the reaction of 10B(n, α7Li. Temperature-stimulated helium evolution was observed via small-angle X-ray scattering (SAXS and was confirmed via transmission electron microscopy (TEM. The SAXS results show that the volume fraction of helium bubbles significantly increased with temperature. The amount of helium bubbles reached its maximum at 600°C, and the most probable diameter of the helium bubbles increased with temperature until 14.6 nm at 700°C. A similar size distribution of helium bubbles was obtained via TEM after in situ SAXS measurement at 700°C, except that the most probable diameter was 3.9 nm smaller.

  11. Phase stability, swelling, microstructure and strength of Ti3SiC2-TiC ceramics after low dose neutron irradiation

    Science.gov (United States)

    Ang, Caen; Zinkle, Steven; Shih, Chunghao; Silva, Chinthaka; Cetiner, Nesrin; Katoh, Yutai

    2017-01-01

    Mn+1AXn (MAX) phase Ti3SiC2 materials were neutron irradiated at ∼400, ∼630, and 700 °C to a fluence of ∼2 × 1025 n/m2 (E > 0.1 MeV). After irradiation at ∼400 °C, anisotropic c-axis dilation of ∼1.5% was observed. Room temperature strength was reduced from 445 ± 29 MPa to 315 ± 33 MPa and the fracture surfaces showed flat facets and transgranular cracks instead of typical kink-band deformation and bridging ligaments. XRD phase analysis indicated an increase of 10-15 wt% TiC. After irradiation at ∼700 °C there were no lattice parameter changes, ∼5 wt% decomposition to TiC occurred, and strength was 391 ± 71 MPa and 378 ± 31 MPa. The fracture surfaces indicated kink-band based deformation but with lesser extent of delamination than as-received samples. Ti3SiC2 appears to be radiation tolerant at ∼400 °C, and increasingly radiation resistant at ∼630-700 °C, but a higher temperature may be necessary for full recovery.

  12. Low-temperature low-dose neutron irradiation effects on Brush Wellman S65-C and Kawechi Berylco P0 beryllium

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

    The mechanical property results for two high quality beryllium materials subjected to low temperature, low dose neutron irradiation in water moderated reactors are presented. Materials chosen were the S65-C ITER candidate material produced by Brush Wellman, and Kawecki Berylco Industries P0 beryllium. Both materials were processed by vacuum hot pressing. Mini sheet tensile and thermal diffusivity specimens were irradiated in the temperature range of {approximately}100--275 C from a fast (E > 0.1 MeV) neutron dose of 0.05 to 1.0 {times} 10{sup 25} n/m{sup 2} in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory and the High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory. As expected from earlier work on beryllium, both materials underwent significant embrittlement with corresponding reduction in ductility and increased strength. Both thermal diffusivity and volumetric expansion were measured and found to be negligible in this temperature and fluence range. Of significance from this work is that while both materials rapidly embrittle at these ITER relevant irradiation conditions, some ductility (>1--2%) remains, which contrasts with a body of earlier work including recent work on the Brush-Wellman S65-C material irradiated to slightly higher neutron fluence.

  13. The Structure and Stability of Molybdenum Ditelluride Thin Films

    Directory of Open Access Journals (Sweden)

    Zhouling Wang

    2014-01-01

    Full Text Available Molybdenum-tellurium alloy thin films were fabricated by electron beam evaporation and the films were annealed in different conditions in N2 ambient. The hexagonal molybdenum ditelluride thin films with well crystallization annealed at 470°C or higher were obtained by solid state reactions. Thermal stability measurements indicate the formation of MoTe2 took place at about 350°C, and a subtle weight-loss was in the range between 30°C and 500°C. The evolution of the chemistry for Mo-Te thin films was performed to investigate the growth of the MoTe2 thin films free of any secondary phase. And the effect of other postdeposition treatments on the film characteristics was also investigated.

  14. Effect of bonding and bakeout thermal cycles on the properties of copper alloys irradiated at 350 degrees C

    DEFF Research Database (Denmark)

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

    2001-01-01

    Screening experiments were carried out to determine the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile...... results are described and their salient features discussed. The most significant effect of neutron irradiation is a severe loss of ductility in the case of CuNiBe alloys. (C) 2001 Elsevier Science B.V. All rights reserved....

  15. Density of Liquid Ni-Mo Alloys Measured by a Modified Sessile Drop Method

    Institute of Scientific and Technical Information of China (English)

    Liang FANG; Zushu LI; ZaiNan TAO; Feng XIAO

    2004-01-01

    The density of liquid binary Ni-Mo alloys with molybdenum concentration from 0 to 20% (mass fraction) was measured by a modified sessile drop method. It has been found that the density of the liquid Ni-Mo alloys decreases with increasing temperature, but increases with the increase of molybdenum concentration in the alloys. The molar volume of liquid Ni-Mo binary alloys increases with the increase of temperature and molybdenum concentration. The partial molar volume of molybdenum in Ni-Mo binary alloy has been approximately calculated as [13.18 - 2.65 × 10-3T + (-47.94 + 3.10 × 10-2T) × 10-2XMo] × 10-6m3·mol-1. The molar volume of Ni-Mo alloy determined in the present work shows a negative deviation from the ideal linear mixing molar volume.

  16. Duct and cladding alloy

    Science.gov (United States)

    Korenko, Michael K.

    1983-01-01

    An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

  17. Order-Disorder Phenomena in the Binary Alloys Platinum(x)titanium, Platinum(x)vanadium (3 Less than or Equal to X Less than or Equal to 8) and NICKEL(3)MOLYBDENUM Studied by High Resolution Electron Microscopy.

    Science.gov (United States)

    Schryvers, Dominique

    In this work the results of an experimental study on the order-disorder behaviour of three different binary alloy systems are presented. The investigations were performed mainly using electron diffraction and high resolution electron microscopy techniques. In the first chapter an introduction with respect to the general aspects of ordering in alloys is given. The second chapter describes the most important properties of the investigation techniques. Both chapters are written in function of the needs in following chapters. Chapter III comprehends the results obtained in the platinum rich part of the Pt-Ti alloy system, while those of Pt-V are described in chapter IV. The main results in these alloy systems concern the characterization of new ordered phases. The complex dynamical ordering mechanism existing in {rm Ni_3 Mo} is discussed in chapter V. In chapter VI a calculation of possible ground state structures of a binary system based on the face centered cubic lattice and including the fourth nearest neighbour pair interaction is presented. A comparison between the theoretical results and the experimental ones of chapters III and IV is given.

  18. Experimental comparisons among various models for the reverse annealing of the effective concentration of ionized space charges (N{sub eff}) of neutron irradiated silicon detectors

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zheng

    1994-03-01

    Experimental data of the reverse annealing of the effective concentration of ionized space charges (N{sub eff}, also called effective doping or impurity concentration) of neutron irradiated high resistivity silicon detectors has been compared with various models: compensation model (first order), cluster model of the first order, neutral to acceptor model (first order), and cluster model of the second order. Detectors irradiated to various neutron fluences have been annealed at 80{degree}C for up to 17 hours to reach the saturation of the first apparent stage of the N{sub eff} reverse anneal, which is equivalent of about one year of room temperature (RT) anneal. The anneal time constant, defined as the time at half saturation {tau}{sub {1/2}}, has been found virtually a constant ({approximately}140 minutes {plus_minus} 14%) for detectors irradiated to fluences ranging from 8.2 {times} 10{sup 12} n/cm{sup 2} to 3.2 {times} 10{sup 13} n/cm{sup 2}, which is characteristic of the first order process. The least square fit of the data to the first order models has shown a time constant of 221.7 minutes with 14% error and that to the second order model has shown a k constant of 7.3 {times} 10{sup {minus}5} s{sup {minus}1} with 37% error. The best fit, however, is a first order fit with two time constants: a short one ({approximately}44 minutes {plus_minus} 25%) with a small amplitude and a longer one ({approximately}290 minutes {plus_minus} 12%) with almost five times as larger amplitude, suggesting that even for the apparent first stage of the N{sub eff} reverse anneal, there may be two stages. There is also evidence that even after the apparent first stage anneal, there is at least another stage which is showing up in higher temperature anneal (150{degree}C).

  19. Effects of neutron irradiation on glass ceramics as pressure-less joining materials for SiC based components for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Ferraris, M., E-mail: monica.ferraris@polito.it [Department of Applied Science and Technology, Politecnico di Torino, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Casalegno, V.; Rizzo, S.; Salvo, M. [Department of Applied Science and Technology, Politecnico di Torino, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Van Staveren, T.O. [NRG (Nuclear Research and Consultancy Group) Petten (Netherlands); Matejicek, J. [Institute of Plasma Physics, Prague (Czech Republic)

    2012-10-15

    This paper reports on the microstructure and properties of two glass-ceramics based on SiO{sub 2}-Al{sub 2}O{sub 3}-MgO (SAMg) and SiO{sub 2}-Al{sub 2}O{sub 3}-Y{sub 2}O{sub 3} (SAY), which have been designed to be used as pressure-less low activation joining materials for SiC/SiC and SiC based components for nuclear applications. Glass-ceramic pellets (SAY and SAMg) were irradiated for approximately 1 year in the reactor core of the LVR-15 research reactor at Nuclear Research Institute Rez, Czech Republic, at about 50 Degree-Sign C, 6.92 Multiplication-Sign 10{sup 24} n/m{sup 2} (E > 1 MeV, about 1 dpa in steel); SiC/SiC composites joined by SAY were irradiated about 1 year at High Flux Reactor (HFR), Petten, The Netherlands, 550 Degree-Sign C, 9-11 Multiplication-Sign 10{sup 24} n/m{sup 2} (E > 1 MeV, about 1.4-1.8 dpa in C), 600 Degree-Sign C, 16-22 Multiplication-Sign 10{sup 24} n/m{sup 2} (E > 1 MeV, about 2.6-3.3 dpa in C) and 820 Degree-Sign C 31-32 Multiplication-Sign 10{sup 24} n/m{sup 2}(E > 1 MeV, about 5 dpa in C). Optical microscopy with image analysis and scanning electron microscopy (SEM) with X-ray microanalysis (EDS) were used to investigate the glass-ceramics morphology and composition, showing a remarkable similarity before and after neutron irradiation for both glass-ceramics. Comparison of bending strength for irradiated and non-irradiated SAY joined SiC/SiC indicate that the mechanical strength is unaffected by irradiation at these conditions.

  20. Study of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} lithium ceramics after long-term neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chikhray, Y. [Kazakh National University, Almaty (Kazakhstan)], E-mail: john@physics.kz; Shestakov, V. [Kazakh National University, Almaty (Kazakhstan); Maksimkin, O.; Turubarova, L.; Osipov, I. [Institute of Nuclear Physics, Almaty (Kazakhstan); Kulsartov, T.; Kuykabayeba, A.; Tazhibayeva, I. [National Nuclear Center, Kurchatov (Kazakhstan); Kawamura, H.; Tsuchiya, K. [JAEA, Oarai (Japan)

    2009-04-30

    Given work presents the results of complex material-science studies of 1 mm diameter ceramic pebbles manufactured of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} ceramics before and after long-time neutron irradiation. Ceramic samples were placed in specially ampoules (six items) made of stainless steel Cr18Ni10Ti which were vacuumized and filled with helium. Irradiation of ampoules was carried out in the loop channel of WWRK reactor (Almaty, Kazakhstan) during 223 days at 6 MW power. After irradiation light-colored pebbles became grey-colored due to structure changes which generation of grey-colored inclusions (lithium oxide) with low density and microhardness. There is a radiation softening of lithium ceramic and that effect is higher for lower irradiation temperature 760 K than for 920 K. The value of maximum permissible load (pebble crash limit) at that is low and comprises {approx}37.9 N. The content of residual tritium is higher for ceramic irradiated at 760 K (6.6 {+-} 0.6 x 10{sup 11} Bq/kg) than for ceramic irradiated at 920 K (17 {+-} 3 x 10{sup 10} Bq/kg). The size change indicates that pebble increase more after irradiation at 760 K than at 920 K where the bigger portion of tritium leaves the pebble. X-ray analysis shows radiation modification of Li{sub 2}TiO{sub 3} + 5 mol% TiO{sub 2} phase composition and generation of new phases: LiTi{sub 2}O{sub 4}, LiTiO{sub 2} and Li{sub 4}Ti{sub 5}O{sub 12}.

