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Sample records for neutron-irradiated molybdenum alloys

  1. 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

    The influence of neutron irradiation on microstructural evolution and mechanical properties of mono- and polycrystalline molybdenum and its alloys has been investigated. Tensile specimens and 3 mm diameter discs of monocrystals of pure molybdenum and Mo-5%Re were irradiated with fission neutrons...... microscopy are presented and discussed in terms of intracascade clustering of self-interstitial atoms and the role of one-dimensional glide of these clusters in controlling microstructural evolution and the resulting mechanical properties....

  2. PDS 1-5. Divertor heat sink materials pre- and post-neutron irradiation. Tensile and fatigue tests of brazed joints of molybdenum alloys and 316L stainless steel

    International Nuclear Information System (INIS)

    Lind, Anders.

    1994-01-01

    Tensile specimens from brazed joints of molybdenum alloys (TZM or Mo-5%Re) and Type 316L austenitic stainless steel tubes have been tested at ambient temperature and 127 degrees C before and after neutron irradiation at about 40 degrees C to approximately 0.2 dpa. The unirradiated specimens showed generally ductile behaviour, but the irradiated specimens were notch sensitive and failed in a brittle manner with zero elongation; in all cases the fracture occurred in the molybdenum alloy. The brittle behaviour is consistent with previously published data and results from the increase in strength (radiation hardening) and the associated increase in the ductile-brittle transition temperature (radiation embrittlement) induced in the body-centered-cubic (BCC) molybdenum alloys by irradiation to relatively low displacement doses. The same type of irradiated specimens were also used in fatigue tests. However, the results from the fatigue tests are too limited and complementary studies are needed. During exposure to water locally up to 25% of the wall thickness of the Mo-alloys has corroded away. These observations cast serious doubts on the viability of the molybdenum alloys for divertor applications in fusion systems. 8 refs, 29 figs

  3. Positron lifetime study of neutron-irradiated molybdenum

    International Nuclear Information System (INIS)

    Hinode, Kenji; Tanigawa, Shoichiro; Kumakura, Hiroaki; Doyama, Masao; Shiraishi, Kensuke.

    1978-01-01

    Annealing behavior of fast-neutron-irradiated molybdenum was studied by means of positron lifetime technique. It was found that Stage III annealing can be mainly identified as the vacancy migration process from the detailed analyses of data. The void growth after successive high temperature annealings was clearly detected through the changes of positron lifetime parameters. An attempt to analyse the size distribution of voids from positron lifetime spectra was presented, and discussions on the evaluation of void concentration from positron data are also given. (author)

  4. Effect of neutron irradiation on vanadium alloys

    International Nuclear Information System (INIS)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

  5. Effect of neutron irradiation on vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  6. EDX microanalysis of neutron-irradiated alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1981-09-01

    Energy-dispersive X-ray (EDX) spectrometry of 50 nm thick specimens in the scanning transmission electron microscope provides quantitative elemental analyses of selected regions as small as 20 nm in diameter. To analyze highly radioactive neutron-irradiated alloys it is necessary to reduce the high counting deadtimes caused by energetic γ-Compton scattering in the Si(Li) detector, and to account for spurious background contributions from γ-rays and characteristic x-ray emissions. Several simple methods for overcoming effects of specimen radioactivity are described, including use of a tungsten collimator to attenuate γ and x-rays coming from the thick edges of self-supporting disk specimens. These methods allow analyses of Fe-Cr-Ni based alloys with γ-activities up to 1000 μC/sub i/. Techniques used to maintain high spatial resolution and accuracy in quantitatve analysis are also described, and their use is illustrated

  7. Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

    International Nuclear Information System (INIS)

    Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

    2011-01-01

    The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

  8. Swelling in neutron-irradiated titanium alloys

    International Nuclear Information System (INIS)

    Peterson, D.T.

    1982-04-01

    Immersion density measurements have been performed on a series of titanium alloys irradiated in EBR-II to a fluence of 5 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 and 550 0 C. The materials irradiated were the near-alpha alloys Ti-6242S and Ti-5621S, the alpha-beta alloy Ti-64, and the beta alloy Ti-38644. Swelling was observed in all alloys with the greater swelling being observed at 550 0 C. Microstructural examination revealed the presence of voids in all alloys. Ti-38644 was found to be the most radiation resistant. Ti-6242S and Ti-5621S also displayed good radiation resistance, whereas considerable swelling and precipitation were observed in Ti-64 at 550 0 C

  9. Swelling in neutron irradiated nickel-base alloys

    International Nuclear Information System (INIS)

    Brager, H.R.; Bell, W.L.

    1972-01-01

    Inconel 625, Incoloy 800 and Hastelloy X were neutron irradiated at 500 to 700 0 C. It was found that of the three alloys investigated, Inconel 625 offers the greatest swelling resistance. The superior swelling resistance of Inconel 625 relative to that of Hastelloy-X is probably related to differences in the concentrations of the minor rather than major alloy constituents, and can involve (a) enhanced recombination of defects in the Inconel 625 and (b) preferential attraction of vacancies to incoherent precipitates. (U.S.)