  1. The effect of neutron irradiation defects on electrical resistivity in FZ-silicon samples irradiated at Es-Salam research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbaci, M. [Centre de Recherche Nucleaire de Birine, BP 180, Ain Oussera W. Djelfa 17200 (Algeria)]. E-mail: abbacim@yahoo.fr; Meglali, O. [Centre de Recherche Nucleaire de Birine, BP 180, Ain Oussera W. Djelfa 17200 (Algeria); Saim, A. [Centre de Recherche Nucleaire de Birine, BP 180, Ain Oussera W. Djelfa 17200 (Algeria); Osmani, N. [Centre de Recherche Nucleaire de Birine, BP 180, Ain Oussera W. Djelfa 17200 (Algeria); Doghmane, N. [Departement d' Electronique, Faculte des Sciences de l' Ingenieur, Universite Badji Mokhtar, BP 12, Annaba 23000 (Algeria)

    2006-09-15

    This paper reports the effect of neutron irradiation defects on electrical properties of n-type FZ-silicon via the measurements of electrical resistivity. For this purpose, FZ-silicon single crystal was irradiated with neutron fluences ranging between 1.54 x 10{sup 16} and 2.5 x 10{sup 18} cm{sup -2}. The samples irradiated at F {sub 1} = 1.54 x 10{sup 16} cm{sup -2}, F {sub 2} = 7.43 x 10{sup 17} cm{sup -2} and F {sub 3} = 2.5 x 10{sup 18} cm{sup -2} were isochronally annealed from room temperature up to 750 {sup o}C. It is found that, for fluences ranging respectively from 1.54 x 10{sup 16} cm{sup -2} to 1.23 x 10{sup 17} cm{sup -2} (stage I) and from 3.09 x 10{sup 17} cm{sup -2} to 2.5 x 10{sup 18} cm{sup -2} (stage II) the resistivity is linearly related to the neutron fluence with two different slopes. The annealing temperature dependence on the electrical resistivity fits well the relation{rho} {sub 0} exp(-CT), where C is a constant depending on the neutron fluence and {rho} {sub 0} is approximately equal to the resistivity after irradiation. For annealing temperatures higher than 550 {sup o}C, we have found that the resistivity is a decreasing function with respect to the neutron fluence and the transmutation-doped phosphorus atoms become electrically active.

  2. Experimental Study of Neutron Irradiation Effects on GaN HEMT Devices%GaN HEMT器件中子辐照效应实验研究

    Institute of Scientific and Technical Information of China (English)

    王燕萍; 罗尹虹; 张科营; 王园明

    2011-01-01

    The online testing techniques and the experimental methods of neutron radiation effects on GaN HEMT (high electron mobility transistor) devices have been established, and the radiation effects experiments of pulsed neutron reactor on GaN HEMT devices were carried out. The emphasis was on device performance degradation caused by ionizing radiation and displacement damage. The GaN HEMT neutron displacement damage sensitive parameters and effected pattern were obtained. The threshold voltage, gate leakage current and leakage current as the sensitive parameters of the neutron irradiation damage have been validated by experiments, and the degradation of the device performance of various radiation damage mechanism has been discussed.%建立了GaN HEMT器件(氮化镓高电子迁移率晶体管)中子原位测试技术和辐照效应实验方法,开展了GaN HEMT器件脉冲反应堆中子辐照效应实验研究,重点研究了电离辐射和位移损伤对器件性能退化的影响,获取了GaN HEMT中子位移损伤效应敏感参数和效应规律.结果表明,阈值电压、栅极泄漏电流以及漏极电流是中子辐照损伤的敏感参数,讨论了器件性能退化的各种辐射损伤机制.

  3. Study of Li 2TiO 3 + 5 mol% TiO 2 lithium ceramics after long-term neutron irradiation

    Science.gov (United States)

    Chikhray, Y.; Shestakov, V.; Maksimkin, O.; Turubarova, L.; Osipov, I.; Kulsartov, T.; Kuykabayeba, A.; Tazhibayeva, I.; Kawamura, H.; Tsuchiya, K.

    2009-04-01

    Given work presents the results of complex material-science studies of 1 mm diameter ceramic pebbles manufactured of Li 2TiO 3 + 5 mol% TiO 2 ceramics before and after long-time neutron irradiation. Ceramic samples were placed in specially ampoules (six items) made of stainless steel Cr18Ni10Ti which were vacuumized and filled with helium. Irradiation of ampoules was carried out in the loop channel of WWRK reactor (Almaty, Kazakhstan) during 223 days at 6 MW power. After irradiation light-colored pebbles became grey-colored due to structure changes which generation of grey-colored inclusions (lithium oxide) with low density and microhardness. There is a radiation softening of lithium ceramic and that effect is higher for lower irradiation temperature 760 K than for 920 K. The value of maximum permissible load (pebble crash limit) at that is low and comprises ˜37.9 N. The content of residual tritium is higher for ceramic irradiated at 760 K (6.6 ± 0.6 × 10 11 Bq/kg) than for ceramic irradiated at 920 K (17 ± 3 × 10 10 Bq/kg). The size change indicates that pebble increase more after irradiation at 760 K than at 920 K where the bigger portion of tritium leaves the pebble. X-ray analysis shows radiation modification of Li 2TiO 3 + 5 mol% TiO 2 phase composition and generation of new phases: LiTi 2O 4, LiTiO 2 and Li 4Ti 5O 12.

  4. Molybdenum sulfide/carbide catalysts

    Science.gov (United States)

    Alonso, Gabriel; Chianelli, Russell R.; Fuentes, Sergio; Torres, Brenda

    2007-05-29

    The present invention provides methods of synthesizing molybdenum disulfide (MoS.sub.2) and carbon-containing molybdenum disulfide (MoS.sub.2-xC.sub.x) catalysts that exhibit improved catalytic activity for hydrotreating reactions involving hydrodesulfurization, hydrodenitrogenation, and hydrogenation. The present invention also concerns the resulting catalysts. Furthermore, the invention concerns the promotion of these catalysts with Co, Ni, Fe, and/or Ru sulfides to create catalysts with greater activity, for hydrotreating reactions, than conventional catalysts such as cobalt molybdate on alumina support.

  5. U.S. Contribution 1994 Summary Report Task T12: Compatibility and irradiation testing of vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L. [comp.

    1995-03-01

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication.

  6. Transmission electron microscopy investigation of the microstructure of Fe–Cr alloys induced by neutron and ion irradiation at 300 °C

    Energy Technology Data Exchange (ETDEWEB)

    Hernández-Mayoral, M., E-mail: m.mayoral@ciemat.es [CIEMAT, Division of Structural Materials, Avenida Complutense, 40, 28040 Madrid (Spain); Heintze, C. [HZDR, Institute of Ion Beam Physics and Materials Research, Bautzner Landstrasse 400, 01328 Dresden (Germany); Oñorbe, E. [CIEMAT, Division of Structural Materials, Avenida Complutense, 40, 28040 Madrid (Spain)

    2016-06-15

    Four Fe–Cr binary alloys, with Cr content from 2.5 up to 12wt%, were neutron or ion irradiated up to a dose of 0.6 dpa at 300 °C. The microstructural response to irradiation has been characterised using Transmission Electron Microscopy (TEM). Both, neutrons and ions, gave rise to the formation of dislocation loops. The most striking difference between ion and neutron irradiation is the distribution of these loops in the sample. Except for the lowest Cr content, loops are distributed mainly along grain boundaries and dislocations in the neutron irradiated samples. The inhomogeneous distribution of dislocation loops could be related to the presence of α′ precipitates in the matrix. In contrast, a homogeneous distribution is observed in all ion irradiated samples. This important difference is attributed to the orders of magnitude difference in dose rate between these two irradiation conditions. Moreover, the density of loops depends non-monotonically on Cr content in case of neutron irradiation, while it seems to increase with Cr content for ion implantation. Differences are also observed in terms of cluster size, with larger sizes for neutron irradiation than for ion implantation, again pointing towards an effect of the dose rate. - Highlights: • Fe–Cr binary alloys were irradiated with neutrons and ions at 300 °C, to 0.6 and 0.5 dpa, respectively. Both, neutrons and ions, gave rise to the formation of dislocation loops. • An effect of Cr is observed in the microstructural response to neutron irradiation and the main difference between alloys consists in the distribution of the dislocation loops throughout the material. While loops appeared homogeneously distributed in Fe–2.5Cr, loops appeared preferentially close to grain boundaries and dislocation lines in the alloys with higher Cr content. • Ion irradiation in similar conditions in terms of dpa resulted in a different damaged microstructure, mainly characterized by a spatially homogeneous

  7. Gel Fabrication of Molybdenum “Beads”

    Energy Technology Data Exchange (ETDEWEB)

    Lowden, Richard Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Armstrong, Beth L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Cooley, Kevin M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-11-01

    Spherical molybdenum particles or “beads” of various diameters are of interest as feedstock materials for the additive manufacture of targets and assemblies used in the production of 99Mo medical isotopes using accelerator technology. Small metallic beads or ball bearings are typically fabricated from wire; however, small molybdenum spheres cannot readily be produced in this manner. Sol-gel processes are often employed to produce small dense microspheres of metal oxides across a broad diameter range that in the case of molybdenum could be reduced and sintered to produce metallic spheres. These Sol-gel type processes were examined for forming molybdenum oxide beads; however, the molybdenum trioxide was chemically incompatible with commonly used gelation materials. As an alternative, an aqueous alginate process being assessed for the fabrication of oxide spheres for catalyst applications was employed to form molybdenum trioxide beads that were successfully reduced and sintered to produce small molybdenum spheres.

  8. HYDROGEN VACANCY INTERACTION IN MOLYBDENUM

    NARCIS (Netherlands)

    Abd El Keriem, M.S.; van der Werf, D.P.; Pleiter, F

    1993-01-01

    Vacancy-hydrogen interaction in molybdenum was investigated by means of the perturbed angular correlation technique, using the isotope In-111 as a probe. The complex InV2 turned out to trap up to two hydrogen atoms: trapping of a single hydrogen atom gives rise to a decrease of the quadrupole

  9. Volume change of Al/sub 2/O/sub 3/ and MgAl/sub 2/O/sub 4/ induced by 14-MeV neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Tanimura, Katsumi; Itoh, Noriaki; Clinard, F.W.

    1987-10-01

    Volume change of Al/sub 2/O/sub 3/ and MgAl/sub 2/O/sub 4/ induced by irradiation with 14-MeV neutrons at 50/sup 0/C has been measured. It is shown that the volume change of Al/sub 2/O/sub 3/ is anisotropic and is larger than that of MgAl/sub 2/O/sub 4/ about a factor of five. The result for MgAl/sub 2/O/sub 4/ is compared with that of fission neutron irradiation.

  10. DLTS and capacitance transients study of defects induced by neutron irradiation in MOS structures CCD process; Etude des defauts induits par irradiation neutron dans des structures MOS par spectroscopie DLTS et reponses transitoires de capacite en fonction du temps

    Energy Technology Data Exchange (ETDEWEB)

    Ahaitouf, A.; Losson, E.; Charles, J.P. [Metz Univ., Lab. Interfaces Composants et Microelectronique, LICM/CLOES/ Supelec, 57 (France)

    1999-07-01

    The aim of this paper is to study neutron irradiation effects on PMOS capacitors and NMOSFETs transistors. The characterization of induced defects was made by capacitance transients C(t) measurements, DLTS spectroscopy, and optical DLTS (ODLTS). DLTS spectra present three peaks due to deep levels created in the semiconductor and two peaks due to minority carrier generation. Two levels are reported in the literature. Two other minority carrier traps have been observed on ODLTS spectra after irradiation. This can explain the decrease of the minority carrier generation lifetime observed by capacitance transients measurements. (authors)

  11. Mechanical Property of China A508-3 Steel after Neutron Irradiation%国产 A508-3钢辐照性能

    Institute of Scientific and Technical Information of China (English)

    林赟; 宁广胜; 张长义; 佟振峰; 杨文

    2016-01-01

    Reactor pressure vessel (RPV ) is the critical un‐changeable component of the reactor during its service lifetime , w hich prevents the radioactive leak of the nuclear power plant core .The irradiation test (about 10 × 1019 cm -2 ,E≥1 MeV) of the pres‐sure vessel material of China A508‐3 steel in research reactor was carried out ,and the mechanics performance tests were carried out after the neutron irradiation ,including tensile property and impact property .The results show that the yield strength increases by 83 ,108 and 52 MPa ,and the tensile strength increases by 58 ,61 and 49 MPa at-100 ,20 and 288 ℃ , respectively . The ductile‐brittle transition temperature T41J increases by 68 ℃ ,and the upper shelf energy decreases by 61 J .Meanwhile ,by compa‐ring the property of un‐irradiated and irradiated material ,after irradiated to the level of 60 a service life ,A508‐3 steel still meets the reactor operation requirement .%反应堆压力容器(RPV )作为反应堆寿期内不可更换的核心设备,是防止堆芯放射性泄漏的最主要屏障。本文针对国产压力容器材料A508‐3钢,开展了一定剂量水平(约10×1019 cm-2,E≥1 M eV )的研究堆加速辐照试验,并进行了辐照后力学性能测试分析,包括拉伸性能和冲击性能测试。结果显示,辐照后在-100、20、288℃下,A508‐3钢的屈服强度分别增加了83、108、52 M Pa ,抗拉强度分别增加了58、61、49 M Pa ,韧脆转变温度 T41J增加了68℃,上平台能量降低了61 J。A508‐3钢辐照前后性能测试结果表明,在中子辐照至60 a寿期后,A508‐3钢仍能满足反应堆使用要求。

  12. Bystander effect-induced mutagenicity in HPRT locus of CHO cells following BNCT neutron irradiation: Characteristics of point mutations by sequence analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kinashi, Yuko [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)], E-mail: kinashi@rri.kyoto-u.ac.jp; Suzuki, Minoru; Masunaga, Shinichiro; Ono, Koji [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)

    2009-07-15

    To investigate bystander mutagenic effects induced by alpha particles during boron neutron capture therapy (BNCT), we mixed cells that were electroporated with borocaptate sodium (BSH), which led to the accumulation of {sup 10}B inside the cells, with cells that did not contain the boron compound. BSH-containing cells were irradiated with {alpha} particles produced by the {sup 10}B(n,{alpha}){sup 7}Li reaction, whereas cells without boron were only affected by the {sup 1}H(n,{gamma}){sup 2}H and {sup 14}N(n,{rho}){sup 14}C reactions. The frequency of mutations induced in the hypoxanthine-guanine phosphoribosyltransferase (HPRT) locus was examined in Chinese hamster ovary (CHO) cells irradiated with neutrons (Kyoto University Research Reactor: 5 MW). Neutron irradiation of 1:1 mixtures of cells with and without BSH resulted in a survival fraction of 0.1, and the cells that did not contain BSH made up 99.4% of the surviving cell population. Using multiplex polymerase chain reactions (PCRs), molecular structural analysis indicated that most of the mutations induced by the bystander effect were point mutations and that the frequencies of total and partial deletions induced by the bystander effect were lower than those resulting from the {alpha} particles produced by the {sup 10}B(n,{alpha}){sup 7}Li reaction or the neutron beam from the {sup 1}H(n,{gamma}){sup 2}H and {sup 14}N(n,{rho}){sup 14}C reactions. The types of point mutations induced by the BNCT bystander effect were analyzed by cloning and sequencing methods. These mutations were comprised of 65.5% base substitutions, 27.5% deletions, and 7.0% insertions. Sequence analysis of base substitutions showed that transversions and transitions occurred in 64.7% and 35.3% of cases, respectively. G:C{yields}T:A transversion induced by 8-oxo-guanine in DNA occurred in 5.9% of base substitution mutants in the BNCT bystander group. The characteristic mutations seen in this group, induced by BNCT {alpha} particles

  13. Defects annihilation behavior of neutron-irradiated SiC ceramics densified by liquid-phase-assisted method after post-irradiation annealing

    Directory of Open Access Journals (Sweden)

    Mohd Idzat Idris

    2016-12-01

    Full Text Available Numerous studies on the recovery behavior of neutron-irradiated high-purity SiC have shown that most of the defects present in it are annihilated by post-irradiation annealing, if the neutron fluence is less than 1×1026 n/m2 (>0.1MeV and the irradiation is performed at temperatures lower than 973K. However, the recovery behavior of SiC fabricated by the nanoinfiltrated and transient eutectic phase (NITE process is not well understood. In this study, the effects of secondary phases on the irradiation-related swelling and recovery behavior of monolithic NITE-SiC after post-irradiation annealing were studied. The NITE-SiC specimens were irradiated in the BR2 reactor at fluences of up to 2.0–2.5×1024 n/m2 (E>0.1MeV at 333–363K. This resulted in the specimens swelling up ∼1.3%, which is 0.1% higher than the increase seen in concurrently irradiated high-purity SiC. The recovery behaviors of the specimens after post-irradiation thermal annealing were examined using a precision dilatometer; the specimens were heated at temperatures of up to 1673K using a step-heating method. The recovery curves were analyzed using a first-order model, and the rate constants for each annealing step were obtained to determine the activation energy for volume recovery. The NITE-A specimen (containing 12 wt% sintering additives recovered completely after annealing at ∼1573K; however, it shrank because of the volatilization of the oxide phases at 1673K. The NITE-B specimen (containing 18wt% sintering additives did not recover fully, since the secondary phase (YAG was crystallized during the annealing process. The recovery mechanism of NITE-A SiC was based on the recombination of the C and Si Frenkel pairs, which were very closely sited or only slightly separated at temperatures lower than 1223K, as well as the recombination of the slightly separated C Frenkel pairs and the migration of C and Si interstitials at temperatures of 1223–1573K. That is to say, the

  14. Neutron irradiation of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Ermi, R.M. [Pacific Northwest National Lab., Richland, WA (United States); Tsai, H. [Argonne National Lab., IL (United States)

    1998-03-01

    Seven subcapsules from the FFTF/MOTA 2B irradiation experiment containing 97 or 100% dense sintered beryllium cylindrical specimens in depleted lithium have been opened and the specimens retrieved for postirradiation examination. Irradiation conditions included 370 C to 1.6 {times} 10{sup 22} n/cm{sup 2}, 425 C to 4.8 {times} 10{sup 22} n/cm{sup 2}, and 550 C to 5.0 {times} 10{sup 22} n/cm{sup 2}. TEM specimens contained in these capsules were also retrieved, but many were broken. Density measurements of the cylindrical specimens showed as much as 1.59% swelling following irradiation at 500 C in 100% dense beryllium. Beryllium at 97% density generally gave slightly lower swelling values.

  15. The involvement of molybdenum in life.

    Science.gov (United States)

    Williams, R J P; Fraústo da Silva, J J R

    2002-03-29

    Quite extraordinarily molybdenum is an essential element in life for the uptake of nitrogen from both nitrogen gas and nitrate, yet it is a relatively rare heavy trace element. It also functions in a few extremely important oxygen-atom transfer reactions at low redox potential. This review poses the question "Why does life depend upon molybdenum?" The answer has to be based upon the availability of the element and on chemical superiority in carrying out the essential tasks. We illustrate here the peculiarities of molybdenum chemistry and how they have become part of certain enzymes. The uptake and incorporation of molybdenum are dependent on its availability, selective pumps, and carriers (chaperones), but 4.5 x 10(9) years ago molybdenum was not available when both tungsten and vanadium or even iron were possibly used in its place. While these possibilities are explored, they leave many unanswered questions concerning the selection today of molybdenum. (c)2002 Elsevier Science (USA).

  16. [Molybdenum as an air pollutant].

    Science.gov (United States)

    Lindner, R; Junker, E; Hoheiser, H

    1990-07-01

    Investigations into the reasons for the retarded growth and discolouration of a small area of a field of rape situated on the outskirts of Vienna revealed higher than normal levels of molybdenum in the soil (up to 430 micrograms/l) and in the water (up to 9.7 mg/l). The source of the pollution was traced to a neighbouring industrial plant that was emitting the metal via the chimney stack. A review of the literature on the toxic effects of molybdenum in general and as an air pollutant in particular is provided. This shows that, in contrast to animals, this effect is relatively small in humans and plants. Nevertheless, the occupation-related inhalation of the metal has been shown to be associated with pneumoconiosis and gout-like symptoms.

  17. Zirconia-molybdenum disilicide composites

    Science.gov (United States)

    Petrovic, John J.; Honnell, Richard E.

    1991-01-01

    Compositions of matter comprised of molybdenum disilicide and zirconium oxide in one of three forms: pure, partially stabilized, or fully stabilized and methods of making the compositions. The stabilized zirconia is crystallographically stabilized by mixing it with yttrium oxide, calcium oxide, cerium oxide, or magnesium oxide and it may be partially stabilized or fully stabilized depending on the amount of stabilizing agent in the mixture.

  18. Effect of molybdenum addition on microstructure and mechanical properties of plain carbon steel weld

    Directory of Open Access Journals (Sweden)

    Jyoti Menghani

    2016-12-01

    Full Text Available The present investigation has two main objectives; first is optimization of welding process parameters of submerged arc welding (SAW using Taguchi philosophy and second is to improve the mechanical properties such as strength and microhardness of weld joint by alloying with varying amounts of molybdenum. For optimization of welding process, parameters Taguchi philosophy have been applied on a mild steel plate (AISI C- 1020 of 10 mm thickness with 60o groove angle with arc voltage and welding speed as variables and bead width as output variables. A mathematical relationship between bead width, arc voltage and welding speed has also been found using multiple regression analysis for the present base metal plate geometry. After optimizing welding parameters, molybdenum has been added individually to the welding area in varying percentages. The properties of alloyed and unalloyed weld metal bead are compared. The mechanical characterization of weld has been done in terms of microhardness, tensile strength, whereas microstructural characterization has been performed using optical microscopy, XRD and EDS. The presence of molybdenum resulted in bainite structure in weld bead having a refined grain structure, enhancement in tensile strength and microhardness. The XRD results showed the formation of molybdenum carbides justifying the increase in microhardness value.

  19. Austenite Grain Growth and Precipitate Evolution in a Carburizing Steel with Combined Niobium and Molybdenum Additions

    Science.gov (United States)

    Enloe, Charles M.; Findley, Kip O.; Speer, John G.

    2015-11-01

    Austenite grain growth and microalloy precipitate size and composition evolution during thermal processing were investigated in a carburizing steel containing various additions of niobium and molybdenum. Molybdenum delayed the onset of abnormal austenite grain growth and reduced the coarsening of niobium-rich precipitates during isothermal soaking at 1323 K, 1373 K, and 1423 K (1050 °C, 1100 °C, and 1150 °C). Possible mechanisms for the retardation of niobium-rich precipitate coarsening in austenite due to molybdenum are considered. The amount of Nb in solution and in precipitates at 1373 K (1100 °C) did not vary over the holding times evaluated. In contrast, the amount of molybdenum in (Nb,Mo)C precipitates decreased with time, due to rejection of Mo into austenite and/or dissolution of fine Mo-rich precipitates. In hot-rolled alloys, soaking in the austenite regime resulted in coarsening of the niobium-rich precipitates at a rate that exceeded that predicted by the Lifshitz-Slyozov-Wagner relation for volume-diffusion-controlled coarsening. This behavior is attributed to an initial bimodal precipitate size distribution in hot-rolled alloys that results in accelerated coarsening rates during soaking. Modification of the initial precipitate size distribution by thermal processing significantly lowered precipitate coarsening rates during soaking and delayed the associated onset of abnormal austenite grain growth.