  10. DAMAGE IN MOLYBDENUM ASSOCIATED WITH NEUTRON IRRADIATION AND SUBSEQUENT POST-IRRADIATION ANNEALING

    Energy Technology Data Exchange (ETDEWEB)

    Mastel, B.

    1963-07-23

    Molybdemum containing carbon was studied in an attempt to establish the combined effect of impurity content and neutron irradiation on the properties and structure of specific metals. Molybdenum foils were punched into discs and heat treated in vacuum. They were then slow-cooled and irradiated. After irradiation and subsequent decay of radioactivity to a low level the foils were subjected to x-ray diffraction measurements. Cold-worked foils with less than 10 ppm carbon showed no change in microstructure due to irradiation. Molybdenum foils that were annealed prior to irradiation showed spot defects. In foils containing up to 500 ppm carbon, it was concluded that the small loops present after irradiation are due to the clustering of point defects at interstitial carbon atoms, followed by collapse to form a dislocation loop. The amount of lattice expansion after irradiation was strongly dependent on impurity content. Neutron irradiation was found to reduce the number of active slip systems. (M.C.G.)

  11. Property change of advanced tungsten alloys due to neutron irradiation

    International Nuclear Information System (INIS)

    Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki

    2013-01-01

    This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening

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

    International Nuclear Information System (INIS)

    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 x 10 27 neutrons/m 2 (17 to 114 atom displacements per atom [dpa])

  13. Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

    International Nuclear Information System (INIS)

    Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

    2007-01-01

    Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

  14. Studies on neutron irradiation effects of iron alloys and nickel-base heat resistant alloys

    International Nuclear Information System (INIS)

    Watanabe, Katsutoshi

    1987-09-01

    The present paper describes the results of neutron irradiation effects on iron alloys and nickel-base heat resistant alloys. As for the iron alloys, irradiation hardening and embrittlement were investigated using internal friction measurement, electron microscopy and tensile testings. The role of alloying elements was also investigated to understand the irradiation behavior of iron alloys. The essential factors affecting irradiation hardening and embrittlement were thus clarified. On the other hand, postirradiation tensile and creep properties were measured of Hastelloy X alloy. Irradiation behavior at elevated temperatures is discussed. (author)

  15. Fracture mechanics behaviour of neutron irradiated Alloy A-286

    International Nuclear Information System (INIS)

    Mills, W.J.; James, L.A.

    The effect of fast-neutron irradiation on the fatigue-crack propagation and fracture toughness behaviour of Alloy A-286 was characterized using fracture mechanics techniques. The fracture toughness was found to decrease continuously with increasing irradiation damage at both 24 deg. C and 427 deg. C. In the unirradiated and low fluence conditions, specimens displayed appreciable plasticity prior to fracture, and equivalent Ksub(Ic) values were determined from Jsub(Ic) fracture toughness results. At high irradiation exposure levels, specimens exhibited a brittle Ksub(Ic) fracture mode. The 427 deg. C fracture toughness fell from 129 MPa√m in the unirradiated condition to 35 MPa√m at an exposure of 16.2 dpa (total fluence of 5.2x10 22 n/cm 2 ). Room temperature fracture toughness values were consistently 40 to 60 percent higher than the 427 deg. C values. Electron fractography revealed that the reduction in fracture resistance was attributed to a fracture mechanism transition from ductile microvoid coalescence to channel fracture. Fatigue-crack propagation tests were conducted at 427 deg. C on specimens irradiated at 2.4 dpa and 16.2 dpa. Crack growth rates at the lower exposure level were comparable to those in unirradiated material, while those at the higher exposure were slightly higher than in unirradiated material. (author)

  16. Recovery characteristics of neutron-irradiated V-Ti alloys

    International Nuclear Information System (INIS)

    Leguey, T.; Pareja, R.

    2000-01-01

    The recovery characteristics of neutron-irradiated pure V and V-Ti alloys with 1.0 and 4.5 at.% Ti have been investigated by positron annihilation spectroscopy. Microvoid formation during irradiation at 320 K is produced in pure V and V-1Ti but not in V-4.5Ti. The results are consistent with a model of swelling inhibition induced by vacancy trapping by solute Ti during irradiation. The temperature dependencies of the parameter S in the range 8-300 K indicate a large dislocation bias for vacancies and solute Ti. This dislocation bias prevents the microvoid nucleation in V-4.5Ti, and the microvoid growth in V-1Ti, when vacancies become mobile during post-irradiation annealing treatments. A characteristic increase of the positron lifetime is found during recovery induced by isochronal annealing. It is attributed to a vacancy accumulation into the lattice of Ti oxides precipitated during cooling down, or at their matrix/precipitate interfaces. These precipitates could be produced by the decomposition of metastable phases of Ti oxides formed during post-irradiation annealing above 1000 K

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

    International Nuclear Information System (INIS)

    Lambri, O.A.; Zelada-Lambri, G.I.; Cuello, G.J.; Bozzano, P.B.; Garcia, J.A.

    2009-01-01

    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.

  18. 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.