  20. Reduction of molybdenum oxide from steelmaking slags by pure liquid iron

    Directory of Open Access Journals (Sweden)

    Gao Y.M.

    2012-01-01

    Full Text Available The effects of reaction temperature, slag basicity and FeO concentration on the reduction of molybdenum oxide from steelmaking slags by pure liquid iron were investigated experimently. The reduction kinetics of molybdenum oxide by liquid iron was analysed. The reaction models were developed based on the condition that diffusion of [Mo] in liquid iron and CaMoO4 in slag is the control steps, respectively. These reaction models were tested using data from a series of experiments. The results indicate that under the present experimental conditions, the temperature and the FeO content, other than slag basicity, have some effects on the reduction of molybdenum oxide from steelmaking slags by pure liquid iron. Both the molybdenum oxide reduction rate and final reduction ratio increase with an increase of temperature and a decrease of FeO content. The diffusion of CaMoO4 in slag which dominated overall reduction process is the only one ratecontrolling step with its apparent activation energy 294 kJ/mol. The reduction of molybdenum oxide used directly as alloy additive can be further enhanced by strong stirring in the converter practice.

  1. The development of a tensile-shear punch correlation for yield properties of model austenitic alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hankin, G.L.; Faulkner, R.G. [Loughborough Univ. (United Kingdom); Hamilton, M.L.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-08-01

    The effective shear yield and maximum strengths of a set of neutron-irradiated, isotopically tailored austentic alloys were evaluated using the shear punch test. The dependence on composition and neutron dose showed the same trends as were observed in the corresponding miniature tensile specimen study conducted earlier. A single tensile-shear punch correlation was developed for the three alloys in which the maximum shear stress or Tresca criterion was successfully applied to predict the slope. The correlation will predict the tensile yield strength of the three different austenitic alloys tested to within {+-}53 MPa. The accuracy of the correlation improves with increasing material strength, to within {+-} MPa for predicting tensile yield strengths in the range of 400-800 MPa.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  3. Fabrication using a levitation melting method of V-4Cr-4Ti-Si-Al-Y alloys and their mechanical properties

    Science.gov (United States)

    Chuto, Toshinori; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori; Nagasaka, Takuya; Muroga, Takeo

    2002-12-01

    Reduction of interstitial impurities such as O and N is a potential method to improve various properties of vanadium alloys. It was shown that addition of Si, Al and Y was useful to reduce the oxygen concentration and to improve post-irradiation ductility at relatively low temperatures for V-Cr-Ti alloys. Several 2.5 kg alloys of V-4Cr-4Ti-Si-Al-Y type were fabricated by using a levitation melting method. Charpy impact test by an instrumented testing machine has been conducted using miniaturized specimens. Tensile tests have been carried out before and after neutron irradiation. The miniaturized specimens were irradiated up to 8×10 22 n/m 2 ( E>1 MeV) at 290 °C in Japan Materials Testing Reactor. By adopting a levitation melting method, high-purity kg-scale ingots of V-4Cr-4Ti-Si-Al-Y alloys with ˜80 ppm C, <170 ppm O and ˜110 ppm N were obtained. The V-4Cr-4Ti-0.1Si-0.1Al-0.1Y alloy fabricated in this study showed good impact properties compared with a previous laboratory-scale alloy. This alloy showed good tensile properties even after neutron irradiation at 290 °C. Levitation melting can be adopted to produce large ingots of V-Cr-Ti-Si-Al-Y type alloys by controlling the amount of yttrium addition. In this study, the technology for fabrication of high-purity kg-scale ingots of V-4Cr-4Ti-Si-Al-Y alloy has been demonstrated, and has made it possible to investigate systematically various properties of the alloy.

  4. Environmentally Assisted Cracking of Nickel Alloys - A Review

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, R

    2004-07-12

    Nickel can dissolve a large amount of alloying elements while still maintaining its austenitic structure. That is, nickel based alloys can be tailored for specific applications. The family of nickel alloys is large, from high temperature alloys (HTA) to corrosion resistant alloys (CRA). In general, CRA are less susceptible to environmentally assisted cracking (EAC) than stainless steels. The environments where nickel alloys suffer EAC are limited and generally avoidable by design. These environments include wet hydrofluoric acid and hot concentrated alkalis. Not all nickel alloys are equally susceptible to cracking in these environments. For example, commercially pure nickel is less susceptible to EAC in hot concentrated alkalis than nickel alloyed with chromium (Cr) and molybdenum (Mo). The susceptibility of nickel alloys to EAC is discussed by family of alloys.

  5. Nuclear fuel alloys or mixtures and method of making thereof

    Science.gov (United States)

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  6. Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bruemmer; Peter L. Andersen; Gary Was

    2002-12-27

    The discovery of a damage-resistant alloy based on Hf solute additions to a low-carbon 316SS is the highlight of the Phase II research. This damage resistance is supported by characterization of radiation-induced microstructures and microchemistries along with measurements of environmental cracking. The addition of Hf to a low-carbon 316SS reduced the detrimental impact of radiation by changing the distribution of Hf. Pt additions reduced the impact of radiation on grain boundary segregation but did not alter its effect on microstructural damage development or cracking. Because cracking susceptibility is associated with several material characteristics, separate effect experiments exploring strength effects using non-irradiated stainless steels were conducted. These crack growth tests suggest that irradiation strength by itself can promote environmental cracking. The second concept for developing damage resistant alloys is the use of metastable precipitates to stabilize the microstructure during irradiation. Three alloys have been tailored for evaluation of precipitate stability influences on damage evolution. The first alloy is a Ni-base alloy (alloy 718) that has been characterized at low neutron irradiation doses but has not been characterized at high irradiation doses. The other two alloys are Fe-base alloys (PH 17-7 and PH 17-4) that have similar precipitate structures as alloy 718 but is more practical in nuclear structures because of the lower Ni content and hence lesser transmutation to He.

  7. Sodium corrosion behavior of austenitic alloys and selective dissolution of chromium and nickel

    Science.gov (United States)

    Suzuki, T.; Mutoh, I.; Yagi, T.; Ikenaga, Y.

    1986-06-01

    The corrosion behavior of six austenitic alloys and reference Type 316 stainless steel (SS) has been examined in a flowing sodium environment at 700°C for up to about 4000 h. The alloys with a range of nickel content between ~ 15 and 43 wt% were designed and manufactured with an expectation of improved swelling resistance during fast neutron irradiation, compared to reference Type 316 SS. The corrosion loss of the alloys at zero downstream position and the concentrations of chromium, nickel and iron in the surface region were determined as a function of corrosion time. The selective dissolution of nickel and chromium played an important role in sodium corrosion of the alloys. During the initial period, accelerated corrosion took place and selective dissolution of chromium and nickel proceeded at a rapid rate. During the subsequent period, the overall corrosion rate and depletion of chromium and nickel decreased with increasing time until the corrosion rate and the surface concentrations of chromium, nickel and iron, which depended on composition of the alloys, reached the steady-state after about 2000 h. Also, the corrosion rate increased with increasing original nickel content of the alloys. Microstructural examination revealed surface attack of the alloys with higher nickel contents, in particular for the two precipitation strengthened Fe-Ni alloys. The alloys showed a trend of increasing carbon and nitrogen contents.

  8. Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bruemmer; Peter L. Andersen; Gary Was

    2002-12-27

    The discovery of a damage-resistant alloy based on Hf solute additions to a low-carbon 316SS is the highlight of the Phase II research. This damage resistance is supported by characterization of radiation-induced microstructures and microchemistries along with measurements of environmental cracking. The addition of Hf to a low-carbon 316SS reduced the detrimental impact of radiation by changing the distribution of Hf. Pt additions reduced the impact of radiation on grain boundary segregation but did not alter its effect on microstructural damage development or cracking. Because cracking susceptibility is associated with several material characteristics, separate effect experiments exploring strength effects using non-irradiated stainless steels were conducted. These crack growth tests suggest that irradiation strength by itself can promote environmental cracking. The second concept for developing damage resistant alloys is the use of metastable precipitates to stabilize the microstructure during irradiation. Three alloys have been tailored for evaluation of precipitate stability influences on damage evolution. The first alloy is a Ni-base alloy (alloy 718) that has been characterized at low neutron irradiation doses but has not been characterized at high irradiation doses. The other two alloys are Fe-base alloys (PH 17-7 and PH 17-4) that have similar precipitate structures as alloy 718 but is more practical in nuclear structures because of the lower Ni content and hence lesser transmutation to He.

  9. Alloying effects on mechanical and metallurgical properties of NiAl

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C.T.; Horton, J.A.; Lee, E.H.; George, E.P.

    1993-06-01

    Alloying effects were investigated in near-stoichiometric NiAl for improving its mechanical and metallurgical properties. Ternary additions of 19 elements at levels up to 10 at. % were added to NiAl; among them, molybdenum is found to be most effective in improving the room-temperature ductility and high-temperature strength. Alloying with 1.0 {plus_minus} 0.6% molybdenum almost doubles the room-temperature tensile ductility of NiAl and triples its yield strength at 1000C. The creep properties of molybdenum-modified NiAl alloys can be dramatically improved by alloying with up to 1% of niobium or tantalum. Because of the low solubilities of molybdenum and niobium in NiAl, the beneficial effects mainly come from precipitation hardening. Fine and coarse precipitates are revealed by both transmission electron microscopy (TEM) and electron microprobe analyses. Molybdenum-containing alloys possess excellent oxidation resistance and can be fabricated into rod stock by hot extrusion at 900 to 1050C. This study of alloying effects provides a critical input for the alloy design of ductile and strong NiAl aluminide alloys for high-temperature structural applications.

  10. Molybdenum Disilicide Oxidation Kinetics in High Temperature Steam

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Stephen Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-07

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials, or the existing zirconium alloy cladding can be coated with a more oxidation resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi2 was identified based on its high temperature oxidation resistance in O2 atmospheres (e.g. air and “wet air”). However, its behavior in H2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O2) exposures were also performed and those results are presented here for a comparative analysis. It was determined that MoSi2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670-773 K

  11. Molybdenum Disilicide Oxidation Kinetics in High Temperature Steam

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Stephen Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Andrew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-07

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials, or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi2 was identified based on its high temperature oxidation resistance in O2 atmospheres (e.g. air and “wet air”). However, its behavior in H2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670

  12. Creep behavior of Zr-Nb alloys

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Chan; Kim, Young Suk; Cheong, Yong Mu; Kwon, Sang Chul; Kim, Sung Soo; Choo, Ki Nam [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    The creep characteristics of Zirconium alloy is affected by several parameters. Out-reactor creep increases both with an increasing amount of Nb, Sn and S contained in alpha-Zr and decreases with the increasing volume of alpha-Zr. Especially, the creep of Zr-2.5Nb alloy depends on the solubility of Nb in alpha-Zr, which is associated with the decomposition of beta-Zr. Since Zr of the hcp structure is strongly anisotropic, it shows the characteristics of texture and results in the anisotropy of creep. Due to the circumferential texture of Zr-2.5%Nb alloy (CANDU Pressure tube), the longitudinal slip is easier than the circumferential one, resulting in the high creep rate. The irradiation creep also increases with increasing neutron fluence. The neutron irradiation increases the strength of the zirconium alloys but decreases their creep strength. In contrast to the out-reactor creep, the irradiation creep is little sensitive to temperature, resulting in the lower activation energy. The most important factor to affect the in-reactor and out-reactor creep of niobium containing alloys seems to be the solution hardening by Nb or Sn which is soluble in alpha-zirconium and the texture as well. Irradiation growth is the mechanism which is caused only by the irradiation. It becomes saturated at lower fluence than the critical fluence but beyond it, shows the break-away growth. The onset of accelerated irradiation growth corresponds with the c-dislocation loop formation, though its mechanism needs better understanding. Generally, the irradiation growth of Zr-Nb alloys increases with an increase in fluence, cold working, dislocation, density and temperature, and with a decrease in the grain size. 141 refs., 59 figs., 10 tabs. (Author)

  13. Study of macro- and micro-segregation of iridium in molybdenum single crystals after electron beam zone melting

    Energy Technology Data Exchange (ETDEWEB)

    Drapala, Jaromir; Skotnicova, Katerina [VSB-Technical University of Ostrava (Czech Republic). Dept. of Non-ferrous Metals, Refining and Recycling

    2013-01-15

    The aim of the work was to study the creation of micro- and macro-segregation of iridium in low-alloyed molybdenum single crystals after electron beam zone melting (floating zone technique) depending on various conditions of crystallization. In order to evaluate relations between the chemical inhomogeneity and structural defects and their influence on properties of single crystals, the dependence of concentration and character of distribution of admixtures under various crystallization conditions on the origin of concentration undercooling and dislocation substructure of molybdenum single crystals prepared by electron beam floating zone melting was experimentally investigated.

  14. Price Hike in Molybdenum Industrial Chain Picked Up Speed

    Institute of Scientific and Technical Information of China (English)

    2016-01-01

    Affected by industry-wide joint production restriction,prices of molybdenum concentrate,molybdenum oxide and ferromolybdenum surged across the board.Following the average rising margin up to 1%to 2%on May 23,on May 24,prices of molybdenum concentrate and molybdenum oxide again were adjusted upward by 50 yuan per tonne,rising by 5%,which

  15. Decommissioning of an Irradiator MPX-{gamma} - 25M and a neutron Irradiator; Desmantelamiento de un irradiador tipo MPX-{gamma}-25M y de un irradiador de neutrones

    Energy Technology Data Exchange (ETDEWEB)

    Soguero, Dania; Guerra, Mercedes; Prieto, Enrique; Desdin, Luis, E-mail: sdania@ceaden.edu.cu [Centro de Aplicaciones Tecnologica y Desarrollo Nuclear (CEADEN), La Habana (Cuba)

    2013-07-01

    In this paper a technology is developed with its procedures in radiation protection to ensure the safety of the process of decommissioning of two irradiators. Both processes are described, the process of decommissioning of a neutron Irradiator 4. 44{center_dot}10{sup 11}Bq, employed in the vegetal radio mutagenesis, and disassembling of an installation of gamma irradiation of 3.33 * 10{sup 12} Bq, self-shielded of category I, model MPX - {gamma} - 25 M. The specific objectives are: a) identify aspects of the contractual assurance, of human and technical resources, b) to evaluate the radiological situation of the process and c) analyze the potential radiological extraordinary events in each of the steps of the process, ensuring the right answers. Evaluation of radiological successful events described can be considered as reference to address the process of disassembling of other similar irradiators.

  16. Hardness behavior of binary and ternary niobium alloys at 77 and 300 K

    Science.gov (United States)

    Stephens, J. R.; Witzke, W. R.

    1974-01-01

    The effects of alloy additions of zirconium, hafnium, molybdenum, tungsten, rhenium, ruthenium, osmium, rhodium, and iridium on the hardness of niobium was determined. Both binary and ternary alloys were investigated by means of hardness tests at 77 K and 300 K. Results showed that atomic size misfit plays a dominant role in controlling hardness of binary niobium alloys. Alloy softening, which occurred at dilute solute additions, is most likely due to an extrinsic mechanism involving interaction between solute elements and interstitial impurities.

  17. Iron-based amorphous alloys and methods of synthesizing iron-based amorphous alloys

    Science.gov (United States)

    Saw, Cheng Kiong; Bauer, William A.; Choi, Jor-Shan; Day, Dan; Farmer, Joseph C.

    2016-05-03

    A method according to one embodiment includes combining an amorphous iron-based alloy and at least one metal selected from a group consisting of molybdenum, chromium, tungsten, boron, gadolinium, nickel phosphorous, yttrium, and alloys thereof to form a mixture, wherein the at least one metal is present in the mixture from about 5 atomic percent (at %) to about 55 at %; and ball milling the mixture at least until an amorphous alloy of the iron-based alloy and the at least one metal is formed. Several amorphous iron-based metal alloys are also presented, including corrosion-resistant amorphous iron-based metal alloys and radiation-shielding amorphous iron-based metal alloys.

  18. MOLYBDENUM

    African Journals Online (AJOL)

    electron donor), have been found to have 3:4 square pyramidal structures with the cyclopentadienyl .... The relative ratios of the cis to trans ring proton resonances gives the proportion of cis to .... relationship is not obvious by our analysis.

  19. Recovery of molybdenum using alumina microspheres and precipitation with selective organic reagents

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Fatima Maria Sequeira de; Abrao, Alcidio [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Engenharia Quimica e Ambiental]. E-mail: fatimamc@net.ipen.br; aabrao@net.ipen.br

    1998-07-01

    In this paper is presented a study for the optimization of dissolution of the UAL{sub x} plates used for irradiation and production of radiomolybdenum. The alloy is dissolved in nitric acid with mercury as catalyst. The separation and concentration of the molybdenum was achieved using a chromatographic grade alumina microspheres column. the purified eluted molybdenum is finally precipitated using one of the selective reagents: alizarine blue, {alpha},{alpha}'- bipyridine and 1,10-phenanthroline. Any one of the obtained precipitate can be fired to the molybdenum trioxide. The interference of the following elements was studied: Re(VII), U(VI), Cr(VI), W(VI), V(V), Te(IV), Ti(IV), Zr(IV), Th(IV), Fe(III), Au(III), Ru(III), Al(III), Bi(III), Sb(III), Ce(IV), Pr(III), Sc(III), Y(III), Sm(III), Ba(II), Sr(II), Ni(II), Co(II), Cs(I). The molybdenum precipitates were characterized by gravimetric, CHN, TG, DTG, IR and X-ray diffraction analyses. (author)

  20. Spectrophotometric determination of molybdenum after separation by the adsorption of its trifluoroethylxanthate on naphthalene

    Energy Technology Data Exchange (ETDEWEB)

    Hussain, M.F.; Katyal, Mohan; Puri, B.K.; Satake, Masatada

    1986-10-01

    Molybdenum reacts with potassium trifluoroethylxanthate to form a water-insoluble complex in the acid concentration range 0.1-3 M, pH 1.0-3.5. This complex is easily adsorbed on to microcrystalline naphthalene from an acetone solution and absorbs in the range 360-370 nm. Beer's law is obeyed over the concentration range 5.0-75.0 ..mu..g of molybdenum in 10 ml of chloroform solution. The molar absorptivity and Sandell sensitivity are 1.041 x 10/sup 4/ l mol/sup -1/ cm-/sup 11/ and 0.0092 ..mu..g cm/sup -2/, respectively. Ten replicate analyses of a sample solution containing 30 ..mu..g of molybdenum gave a mean absorbance of 0.325 with a relative standard deviation of 0.60%. The interferences of various ions were studied and conditions were developed for the determination of molybdenum in some alloy samples.

  1. Mechanical properties and constitutive relations for molybdenum under high-rate deformation

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S.R.; Maudlin, P.J.; Gray, G.T. III

    1998-01-01

    Molybdenum and its alloys have received increased interest in recent years for ballistic applications. The stress-strain behavior of several molybdenums possessing various compositions, manufacturing sources, and the degree of pre-straining, were investigated as a function of temperature from 77 to 1,273 K, and strain rate from 10{sup {minus}3} s{sup {minus}1} to 8,000 s{sup {minus}1}. The yield stress was found to be sensitive to the test temperature and strain rate, however, the strain hardening remained rate-insensitive. The constitutive response of a powder-metallurgy molybdenum was also investigated; similar mechanical properties compared to conventionally wrought processed molybdenums were achieved. Constitutive relations based upon the Johnson-Cook, the Zerilli-Armstrong and the Mechanical Threshold Stress (MTS) models were evaluated and fit for the various Mo-based materials. The capabilities and limitations of each model for large-strain applications were examined. The differences between the three models are demonstrated using model comparisons to Taylor cylinder validation experiments.

  2. New alloys to conserve critical elements

    Science.gov (United States)

    Stephens, J. R.

    1978-01-01

    Based on availability of domestic reserves, chromium is one of the most critical elements within the U.S. metal industry. New alloys having reduced chromium contents which offer potential as substitutes for higher chromium containing alloys currently in use are being investigated. This paper focuses primarily on modified Type 304 stainless steels having one-third less chromium, but maintaining comparable oxidation and corrosion properties to that of type 304 stainless steel, the largest single use of chromium. Substitutes for chromium in these modified Type 304 stainless steel alloys include silicon and aluminum plus molybdenum.

  3. Handbook of International alloy Compositions and Designations. Volume II. Superalloys

    Science.gov (United States)

    1978-12-01

    Melted Alloys. Type VMA15: Nickel base-10% cobalt 10% tungsten 9% chromium b.5% aluminium 2.5% tantalum 1.5% hafnium 1.5% titanium vacuum melted... Aluminium Nb — Niobium C - Chromium S - Silicon D — Molybdenum T — Titanium Fe- Iron Ta - Tantalum G — Magnesium U — Copper H - Thorium V...chromium- aluminium -tungsten- molybdenum-nlobium alloy castings (Cr 11.0, AI6.0, W3.5,Mo3.0, Nb2.0) M Gr 1 (ISBN: 0 580 07218 5) 0&73

  4. Process for Functionalizing Biomass using Molybdenum Catalysts

    DEFF Research Database (Denmark)

    2015-01-01

    The present invention concerns a process for converting biomass into useful organic building blocks for the chemical industry. The process involves the use of molybdenum catalysts of the formula Aa+a(MovXxR1yR2zR3e)a*3-, which may be readily prepared from industrial molybdenum compounds.......The present invention concerns a process for converting biomass into useful organic building blocks for the chemical industry. The process involves the use of molybdenum catalysts of the formula Aa+a(MovXxR1yR2zR3e)a*3-, which may be readily prepared from industrial molybdenum compounds....

  5. Investigation of hydrogen evolution activity for the nickel, nickel-molybdenum nickel-graphite composite and nickel-reduced graphene oxide composite coatings

    Science.gov (United States)

    Jinlong, Lv; Tongxiang, Liang; Chen, Wang

    2016-03-01

    The nickel, nickel-molybdenum alloy, nickel-graphite and nickel-reduced graphene oxide composite coatings were obtained by the electrodeposition technique from a nickel sulfate bath. Nanocrystalline molybdenum, graphite and reduced graphene oxide in nickel coatings promoted hydrogen evolution reaction in 0.5 M H2SO4 solution at room temperature. However, the nickel-reduced graphene oxide composite coating exhibited the highest electrocatalytic activity for the hydrogen evolution reaction in 0.5 M H2SO4 solution at room temperature. A large number of gaps between 'cauliflower' like grains could decrease effective area for hydrogen evolution reaction in slight amorphous nickel-molybdenum alloy. The synergistic effect between nickel and reduced graphene oxide promoted hydrogen evolution, moreover, refined grain in nickel-reduced graphene oxide composite coating and large specific surface of reduced graphene oxide also facilitated hydrogen evolution reaction.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.