  19. Void formation by annealing of neutron-irradiated plastically deformed molybdenum

    International Nuclear Information System (INIS)

    Petersen, K.; Nielsen, B.; Thrane, N.

    1976-01-01

    The positron annihilation technique has been used in order to study the influence of plastic deformation on the formation and growth of voids in neutron irradiated molybdenum single crystals treated by isochronal annealing. Samples were prepared in three ways: deformed 12-19% before irradiation, deformed 12-19% after irradiation, and - for reference purposes -non-deformed. In addition a polycrystalline sample was prepared in order to study the influence of the grain boundaries. All samples were irradiated at 60 0 C with a flux of 2.5 x 10 18 fast neutrons/cm 2 . After irradiation the samples were subjected to isochronal annealing. It was found that deformation before irradiation probably enhanced the formation of voids slightly. Deformation after irradiation strongly reduced the void formation. The presence of grain boundaries in the polycrystalline sample had a reducing influence on the growth of voids. (author)

  20. Neutron irradiation effects on magnetic properties of some Heusler alloys

    International Nuclear Information System (INIS)

    Onodera, Hideya; Shinohara, Takeshi; Yamamoto, Hisao; Watanabe, Hiroshi

    1975-01-01

    The neutron irradiation effects were studied with measurements of temperature dependence of magnetization in ordered and disordered Heusler alloys. The irradiation was carried out in JMTR with a total flux of fast neutrons of 10 20 nvt. Fully ordered Cu 2 MnIn, partially ordered Cu 2 MnAl and completely disordered Cu 2 MnSn were prepared with various temperature treatments. The magnetization-temperature curves of each specimen were measured before and after irradiation. In the irradiated Cu 2 MnIn, the disordering by the irradiation gave rise to a decrease of magnetization, and the temperature dependence of magnetization showed that the disordered region contained various regions with different degrees of disorder. For the distribution of the disordered region, the calculation based on the theory of temperature spike by Seitz and Koekler gave a feasible result that a disordered region comprised a central core with a radius of 5.4 A which was completely disordered and a periphery of 3.3 A thickness which was partially disordered. From the magnetization-temperature curves of Cu 2 MnAl, it was considered that the disordered regions induced by the irradiation had different properties from those induced by the heat treatment. The former were the localized and comprised regions corresponding to various degrees of disorder, while the latter spread spatially in a wide range with a certain degree of disorder. The ordering by enhanced diffusion occurred simultaneously to an extent comparable to the disordering, and so it played an important role in the magnetization in the partially disordered Cu 2 MnAl. In the disordered Cu 2 MnSn, however, the ordering effect was very small. It is supposed to be difficult for the A2 structure to transform into the L2 1 structure by the enhanced diffusion. (auth.)

  1. Neutron irradiation effects of iron alloys and ceramics

    International Nuclear Information System (INIS)

    Kuramoto, Eiichi; Takenaka, Minoru; Hasegawa, Masayuki.

    1991-01-01

    Positron annihilation angular correlation measurements have been performed for the neutron irradiated various metals and ceramics in order to obtain the information of the microvoids and positronium formation in them. Positronium (Ps) formation was observed in Nb containing a small amount of oxygen and Fe-15%Cr-16%Ni-0.006%B 10 . In practical steels such as JPCA and JFMS no Ps formation was observed. High temperature deformation might induce microvoids into metals, but the positron annihilation angular correlation measurements could not confirm this. In non-metallic materials neutron irradiated no Ps formation has so far been observed. (author)

  2. Neutron irradiation test of copper alloy/stainless steel joint materials

    International Nuclear Information System (INIS)

    Yamada, Hirokazu; Kawamura, Hiroshi

    2006-01-01

    As a study about the joint technology of copper alloy and stainless steel for utilization as cooling piping in International Thermonuclear Experimental Reactor (ITER), Al 2 O 3 -dispersed strengthened copper or CuCrZr was jointed to stainless steel by three kinds of joint methods (casting joint, brazing joint and friction welding method) for the evaluation of the neutron irradiation effect on joints. A neutron irradiation test was performed to three types of joints and each copper alloy. The average value of fast neutron fluence in this irradiation test was about 2 x 10 24 n/m 2 (E>1 MeV), and the irradiation temperature was about 130degC. As post-irradiation examinations, tensile tests, hardness tests and observation of fracture surface after the tensile tests were performed. All type joints changed to be brittle by the neutron irradiation effect like each copper alloy material, and no particular neutron irradiation effect due to the effect of joint process was observed. On the casting and friction welding, hardness of copper alloy near the joint boundary changed to be lower than that of each copper alloy by the effect of joint procedure. However, tensile strength of joints was almost the same as that of each copper alloy before/after neutron irradiation. On the other hand, tensile strength of joints by brazing changed to be much lower than CuAl-25 base material by the effect of joint process before/after neutron irradiation. Results in this study showed that the friction welding method and the casting would be able to apply to the joint method of piping in ITER. This report is based on the final report of the ITER Engineering Design Activities (EDA). (author)

  3. Neutron irradiation damage of a stress relieved TZM alloy

    International Nuclear Information System (INIS)

    Abe, K.; Masuyama, T.; Satou, M.; Hamilton, M.L.