    1996-04-01

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

  7. The influence of molybdenum on stress corrosion in Ultra Low Carbon Steels with copper addition

    Directory of Open Access Journals (Sweden)

    M. Mazur

    2010-07-01

    Full Text Available The influence of molybdenum content on the process of stress corrosion of ultra-low carbon structural steels with the addition of copper HSLA (High Strength Low Alloy was analyzed. The study was conducted for steels after heat treatment consisting of quenching andfollowing tempering at 600°C and it was obtained microstructure of the tempered martensite laths with copper precipitates and the phaseLaves Fe2Mo type. It was found strong influence of Laves phase precipitate on the grain boundaries of retained austenite on rate anddevelopment of stress corrosion processes. The lowest corrosion resistance was obtained for W3 steel characterized by high contents ofmolybdenum (2.94% Mo which should be connected with the intensity precipitate processes of Fe2Mo phase. For steels W1 and W2which contents molybdenum equals 1.02% and 1.88%, respectively were obtained similar courses of corrosive cracking.

  8. Nitrogen diffusion in near-surface range of ion doped molybdenum

    CERN Document Server

    Zamalin, E Y

    2001-01-01

    The dynamics of change in nitrogen near-the-surface concentration in the Mo ion-alloyed monocrystalline foil is studied through the Auger-electron spectroscopy and the secondary ion mass spectrometry. The implantation dose constituted 5 x 10 sup 1 sup 7 ion/cm sup 2 and the implantation energy equaled 50 and 100 keV. The samples diffusion annealing was performed at the temperature of 800-900 deg C. The evaluation of the nitrogen diffusion coefficient indicates the values by 3-5 orders lesser than the diffusion coefficient in the nitrogen solid-state solution in the molybdenum. At the same time the molybdenum self-diffusion coefficient value is by 3-5 orders lesser as compared to the obtained value. The supposition is made, the the surplus nitrogen relative to the solubility limit is deposited on the radiation defects and in the process of the diffusion annealing it nitrates together with them

  9. Microstructures and mechanical properties of compositionally complex Co-free FeNiMnCr{sub 18} FCC solid solution alloy

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Z. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Materials Science and Engineering Department, University of Tennessee, Knoxville, TN 37996 (United States); Bei, H., E-mail: beih@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2015-07-29

    Recently, a structurally-simple but compositionally-complex FeNiCoMnCr high entropy alloy was found to have excellent mechanical properties (e.g., high strength and ductility). To understand the potential of using high entropy alloys as structural materials for advanced nuclear reactor and power plants, it is necessary to have a thorough understanding of their structural stability and mechanical properties degradation under neutron irradiation. This requires us to develop a similar model alloy without Co because material with Co will make post-neutron-irradiation testing difficult due to the production of the {sup 60}Co radioisotope. To achieve this goal, a FCC-structured single-phase alloy with a composition of FeNiMnCr{sub 18} was successfully developed. This near-equiatomic FeNiMnCr{sub 18} alloy has good malleability and its microstructure can be controlled by thermomechanical processing. By rolling and annealing, the as-cast elongated-grained-microstructure is replaced by homogeneous equiaxed grains. The mechanical properties (e.g., strength and ductility) of the FeNiMnCr{sub 18} alloy are comparable to those of the equiatomic FeNiCoMnCr high entropy alloy. Both strength and ductility increase with decreasing deformation temperature, with the largest difference occurring between 293 and 77 K. Extensive twin-bands which are bundles of numerous individual twins are observed when it is tensile-fractured at 77 K. No twin bands are detected by EBSD for materials deformed at 293 K and higher. The unusual temperature-dependencies of UTS and uniform elongation could be caused by the development of the dense twin substructure, twin-dislocation interactions and the interactions between primary and secondary twinning systems which result in a microstructure refinement and hence cause enhanced strain hardening and postponed necking.

  10. Crystallization of Low-alloyed Construction Cast Steel Modified with V and Ti

    Directory of Open Access Journals (Sweden)

    A. Studnicki

    2013-07-01

    Full Text Available In this paper crystallization studies of low-alloyed construction cast steel were presented for different additions of chromium, nickel and molybdenum modified with vanadium and titanium. Studies were conducted using developed TDA stand, which additionally enabled evaluation of cooling rate influence on crystallization process of investigated alloys.

  11. The Influence of Casting Conditions on the Microstructure of As-Cast U-10Mo Alloys: Characterization of the Casting Process Baseline

    Energy Technology Data Exchange (ETDEWEB)

    Nyberg, Eric A.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean M.; Burkes, Douglas

    2013-12-13

    Sections of eight plate castings of uranium alloyed with 10 wt% molybdenum (U-10Mo) were sent from Y-12 to the Pacific Northwest National Laboratory (PNNL) for microstructural characterization. This report summarizes the results from this study.

  12. The Influence of Casting Conditions on the Microstructure of As-Cast U-10Mo Alloys: Characterization of the Casting Process Baseline

    Energy Technology Data Exchange (ETDEWEB)

    Nyberg, Eric A.; Joshi, Vineet V.; Lavender, Curt A.; Paxton, Dean M.; Burkes, Douglas

    2013-12-13

    Sections of eight plate castings of uranium alloyed with 10 wt% molybdenum (U-10Mo) were sent from Y-12 to the Pacific Northwest National Laboratory (PNNL) for microstructural characterization. This report summarizes the results from this study.

  13. Molybdenum Metallopharmaceuticals Candidate Compounds - The "Renaissance" of Molybdenum Metallodrugs?

    Science.gov (United States)

    Jurowska, Anna; Jurowski, Kamil; Szklarzewicz, Janusz; Buszewski, Boguslaw; Kalenik, Tatiana; Piekoszewski, Wojciech

    2016-01-01

    Metal-based drugs, also called "metallopharmaceuticals" or "metallodrugs", are examples of sophisticated compounds that have been used in inorganic medicinal chemistry as therapeutic agents for a long time. Few of them have shown substantially promising results and many of them have been used in different phases of clinical trials. The Mo-based metallodrugs were successfully applied in the past for treating conditions such as anemia or Wilson's disease. Moreover, Mo complexes are supposed to exert their effect by intercalation/ cleavage of DNA/RNA, arrest of the cell cycle, and alteration of cell membrane functions. However, in the current literature, there are no reliable and in-depth reviews about the hypothetical therapeutic applications of all of the known molybdenum complexes as metallopharmaceuticals/ metallodrugs. The main emphasis was on the in-depth review of the potential applications of Mo-based complexes in medicinal chemistry as metallopharmaceuticals in treating diseases such as cancer and tumors, Wilson's disease, diabetes mellitus, Huntington's disease, atherosclerosis, and anemia. It must be emphasized that today the development of innovative and new Mo-based metalo-pharmaceuticals is not rapid, and hence the aim of this paper was also to inspire colleagues working in the field of Mo compounds who are trying to find "signpost" for research. The authors hope that this article will increase interest and initiate the Renaissance of Mo-compounds among medicinal inorganic chemists. This paper is the first review article in the literature that refers to and emphasizes many different and complex aspects of possible applications and capabilities of Mo-based metallodrugs.

  14. Preparation of selective molybdenum concentrate from collective coppermolybdenum concentrate

    Directory of Open Access Journals (Sweden)

    N. Tusupbaev

    2016-06-01

    Full Text Available The paper considers possibilities of selective separation of the concentrate of copper and molybdenum from a collective copper-molybdenum concentrate of Aktogay deposit using regrinding and conventional flotation reagents. In the case of conventional flotoreagents, the content of molybdenum in a molybdenum concentrate was 8.0% at extraction effectiveness 83.12%. At 27.96% extraction degree of copper, it’s content in the concentrate equaled to 21.3%. After regrinding, molybdenum content in the concentrate was 24.0% at the extraction effectiveness 59.63%, and copper content in the concentrate was 21.9% at the recovery of 61.23%. Thus, the regrinding of a collective copper-molybdenum concentrate resulted in an increase in the content of molybdenum in molybdenum concentrate by 16%, and the copper concentration increased by 0.6%.

  15. Zunyi Molybdenum & Nickel Mining Enterprises Are Still in Suspension Status

    Institute of Scientific and Technical Information of China (English)

    2015-01-01

    According to the 2015 mid-year report of Tiancheng Holding,because Guizhou Province is still enforcing policy regulation&environmental; protection policy for molybdenum&nickel; mining industry,currently all molybdenum and nickel mining

  16. FRACTURE TOUGHNESS OF 6.4 MM (0.25 INCH) ARC-CAST MOLOBDENUM AND MOLYBDENUM-TZM PLATE AT ROOM TEMPERATURE AND 300 DEGREES C

    Energy Technology Data Exchange (ETDEWEB)

    J. A. SHIELDS, JR.; P. LIPETZKY; A. J. MUELLER

    2001-04-11

    THE FRACTURE TOUGHNESS OF 6.4 mm (0.25 INCH) LOW CARBON ARC-CAST (LCAC) MOLYBDENUM AND ARC-CAST MOLYBDENUM-TZM ALLOY PLATE WERE MEASURED AT ROOM TEMPERATURE AND 300 DEGREES C USING COMPACT TNESION SPECIMENTS. THE EFFECT OF CRACK PLANE ORIENTATION (LONGITUDINAL VS. TRANSVERSE) AND ANNEALING PRACTICE (STRESS-RELIEVED VS. RECRYSTALLIZED) WERE EVALUATED. DEPENDING UPON THE TEST TEMPERATURE EITHER A STANDARD K[SUB]IC OR A J-INTEGRAL ANALYSIS WAS USED TO OBTAIN THE TOUGHNESS VALUE. AT ROOM TEMPERATURE, REGARDLESS OF ALLOY, ORIENTATION, OR MICROSTURECTURE, FRACTURE TOUGHNESS VALUES BETWEEN 15 AND 22 MPa m{sup 1/2} (14 AND 20 KSI IN{sup 1/2}) WERE MEASURED. THESE K[SUB]IC VALUES WERE CONSISTENT WITH MEASUREMENTS BY THE AUTHORS. INCREASING TEMPERATURE IMPROVES THE TOUGHNESS, DUE TO THE FACT THAT ONE TAKES ADVANTAGE OF THE DUCTIVE-BRITTLE TRANSITION BEHAVIOR OF MOLYBDENUM. AT 300 DEGREES C, THE FRACTURE TOUGHNESS OF RECRYSTALLIZED LCAC AND ARC-CAST TZM MOLYBDENUM WERE ALSO SIMILAR AT APPROXI MATELY 64 MPa m{sup 1/2} (58 KSI IN{sup 1/2}). IN THE STRESS-RELIEVED CONDITION, HOWEVER, THE TOUGHNESS OF ARC-CAST TZM (91 MPa m{sup 1/2}/83 KSI IN{sup 1/2}) WAS HIGHER THAN THAT OF THE LCAC MOLYBDENUM (74 MPa m{sup 1/2}/67 KSI IN{sup 1/2}).

  17. Reduction property of rare earth oxide doped molybdenum oxide

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Rare earth oxide doped molybdenum powders were prepared by the reduction of rare earth nitrites doped MoO3. The effect of rare earth oxide on the reduction behavior of molybdenum oxide had been studied by means of Temperature Programmed Reduction (TPR), thermal analysis, X-ray diffraction. Doping rare earth oxide in the powder could lower the reduction temperature of molybdenum oxide and decrease the particle size of molybdenum. The mechanism for the effects had been discussed in this paper.

  18. Tensile and impact properties of vanadium-base alloys irradiated at low temperatures in the ATR-A1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Nowicki, L.J.; Billone, M.C.; Chung, H.M.; Smith, D.L. [Argonne National Lab., IL (United States)

    1998-03-01

    Subsize tensile and Charpy specimens made from several V-(4-5)Cr-(4-5)Ti alloys were irradiated in the ATR-A1 experiment to study the effects of low-temperature irradiation on mechanical properties. These specimens were contained in lithium-bonded subcapsules and irradiated at temperatures between {approx}200 and 300 C. Peak neutron damage was {approx}4.7 dpa. Postirradiation testing of these specimens has begun. Preliminary results from a limited number of specimens indicate a significant loss of work-hardening capability and dynamic toughness due to the irradiation. These results are consistent with data from previous low-temperature neutron irradiation experiments on these alloys.

  19. Contribution to the study of recoil species produced by potassium ferrocyanide neutron irradiation; Contribution a l'etude de la reactivite des especes de recul dans le ferrocyanure de potassium irradie aux neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Meriadec Vernier de Byans, B. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-04-01

    The chemical species produced by potassium ferrocyanide neutron irradiation were separated and identified. The study of their behaviour upon thermal annealing has allowed to establish a scheme of reaction as well as a kinetic treatment of the data. Activation energies are determined in different conditions and the effects of radiation dose, oxygen and water of crystallisation upon the activation energies were studied. Preliminary E.S.R. data and its relevance to the decomposition process is also discussed. (authors) [French] On a separe et identifie les differentes especes chimiques produites par irradiation neutronique de ferrocyanure de potassium. L'etude de leur comportement au cours du recuit thermique a permis l'application de differentes cinetiques et l'etablissement d'un schema de reaction. On a deterrmine la valeur des energies d'activation dans differentes conditions de recuit ainsi que l'influence de la dose de radiations, de l'oxygene et de l'eau de cristallisation sur ces energies. Une serie de mesures de resonance paramagnetique electronique a complete cette etude. (auteurs)

  20. 法国900MWe压水堆RPV中子辐照脆化寿命管理策略研究%Strategies for life management of French 900 MWe PWR RPV due to neutron irradiation embrittlement

    Institute of Scientific and Technical Information of China (English)

    万强茂; 束国刚; 王荣山; 丁辉; 任爱; 彭啸; 张琪; 雷静

    2011-01-01

    针对法国压水堆( PWR)核电站,介绍其长寿命运行计划情况,分析反应堆压力容器(RPV)辐照监督大纲和评价方法,总结已有辐照监督数据,重点论述法国实施的RPV中子辐照脆化寿命评价技术和管理策略、研发活动等,以期对我国开展RPV中子辐照脆化寿命管理提供有益的借鉴作用.%In this paper, the situation of the long-life operation plans for French PWR nuclear power plants was presented, the RPV irradiation surveillance program and its assessment approach were analyzed, and then data obtained from 180 surveillance capsules were summarized and compared with the predictive values. It specially argued for the RPV lifetime assessment techniques and management strategies concerning the neutron irradiation embrittlement, as well as R&D activities, with an aim to provide useful reference to the RPV life management of PWR NPPs in China.

  1. Evolution of microstructure in zirconium alloys during irradiation

    CERN Document Server

    Griffiths, M; Winegar, J E

    1997-01-01

    X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 x 10 sup 2 sup 5 n.m sup - sup 2 (E > I MeV) for a range of temperatures between 330 and 580 K. Irradiation modifies the dislocation structure through nucleation and growth of dislocation loops and, for cold-worked materials in particular, climb of existing network dislocations. In general, the a-type dislocation structure tends to saturate at low fluences (10 x l0 sup 2 sup 5 n.m sup - sup 2 - in some cases). The phase structure is also modified by irradiation. The common alloying/impurity elements, Fe, Cr, and Ni, are relatively insoluble in the alpha-phase but are dispersed into the alpha-phase during irradiation irrespective of the state of the phase initial...

  2. The Nature of Surface Oxides on Corrosion-Resistant Nickel Alloy Covered by Alkaline Water

    Directory of Open Access Journals (Sweden)

    Gervasio DF

    2010-01-01

    Full Text Available Abstract A nickel alloy with high chrome and molybdenum content was found to form a highly resistive and passive oxide layer. The donor density and mobility of ions in the oxide layer has been determined as a function of the electrical potential when alkaline water layers are on the alloy surface in order to account for the relative inertness of the nickel alloy in corrosive environments.

  3. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  4. Effect of La203 nanoparticles on properties of molybdenum powder

    Institute of Scientific and Technical Information of China (English)

    王金淑; 周美玲; 左铁镛; 聂祚仁; 张久兴; 刘娟

    2001-01-01

    The properties of La2O3-doped molybdenum powder were studied. The La2O3 nanoparticles on the surface of molybdenum powder which is produced by the reduction of La(NO3)3-doped MoO2 in hydrogen decrease the intensity of feature energy loss peak of molybdenum substrate; but increase that of peak of Mo 3d. The surface of molybdenum powder exposed to the atmosphere can be reduced because the surface is mainly covered with La2O3 nanoparticles. As a result, the capability of anti-oxidation of molybdenum is improved.

  5. ANALYSIS ON THE INFLUENCING FACTORS OF MOLYBDENUM ANVIL SERVICE LIFE%影响钼合金顶头使用寿命因素浅析

    Institute of Scientific and Technical Information of China (English)

    史振琦; 黄晓玲; 易永鹏

    2014-01-01

    通过对粉末粒度、元素配比以及烧结三方面因素的讨论,浅析了其对钼合金顶头使用寿命的影响原因。结果表明:控制钼粉的粒度以及添加试剂的粉末状态,可有效提高钼合金的密度并细化晶粒;稀土元素添加总量控制在1.0%左右,可减少氧化物在钼基体中的富集,杜绝裂纹源的形成,提高钼顶头的寿命;采用氢气烧结的方式可得到质量稳定的钼顶头。%The factors that influence molybdenum anvil service life such as powder particle size,element mixture ra-tio and alloy sintering processing were analyzed in this paper. The results show that the alloy density can be effec-tively improved and the alloy grain size can be refined by controlling the particle size of molybdenum powder and the phase of addition reagent powder. The oxides eliguation in the molybdenum alloy was reduced,crack source was eliminated and service life of the molybdenum anvil can be improved when the tatal rare earth addition was con-trolled less than 1%. Molybdenum anvil quality was steady when sintering in hydrogen atmosphere.

  6. Deformation-induced structural transition in body-centred cubic molybdenum.

    Science.gov (United States)

    Wang, S J; Wang, H; Du, K; Zhang, W; Sui, M L; Mao, S X

    2014-03-07

    Molybdenum is a refractory metal that is stable in a body-centred cubic structure at all temperatures before melting. Plastic deformation via structural transitions has never been reported for pure molybdenum, while transformation coupled with plasticity is well known for many alloys and ceramics. Here we demonstrate a structural transformation accompanied by shear deformation from an original -oriented body-centred cubic structure to a -oriented face-centred cubic lattice, captured at crack tips during the straining of molybdenum inside a transmission electron microscope at room temperature. The face-centred cubic domains then revert into -oriented body-centred cubic domains, equivalent to a lattice rotation of 54.7°, and ~15.4% tensile strain is reached. The face-centred cubic structure appears to be a well-defined metastable state, as evidenced by scanning transmission electron microscopy and nanodiffraction, the Nishiyama-Wassermann and Kurdjumov-Sachs relationships between the face-centred cubic and body-centred cubic structures and molecular dynamics simulations. Our findings reveal a deformation mechanism for elemental metals under high-stress deformation conditions.

  7. Deformation-induced structural transition in body-centred cubic molybdenum

    Science.gov (United States)

    Wang, S. J.; Wang, H.; Du, K.; Zhang, W.; Sui, M. L.; Mao, S. X.

    2014-03-01

    Molybdenum is a refractory metal that is stable in a body-centred cubic structure at all temperatures before melting. Plastic deformation via structural transitions has never been reported for pure molybdenum, while transformation coupled with plasticity is well known for many alloys and ceramics. Here we demonstrate a structural transformation accompanied by shear deformation from an original -oriented body-centred cubic structure to a -oriented face-centred cubic lattice, captured at crack tips during the straining of molybdenum inside a transmission electron microscope at room temperature. The face-centred cubic domains then revert into -oriented body-centred cubic domains, equivalent to a lattice rotation of 54.7°, and ~15.4% tensile strain is reached. The face-centred cubic structure appears to be a well-defined metastable state, as evidenced by scanning transmission electron microscopy and nanodiffraction, the Nishiyama-Wassermann and Kurdjumov-Sachs relationships between the face-centred cubic and body-centred cubic structures and molecular dynamics simulations. Our findings reveal a deformation mechanism for elemental metals under high-stress deformation conditions.

  8. Enhanced photochromism in nanostructured molybdenum trioxide films

    Science.gov (United States)

    Beydaghyan, Gisia; Doiron, Serge; Haché, Alain; Ashrit, P. V.

    2009-08-01

    We present evidence of enhancement of photochromism in nanostructured thin films of molybdenum oxide fabricated by glancing angle deposition. The strong correlation of coloration response with the internal surface area of the films provides evidence of the importance of nanostructuring on the photochromic effect and the vital role played by the availability of water in the photochromic mechanism.

  9. Exploring atomic defects in molybdenum disulphide monolayers

    KAUST Repository

    Hong, Jinhua

    2015-02-19

    Defects usually play an important role in tailoring various properties of two-dimensional materials. Defects in two-dimensional monolayer molybdenum disulphide may be responsible for large variation of electric and optical properties. Here we present a comprehensive joint experiment-theory investigation of point defects in monolayer molybdenum disulphide prepared by mechanical exfoliation, physical and chemical vapour deposition. Defect species are systematically identified and their concentrations determined by aberration-corrected scanning transmission electron microscopy, and also studied by ab-initio calculation. Defect density up to 3.5 × 10 13 cm \\'2 is found and the dominant category of defects changes from sulphur vacancy in mechanical exfoliation and chemical vapour deposition samples to molybdenum antisite in physical vapour deposition samples. Influence of defects on electronic structure and charge-carrier mobility are predicted by calculation and observed by electric transport measurement. In light of these results, the growth of ultra-high-quality monolayer molybdenum disulphide appears a primary task for the community pursuing high-performance electronic devices.

  10. Defects in hyperpure Fe-based alloys created by 3 MeV e{sup -}-irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Li, X.H.; Moser, P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M.; Van Duysen, C. [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    Information about vacancy defects created in RPV (Reactor Pressure Vessels) steels after neutron irradiations are obtained via a simulation: the RPV steels are simulated by a series of high purity Fe-based alloys; the neutron irradiation is simulated by a 3 MeV electron irradiation; vacancy defects characteristics are obtained by positron lifetime techniques. Irradiations are made at 150 or 288 deg C, with a dose of 4*10{sup 19} e-/cm{sup 2}, and followed by isochronal annealing in the range 20-500 deg C. The observed vacancy defects are single trapped vacancies and small vacancy clusters, the size of which being lower than 10 empty atomic volumes (vacancy clusters containing more than 50 empty atomic volumes were never found). A large recovery step is observed between 200 and 400 deg C, after 150 deg C irradiation and attributed to vacancy-impurity detrapping, and also, vacancy cluster evaporation. The influence of C, Cu and Mo are presented. These results are in agreement with a model supposing, in pure Fe, single vacancy migration at -50 deg C and vacancy-impurity detrapping at 200 deg C. (authors). 4 figs., 15 refs.

  11. The role of interstitial binding in radiation induced segregation in W-Re alloys

    Science.gov (United States)

    Gharaee, Leili; Marian, Jaime; Erhart, Paul

    2016-07-01

    Due to their high strength and advantageous high-temperature properties, tungsten-based alloys are being considered as plasma-facing candidate materials in fusion devices. Under neutron irradiation, rhenium, which is produced by nuclear transmutation, has been found to precipitate in elongated precipitates forming thermodynamic intermetallic phases at concentrations well below the solubility limit. Recent measurements have shown that Re precipitation can lead to substantial hardening, which may have a detrimental effect on the fracture toughness of W alloys. This puzzle of sub-solubility precipitation points to the role played by irradiation induced defects, specifically mixed solute-W interstitials. Here, using first-principles calculations based on density functional theory, we study the energetics of mixed interstitial defects in W-Re, W-V, and W-Ti alloys, as well as the heat of mixing for each substitutional solute. We find that mixed interstitials in all systems are strongly attracted to each other with binding energies of -2.4 to -3.2 eV and form interstitial pairs that are aligned along parallel first-neighbor strings. Low barriers for defect translation and rotation enable defect agglomeration and alignment even at moderate temperatures. We propose that these elongated agglomerates of mixed-interstitials may act as precursors for the formation of needle-shaped intermetallic precipitates. This interstitial-based mechanism is not limited to radiation induced segregation and precipitation in W-Re alloys but is also applicable to other body-centered cubic alloys.