    1992-01-01

    The objective of this work is to study defect microstructures and irradiation hardening in a stress relieved TZM alloy after irradiation in the Fast Flux Test Facility (FFTF) using the Materials Open Test Assembly (MOTA). Disk specimens of the molybdenum alloy TZM that had been stress relieved at 1199 K (929 C) for 0.9 ks (15 min.) were irradiated in the FFTF/MOTA 1F at 679, 793 and 873 K (406, 520, and 600 C) to a fast fluence of ∼9.6 x 10 22 n/cm 2 . Microstructures were observed in a transmission electron microscope (TEM). Dislocation structures consisted of isolated loops, aggregated loops (rafts) and elongated dislocations. The size of the loops increased with the irradiation temperature. Void swelling was about 1 and 2% at 793 and 873 K (520 and 600 C), respectively. A void lattice was developed in the body centered cubic (bcc) structure with a spacing of 26 - 28 nm. The fine grain size (0.5 - 2 μm) was retained following high temperature irradiation, indicating that the stress relief heat treatment may extend the material's resistance to radiation damage up to high fluence levels. Microhardness measurements indicated that irradiation hardening increased with irradiation temperature. The relationship between the microstructure and the observed hardening was determined

  4. 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.

  5. Effect of neutron irradiation on the cellular stage of Ni-Be alloy decomposition

    International Nuclear Information System (INIS)

    Larikov, L.N.; Borimskaya, S.T.

    1981-01-01

    Effects of neutron irradiation on the cellular stage of decomposition are investigated in deformed supersaturated solid solution Ni-1.92%Be by the X-ray structural and metallographic analyses. Radiation-initiated stimulation of the recovery properties in the deformed alloy and a lower rate of the cellular decomposition in irradiated samples are discovered [ru

  6. Neutronic irradiation effect in FeNi alloys, observed by magnetic measurements

    International Nuclear Information System (INIS)

    Sciani, V.; Lucki, G.

    1986-01-01

    In this work some aspects of radiation damage are analysed through the influence of neutron irradiation on magnetic properties of FeNi alloys. The main points emphasized are: radiation enhanced diffusion, determination of the activation energy for diffusion process and vacancies supersaturation, which is an important parameter from technological point of view and a necessary condition for the void formation. (Author) [pt

  7. Alloys of nickel-iron and nickel-silicon do not swell under fast neutron irradiation

    International Nuclear Information System (INIS)

    Silvestre, G.; Silvent, A.; Regnard, C.; Sainfort, G.

    1975-01-01

    This research is concerned with the effect of fast-neutron irradiation on the swelling of nickel and nickel alloys. Ni-Fe (0-60at%Fe) and Ni-Si (0-8at%Si) were studied, and the fluences were in the range 10 20 -4.3x10 22 n/cm 2 . In dilute alloys, the added elements are dissolved and reduce swelling, silicon being particularly effective. In more concentrated alloys, irradiation of Ni-Fe and Ni-Si alloys brings about the formation of plate-shaped precipitates of Ni 3 X and these alloys do not swell. (Auth.)

  8. 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)

  9. Neutron irradiation effect on thermomechanical properties of shape memory alloys

    International Nuclear Information System (INIS)

    Abramov, V.Ya.; Ionajtis, R.R.; Kotov, V.V.; Loguntsev, E.N.; Ushakov, V.P.

    1996-01-01

    Alloys of Ti-Ni, Ti-Ni-Pd, Fe-Mn-Si, Mn-Cu-Cr, Mn-Cu, Cu-Al-Mn, Cu-Al-Ni systems are investigated after irradiation in IVV-2M reactor at various temperatures with neutron fluence of 10 19 - 10 20 cm -2 . The degradation of shape memory effect in titanium nickelide base alloys is revealed after irradiation. Mn-Cu and Mn-Cu-Cr alloys show the best results. Trends in shape memory alloy behaviour depending on irradiation temperature are found. A consideration is given to the possibility of using these alloys for components of power reactor control and protection systems [ru

  10. Short-range order of amorphous FeNiB alloy after neutron irradiation

    International Nuclear Information System (INIS)

    Miglierini, M.; Sitek, J.; Baluch, S.; Cirak, J.; Lipka, J.

    1990-01-01

    Transmission Moessbauer spectroscopy was used to study irradiation-induced changes in the short-range order of an amorphous Fe 80-x Ni x B 20 alloy. Neutron irradiation led to an increase of the width of a hyperfine field distribution implying atomic rearrangement towards disordering. Changes in a mean value of a HFD and Moessbauer line areas can be associated with a reorientation of spins due to radiation damage. (orig.)

  11. Concentration dependence of solute atoms on vacancy cluster formation in neutron irradiated Ni alloy

    International Nuclear Information System (INIS)

    Sato, K.; Itoh, D.; Yoshiie, T.; Xu, Q.

    2007-01-01

    Full text of publication follows: One dimensional (1-D) motion of interstitial clusters is important for the microstructural evolution in metals. The movement of interstitial clusters was often observed in neutron irradiated metals by transmission electron microscopy (TEM). Alloying elements are expected to affect the motion of interstitial clusters. Yoshiie et al. have studied the effect of alloying elements in Ni. For example, in neutron irradiated pure Ni, well-developed dislocation networks and voids were observed at 573 K at a dose of 0.026 dpa by TEM. After the addition of 2at.%Si (-5.81% volume size factor to Ni) and Sn (74.08% volume size factor), no voids were detected by TEM observation and positron lifetime measurement. Alloying elements of Si and Sn were expected to prevent the 1-D motion of the interstitial clusters. In this study, the concentration dependence of alloying elements on the 1-D motion of the interstitial clusters was investigated by positron annihilation lifetime measurements, and the microstructural evolution was discussed. Specimens irradiated were 99.99 pure Ni (Johnson Matthey) and Ni based binary alloys, which contain Si, Cu, Ge and Sn as solute atoms. The concentration of solute atoms was 0.05at.%o, 0.3at.% and 2at.%. Neutron irradiation was performed with the Kyoto University Reactor (KUR) and Japan materials testing reactor (JMTR) at Japan Atomic Energy Agency. Neutron dose was 6x10 -5 -1x10 -2 dpa at KUR, and 8x10 -3 -0.3 dpa at JMTR. Irradiation temperature was 573 K at KUR and 563 K at JMTR. After the neutron irradiation, positron annihilation lifetime measurements were performed at room temperature. Microvoids were detected in pure Ni, Ni-0.05%Si, Ni-0.05%Sn, Ni-Cu and Ni-Ge alloys. In Ni-Si and Ni-Sn alloys, the size of microvoids decreased as the concentration of solute atoms increased. This is because the frequency of 1-D motion of the interstitial clusters depends on the alloy concentration. High concentration of alloying

  12. Concentration dependence of solute atoms on vacancy cluster formation in neutron irradiated Ni alloy

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K.; Itoh, D.; Yoshiie, T.; Xu, Q. [Kyoto Univ., Research Reactor Institute, Osaka (Japan)

    2007-07-01

    Full text of publication follows: One dimensional (1-D) motion of interstitial clusters is important for the microstructural evolution in metals. The movement of interstitial clusters was often observed in neutron irradiated metals by transmission electron microscopy (TEM). Alloying elements are expected to affect the motion of interstitial clusters. Yoshiie et al. have studied the effect of alloying elements in Ni. For example, in neutron irradiated pure Ni, well-developed dislocation networks and voids were observed at 573 K at a dose of 0.026 dpa by TEM. After the addition of 2at.%Si (-5.81% volume size factor to Ni) and Sn (74.08% volume size factor), no voids were detected by TEM observation and positron lifetime measurement. Alloying elements of Si and Sn were expected to prevent the 1-D motion of the interstitial clusters. In this study, the concentration dependence of alloying elements on the 1-D motion of the interstitial clusters was investigated by positron annihilation lifetime measurements, and the microstructural evolution was discussed. Specimens irradiated were 99.99 pure Ni (Johnson Matthey) and Ni based binary alloys, which contain Si, Cu, Ge and Sn as solute atoms. The concentration of solute atoms was 0.05at.%o, 0.3at.% and 2at.%. Neutron irradiation was performed with the Kyoto University Reactor (KUR) and Japan materials testing reactor (JMTR) at Japan Atomic Energy Agency. Neutron dose was 6x10{sup -5}-1x10{sup -2} dpa at KUR, and 8x10{sup -3} -0.3 dpa at JMTR. Irradiation temperature was 573 K at KUR and 563 K at JMTR. After the neutron irradiation, positron annihilation lifetime measurements were performed at room temperature. Microvoids were detected in pure Ni, Ni-0.05%Si, Ni-0.05%Sn, Ni-Cu and Ni-Ge alloys. In Ni-Si and Ni-Sn alloys, the size of microvoids decreased as the concentration of solute atoms increased. This is because the frequency of 1-D motion of the interstitial clusters depends on the alloy concentration. High

  13. Extra spots in the electron diffraction patterns of neutron irradiated zirconium and its alloys

    International Nuclear Information System (INIS)

    Madden, P.K.

    1977-01-01

    Specimens of neutron irradiated zirconium and its alloys were examined in the transmission electron microscope. Groups of extra spots, often exhibiting four-fold symmetry, were observed in thin foil electron diffraction patterns of these specimens. The 'extra-spot' structure, like the expected black-dot/small scale dislocation loop neutron irradiated damage, is approximately 100 A in size. Its nature is uncertain. It may be related to irradiation damage or to some artefact introduced during specimen preparation. If it is the latter, then published irradiation damage defect size distributions and determined irradiation growth strains of other investigators, may require modification. The present inconclusive results indicate that extra-spot structure is likely to consist of oxide particles, but may correspond to hydride precipitation or decoration effects, or even, to electron beam effects. (author)

  14. Analytical electron microscopy of neutron-irradiated reactor alloys

    International Nuclear Information System (INIS)

    Thomas, L.E.