  12. Quenching and partitioning response of carbon-manganese-silicon sheet steels containing nickel, molybdenum, aluminum and copper additions

    Science.gov (United States)

    Kahkonen, Joonas

    In order to produce passenger vehicles with improved fuel economy and increased passenger safety, car manufacturers are in need of steels with enhanced strength levels and good formability. Recently, promising combinations of strength and ductility have been reported for several, so-called third generation advanced high-strength steels (AHSS) and quenching and partitioning (Q&P) steels are increasingly being recognized as a promising third generation AHSS candidate. Early Q research used conventional TRIP steel chemistries and richer alloying strategies have been explored in more recent studies. However, systematic investigations of the effects of alloying elements on tensile properties and retained austenite fractions of Q&P steels are sparse. The objective of the present research was to investigate the alloying effects of carbon, manganese, molybdenum, aluminum, copper and nickel on tensile properties and microstructural evolution of Q&P heat treated sheet steels. Seven alloys were investigated with 0.3C-1.5Mn-1.5Si (wt pct) and 0.4C-1.5Mn-1.5Si alloys used to study carbon effects, a 0.3C-5Mn-1.6Si alloy to study manganese effects, 0.3C-3Mn-1.5Si-0.25Mo and 0.3C-3Mn-1.5Si-0.25Mo-0.85Al alloys to study molybdenum and aluminum effects and 0.2C-1.5Mn-1.3Si-1.5Cu and 0.2C-1.5Mn-1.3Si-1.5Cu-1.5Ni alloys to study copper and nickel effects. Increasing alloy carbon content was observed to mainly increase the ultimate tensile strength (UTS) up to 1865 MPa without significantly affecting total elongation (TE) levels. Increasing alloy carbon content also increased the resulting retained austenite (RA) fractions up to 22 vol pct. Measured maximum RA fractions were significantly lower than the predicted maximum RA levels in the 0.3C-1.5Mn-1.5Si and 0.4C-1.5Mn-1.5Si alloys, likely resulting from transition carbide formation. Increasing alloy manganese content increased UTS, TE and RA levels, and decreased yield strength (YS) and austenite carbon content (Cgamma) levels

  13. Effect of molybdenum on the microstructure and wear resistance of Fe-based hardfacing coatings

    Energy Technology Data Exchange (ETDEWEB)

    Wang, X.H. [School of Materials Science and Engineering, Shandong University, Jinan 250061 (China)], E-mail: xinhongwang@sdu.edu.cn; Han, F. [Department of Mechanical and Electrical Engineering, College of Weifang, Weifang 261021 (China); Liu, X.M.; Qu, S.Y.; Zou, Z.D. [School of Materials Science and Engineering, Shandong University, Jinan 250061 (China)

    2008-08-20

    Fe-based hardfacing alloys containing molybdenum compound have been deposited on AISI 1020 steel substrates by shield manual arc welding (SMAW) process. The effect of Mo on the microstructure and wear resistance of the Fe-based hardfacing alloys were investigated by means of X-ray diffraction, optical microscopy, scanning electron microscopy (SEM) and electron probe microanalysis, as well as wear test. The results indicated that cuboidal and rod-type complex carbides were synthesized in the lath martensite matrix. The fraction of carbides in hardfacing layer increased with an increasing of Mo content. The hardfacing layer with good cracking resistance and wear resistance could be obtained when the amounts of Fe-Mo was controlled within a range of 3-4 wt.%. The improvement of hardness and wear resistance of the hardfacing layers attributed to the formation of Mo{sub 2}C carbide and the solution strengthening of Mo.

  14. Optimization design for the epithermal neutron duct of in-hospital neutron irradiator mark 1 reactor%医院中子照射器Ⅰ型堆超热中子束流孔道的优化设计

    Institute of Scientific and Technical Information of China (English)

    江新标; 朱养妮; 赵柱民; 陈立新; 周永茂

    2012-01-01

    采用蒙特卡罗程序(Monte Carlo neutron and photo transport code,MCNP)对医院中子照射器Ⅰ型堆(IHNI-1)超热中子束流孔道的慢化层、反射层进行了优化设计.首先对FLUENTAL、Al等材料组成的6种慢化体方案进行了分析比较,给出了孔道出口处超热中子通量密度较大的两种设计方案;基于此两种慢化体设计方案,在保持束流孔道外框尺寸不变情况下,对慢化体周围的反射层进行了分析比较,给出了反射层的推荐方案;基于慢化体和反射层优化方案,最后给出了超热中子束流孔道出口处束流参数的空间分布.%Optimization design for the moderation layer and reflection layer of the epithermal neutron duct at in-hospital neutron irradiator mark 1(IHNI-1) reactor is carried out by using MCNP in this paper. Firstly, six moderator schemes combined with FLUENTAL are compared with Al materials, and two moderation optimization schemes which can obtain intensive epithermal neutron flux density at exit of this duct are chosen. Secondly, based on these two moderation schemes, the optimization design for reflectors around the moderator is introduced, and the recommended reflector schemes are given. Finally, based on the moderation layer and reflection layer optimization schemes, the neutron and gamma space distribution of the epithermal neutron beam at exit of this duct are detailed calculated.

  15. Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

    Energy Technology Data Exchange (ETDEWEB)

    Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R. [Bhabha Atomic Research Centre, Mumbai (India)

    2017-04-15

    The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples.

  16. Heavy ion irradiation induced dislocation loops in AREVA's M5 Registered-Sign alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hengstler-Eger, R.M., E-mail: Rosmarie.Hengstler-Eger@areva.com [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Baldo, P. [Argonne National Laboratory, Materials Science Division, 9700 South Cass Avenue, 60439 Argonne IL (United States); Beck, L. [Maier-Leibnitz-Laboratorium (MLL), Am Coulombwall 6, 85748 Garching (Germany); Dorner, J.; Ertl, K. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Hoffmann, P.B. [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Hugenschmidt, C. [Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II), Lichtenbergstr. 1, 85747 Garching (Germany); Kirk, M.A. [Argonne National Laboratory, Materials Science Division, 9700 South Cass Avenue, 60439 Argonne IL (United States); Petry, W.; Pikart, P. [Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II), Lichtenbergstr. 1, 85747 Garching (Germany); Rempel, A. [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)

    2012-04-15

    Pressurized water reactor (PWR) Zr-based alloy structural materials show creep and growth under neutron irradiation as a consequence of the irradiation induced microstructural changes in the alloy. A better scientific understanding of these microstructural processes can improve simulation programs for structural component deformation and simplify the development of advanced deformation resistant alloys. As in-pile irradiation leads to high material activation and requires long irradiation times, the objective of this work was to study whether ion irradiation is an applicable method to simulate typical PWR neutron damage in Zr-based alloys, with AREVA's M5 Registered-Sign alloy as reference material. The irradiated specimens were studied by electron backscatter diffraction (EBSD), positron Doppler broadening spectroscopy (DBS) and in situ transmission electron microscopy (TEM) at different dose levels and temperatures. The irradiation induced microstructure consisted of - and -type dislocation loops with their characteristics corresponding to typical neutron damage in Zr-based alloys; it can thus be concluded that heavy ion irradiation under the chosen conditions is an excellent method to simulate PWR neutron damage.

  17. Mechanical alloying of Al-3 at. % Mo powders

    Energy Technology Data Exchange (ETDEWEB)

    Zdujic, M. (Srpska Akademija Nauka i Umetnosti, Belgrade (Yugoslavia). Dept. of Technical Science); Kobayashi, K.F. (Osaka Univ., Suita (Japan). Dept. of Welding and Production Engineering); Shingu, P.H. (Kyoto Univ. (Japan). Dept. of Metal Science and Technology)

    1990-05-01

    Mechanical alloying of elemental powders of aluminum and molybdenum (Al-3 at.% Mo) has been carried out in a conventional horizontal ball mill up to 1000 h of milling time. Mechanically alloyed powders were investigated by scanning electron microscopy, X-ray diffraction analysis and differential scanning calorimetry. After prolonged milling time molybdenum was finely dispersed in aluminum matrix. The dispersoid sizes were less than about 100 nm, with average size considerably smaller. By the heat treatment of the mechanically alloyed powders, the intermetallic compound Al{sub 12}Mo was formed. The reaction temperature for the formation of Al{sub 12}Mo decreased with increasing milling time. The Johnson-Mehl-Avrami exponent of n=2.8{plus minus}0.3 for the formation of Al{sub 12}Mo was obtained with the apparent activation energy of 165{plus minus}12 kJ/mol (1.7{plus minus}0.1 eV). (orig.).

  18. Trace Carbon in Biomedical Beta-Titanium Alloys: Recent Progress

    Science.gov (United States)

    Zhao, D.; Ebel, T.; Yan, M.; Qian, M.

    2015-08-01

    Owing to their relatively low Young's modulus, high strength, good resistance to corrosion, and excellent biocompatibility, β-titanium (Ti) alloys have shown great potential for biomedical applications. In β-Ti alloys, carbon can exist in the form of titanium carbide (TiC x ) as well as interstitial atoms. The Ti-C binary phase diagram predicts a carbon solubility value of 0.08 wt.% in β-Ti, which has been used as the carbon limit for a variety of β-Ti alloys. However, noticeable grain boundary TiC x particles have been observed in β-Ti alloys containing impurity levels of carbon well below the predicted 0.08 wt.%. This review focuses its attention on trace carbon (≤0.08 wt.%) in biomedical β-Ti alloys containing niobium (Nb) and molybdenum (Mo), and it discusses the nature and precipitation mechanism of the TiC x particles in these alloys.

  19. Molybdenum-catalyzed deoxydehydration of vicinal diols

    DEFF Research Database (Denmark)

    Dethlefsen, Johannes Rytter; Lupp, Daniel; Oh, Byung Chang

    2014-01-01

    The commercially available (NH4)6Mo7O24 and other molybdenum compounds are shown to be viable substitutes for the typically employed rhenium compounds in the catalytic deoxydehydration of aliphatic diols into the corresponding alkenes. The transformation, which represents a model system for the v......The commercially available (NH4)6Mo7O24 and other molybdenum compounds are shown to be viable substitutes for the typically employed rhenium compounds in the catalytic deoxydehydration of aliphatic diols into the corresponding alkenes. The transformation, which represents a model system...... for the various hydroxyl groups found in biomass-derived carbohydrates, can be conducted in an inert solvent (dodecane), under solvent-free conditions, and in a solvent capable of dissolving biomass-derived polyols (1,5-pentanediol). The reaction is driven by the simultaneous oxidative deformylation of the diol...

  20. Deformation localization and cyclic strength in polycrystalline molybdenum

    Energy Technology Data Exchange (ETDEWEB)

    Sidorov, O.T.; Rakshin, A.F.; Fenyuk, M.I.

    1983-06-01

    Conditions of deformation localization and its interrelation with cyclic strength in polycrystalline molybdenum were investigated. A fatigue failure of polycrystalline molybdenum after rolling and in an embrittled state reached by recrystallization annealing under cyclic bending at room temperature takes place under nonuniform distribution of microplastic strain resulting in a temperature rise in separate sections of more than 314 K. More intensive structural changes take place in molybdenum after rolling than in recrystallized state.