    1982-01-01

    Exposure to the high neutron fluxes and temperatures from 400 to 650 0 C in the core region of a fast breeder reactor profoundly alters the microstructure and properties of structural steels and superalloys. The development of irradiation-induced voids, dislocations and precipitates, as well as segregation of alloying elements on a microscopic scale has been related to macroscopic swelling, creep, hardening and embrittlement which occur during prolonged exposures in reactor. Microanalytical studies using TEM/STEM methods, primarily energy dispersive x-ray (EDX) microanalysis, have greatly aided understanding of alloy behavior under irradiation. The main uses of analytical electron microscopy in studying irradiated alloys have been the identification of irradiation-induced precipitates and determination of the changes in local composition due to irradiation-induced solute segregation

  15. α′ precipitation in neutron-irradiated Fe–Cr alloys

    International Nuclear Information System (INIS)

    Bachhav, Mukesh; Robert Odette, G.; Marquis, Emmanuelle A.

    2014-01-01

    Graphical abstract: -- A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed α′ precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of α′ precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1987-09-01

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

  18. Relationship of microstructure and tensile properties for neutron-irradiated vanadium alloys

    International Nuclear Information System (INIS)

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

    1990-01-01

    The microstructures in V-15Cr-5Ti, V-10Cr-5RTi, V-3Ti-1Si, V-15Ti-7.5Cr, and V-20Ti alloys were examined by transmission electron microscopy after neutron irradiation at 600 degree C to 21--84 atom displacements per atom in the Materials Open Test Assembly of the Fast Flux Test Facility. The microstructures in these irradiated alloys were analyzed to determine the radiation-produced dislocation density, precipitate number density and size, and void number density and size. The results of these analyses were used to compute increases in yield stress and swelling of the irradiated alloys. The computed increase in yield stress was compared with the increase in yield stress determined from tensile tests on these irradiated alloys. This comparison made it possible to evaluate the influence of alloy composition on the evolution of radiation-damaged microstructures and the resulting tensile properties. 11 refs

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

    Science.gov (United States)

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

    1988-07-01

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

  20. Internal friction in Al alloys after neutron irradiation at low temperature

    International Nuclear Information System (INIS)

    Takamura, S.; Kobiyama, M.

    1985-01-01

    Internal friction and elastic modulus of dilute Al alloys have been measured after fast neutron irradiation at about 5 K. The internal friction spectra in Al-Pb, Al-Si, Al-Zn, Al-Ag, Al-Sn and Al-In are very similar. This result suggests that the configuration of the interstitial-solute atom complex in these alloys is very similar. In Al-Mg, the main complexes have the configuration with nearly symmetry, but its internal friction spectrum is different from that of the above-mentioned alloys. The internal friction spectra and their annealing behavior in Al-Be, Al-Mn, Al-Fe and Al-Cu demonstrate that the configuration of their interstitial-solute atom complex seems to be different from each other and the main complex in these alloys is immobile until stage III. (author)

  1. Neutron irradiation effects on the mechanical properties of thorium and thorium--carbon alloy

    International Nuclear Information System (INIS)

    Wang, S.C.P.

    1978-04-01

    The effects of neutron exposure to 3.0 x 10 18 neutrons/cm 2 on the mechanical properties of thorium and thorium-carbon alloy are described. Tensile measurements were done at six different test temperatures from 4 0 K to 503 0 K and at two strain rates. Thorium and thorium-carbon alloy are shown to display typical radiation hardening like other face-centered cubic metals. The yield drop phenomenon of the thorium-carbon alloy is unchanged after irradiation. The variation of shear stress and effective shear stress with test temperature was fitted to Seeger's and Fleischer's equations for irradiated and unirradiated thorium and thorium-carbon alloy. Neutron irradiation apparently contributes an athermal component to the yield strength. However, some thermal component is detected in the low temperature range. Strain-rate parameter is increased and activation volume is decreased slightly for both kinds of metal after irradiation

  2. Atomic rearrangements in ordered fcc alloys during neutron irradiation

    International Nuclear Information System (INIS)

    Kirk, M.A.; Blewitt, T.H.

    1978-01-01

    Three sets of experiments performed at Argonne National Laboratory over the past few years are described. These experiments deal with atomic rearrangements in the ordered alloys Ni 3 Mn and Cu 3 Au during fast and thermal neutron bombardment. The unique magnetic properties of ordered Ni 3 Mn are utilized to investigate radiation damage production mechanisms at low temperature (5 K) where defect migration is not possible and only disordering is observed. In the case of thermal neutron bombardment, the average recoil energy is about 450 eV and significant disordering due to [110] replacement collision sequences is observed. For fast neutron bombardment where typical recoil energies are 20 keV, significant random disordering is observed but no evidence for sizable replacement sequences is found. The bombardment of ordered Cu 3 Au by fast and thermal neutrons at higher temperature (approx. 150 0 C) is studied by electrical resistance techniques. Both ordering and disordering are observed and related to the number of migrating vacancies escaping from the high energy collision cascade

  3. Internal friction study of neutron-irradiation effects on an amorphous Cu40Ti60 alloy

    International Nuclear Information System (INIS)

    Dong, Y.; Wu, G.; Xiao, K.; Li, X.; He, Y.

    1988-01-01

    Effects of neutron irradiation on the structure of an amorphous Cu 40 Ti 60 alloy have been studied by internal friction measurements. After irradiation, the position of the first internal friction peak remains almost unchanged and the shoulder position shifts towards a higher temperature by about 5 K, which indicates that the Cu 40 Ti 60 glass becomes more stable. These results are finally discussed based on the concept of changes of chemical short-range ordering and geometrical short-range ordering due to radiation damage

  4. On the recovery of neutron irradiation defects of some metals and alloys

    International Nuclear Information System (INIS)

    Mohamed, H.G.; Matta, M.K.

    2001-01-01

    This work deals with the recovery of mechanical properties of neutron irradiated material to the pre-irradiating values. Rate of migration of defects responsible for radiation hardening and those inducing radiation embrittlement is analyzed. Role of crystalline structure is also studied. Materials of FCC crystal structure used in these investigations are pure Cu, Cu-5 at. % , Al, Cu-5 at. % Si, some Ni base binary alloys and some austenitic stainless steels mainly of AISI types 304 and 316. Among materials of BCC crystalline structure Fe-6 wt % Cr alloy is used. Alloys with CPH structure used in the present investigations are Zr-l wt. % Nb and Mg - 4.8 wt % Li alloys. History of material is studied such as cold worked state and annealed condition. Character of alloying elements and their amounts were of interest in this study. The result showed that the higher the percentage radiation hardening, the slower is the migration of radiation defects. Irradiated pure metals recovered at a higher temperature than alloys. Cold work accelerated the migration of radiation defects. The amount of alloying elements had little effect on the recovery temperatures. Character of solute alloying elements (substitutional or interstitial) revealed sensitive effect on the migration of radiation defects. Rate of migration of defects causing hardening can be different from those causing embrittlement. (author)

  5. Neutron irradiation effect on the strength of jointed Ti-6Al-4V alloy

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Miya, Naoyuki

    2002-01-01

    In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8 x 10 23 n/m 2 (>0.18 MeV). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of fracture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation. (author)

  6. 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......-irradiation annealing has been carried out. The results are discussed with reference to equivalent Transmission Electron Microscopy results on the microstructure of the materials. The CuNiBe has the lowest conductivity (less than or equal to 55% of that of pure Cu), and Cu-Al(2)O(3) the highest (75-90% of pure Cu). (C...

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

    Science.gov (United States)

    Eldrup, M.; Singh, B. N.

    1998-10-01

    The electrical conductivity of three different types of copper alloys, viz. CuNiBe, CuCrZr and Cu-Al 2O 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-irradiation annealing has been carried out. The results are discussed with reference to equivalent Transmission Electron Microscopy results on the microstructure of the materials. The CuNiBe has the lowest conductivity (⩽55% of that of pure Cu), and Cu-Al 2O 3 the highest (75-90% of pure Cu).

  8. Study of neutron irradiation on F82H alloys by Mössbauer spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Huang, S.S., E-mail: h.shaosong@ht8.ecs.kyoto-u.ac.jp; Kitao, S.; Kobayashi, Y.; Yoshiie, T.; Xu, Q.; Sato, K.; Seto, M.

    2015-01-15

    The effects of neutron irradiation on F82H ferritic/martensitic stainless steel and its model alloys were studied by Mössbauer spectroscopy. The microstructural damage mechanisms of these alloys, during the void incubation period were interpreted using the short range order (SRO) parameters. Results show that within Fe–8Cr alloy, the atoms in the nearest neighbor (NN) of the Fe nuclei were inhomogeneous, prior to irradiation. A configuration trapping model of Cr supported the negative average SRO observed for the NN shells in our Fe–Cr alloys. We found that irradiation also accelerated the SRO in Fe–8Cr through a diffusion mechanism, where Cr atom repulsion was concentration dependent. Finally, comparative studies were conducted on F82H model alloys using the present Mössbauer measurements and our previously reported work on positron annihilation spectroscopy, which further established that irradiation of the alloys promoted the growth of a M{sub 23}C{sub 6} complex.

  9. Mechanical Properties and Microstructure of Neutron Irradiated Cold-worked Al-1050 and Al-6063 Alloys

    International Nuclear Information System (INIS)

    Munitz, A.; Cotler, A; Talianker, M.

    1998-01-01

    The impact of neutron irradiation on the internal microstructure, mechanical properties and fracture morphology of cold-worked Al-1050 and Al-6063 alloys was studied, using scanning and transmission electron microscopy, and tensile measurements. Specimens consisting of 50 mm long and 6 mm wide gauge sections, were punched out from Al-1050 and Al-6063 23% cold-worked tubes. They were exposed to prolonged neutron irradiation of up to 4.5x10 25 and 8x10 25 thermal neutrons/m 2 (E -3 s -1 . In general, the uniform and total elongation, the yield stress, and the ultimate tensile strength increase as functions of fluence. However, for Al-1050 a decrease in the ultimate tensile strength and yield stress was observed up to a fluence of 1x10 25 thermal neutrons/m 2 which then increase with thermal neutrons fluence. Metallographic examination and fractography for Al-6063 revealed a decrease in the local area reduction of the final fracture necking. This reduction is accompanied by 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. In contrast, for Al-1050, fracture morphology remains ductile transgranular shear rupture and the final local area reduction remains almost constant No voids could be observed in either alloy up to the maximum fluence. The dislocation density of cold-worked Al was found to decrease with the thermal neutron fluence. Prolonged annealing of unirradiated cold-worked Al-6063 at 52 degree led to similar results. Thus, it appears that, under our irradiation conditions, whereby the temperature encompassing the samples increases the exposure to this thermal field is the major factor influencing the mechanical properties and microstructure of aluminum alloys

  10. Weldability of neutron-irradiated stainless steel and nickel-base alloy

    International Nuclear Information System (INIS)

    Koyabu, Ken; Asano, Kyoichi; Takahashi, Hidenori; Sakamoto, Hiroshi; Kawano, Shohei; Nakamura, Tomomi; Hashimoto, Tsuneyuki; Koshiishi, Masato; Kato, Takahiko; Katsura, Ryoei; Nishimura, Seiji

    2000-01-01

    Degradation of of weldability caused by helium, which is generated by nuclear transmutation irradiated material, is an important issue to be addressed in planning of proactive maintenance of light water reactor core internal components. In this work, the weldability of neutron.irradiated stainless steel and nickel-base alloy, which are major constituting materials for components, was practically evaluated. The weldability was first examined by TIG welding in relation to the weld heat input and helium content using various specimens (made of SUS304 and SUS316L) sampled from reactor internal components. The specimens were neutron irradiated in a boiling water reactor to fluences from 4 x 10 24 to 1.4 x 10 26 n/ m 2 (E> l MeV ), and resulting helium generation ranged from 0.1 to 103 appm. The weld defects were characterized by dye penetrant test and cross-sectional metallography. The weldability of neutron-irradiated stainless steel was shown to be better at lower weld heat input and lower helium content. To evaluate mechanical properties of welded joints, thick plates (20 mm) specimens of SUS304 and Alloy 600 were prepared and irradiated in Japan Material Test Reactor (JMTR). The helium content of the specimens was controlled to range from 0.11 to 1.34 appm selected to determine threshold helium content to weld successfully. The welded joints had multiple passes by TIG welding process at 10 and 20 kJ/cm heat input. The welded joints of thick plate were characterized by dye penetrant test, cross-sectional metallography, tensile test, side bend test and root bend test. It was shown that irradiated stainless steel containing below 0.14 appm of helium could be welded with conventional TIG welding process (heat input below 20 kJ/cm). Nickel-base alloy, which contained as much helium as stainless steel could be welded successfully, could also be welded with conventional TIG welding process, These results served as basis to evaluate the applicability of repair welding to

  11. Effects of neutron irradiation to 63 dpa on the properties of various commercial copper alloys

    International Nuclear Information System (INIS)

    Brager, H.R.

    1985-04-01

    High purity copper and six commercial copper alloys were neutron irradiated to 47 and 63 dpa at about 450 0 C in the FFTF. Immersion density measurements showed a wide range of swelling behavior after irradiation to 63 dpa. At one extreme was CuBe in the aged and tempered (AT) condition which had densified slightly. At the other extreme was 20% CW Cu-0.1% Ag which swelled over 45%. Electrical resistivity measurements followed trends similar to previously published results for the same alloys irradiated to 16 dpa: a continued change in conductivity with fluence which appears to relate to void formation, transmutation products and coarsening of second phase precipitates. These results were compared with electrical conductivity of unirradiated alloys examined after aging for 10,000 hours. The most irradiation resistant high-conductivity copper alloys examined after 63 dpa are A125 and MZC. Cu-2.0Be, only a moderate-conductivity alloy, exhibits very consistent irradiation resistant properties

  12. Neutron irradiation effects on grain-refined W and W-alloys

    International Nuclear Information System (INIS)

    Hasegawa, A.; Fukuda, M.; Tanno, T.; Nogami, S.; Yabuuchi, K.; Tanaka, T.; Muroga, T.

    2014-10-01

    Microstructural data of neutron irradiated Tungsten (W) such as size and number density of voids and precipitates obtained by W up to 1.5dpa irradiation in the temperature range of 400-800degC were compiled quantitatively. Nucleation and growth process of these defects were clarified and a qualitative prediction of the damage structure development and hardening of W in fusion reactor environments were made taking into account the solid transmutation effects for the first time. To improve recrystallization behavior and low temperature embrittlement, grain refined-W alloys were fabricated by K- or La-doping method. Rhenium addition to the grain refining process was also examined to improve mechanical properties. Characterizations of unirradiated materials were performed. (author)

  13. Posterior magnetic effect on the pure and doped Fe-Ni alloy under neutron irradiation

    International Nuclear Information System (INIS)

    Ferreira, Iris

    1974-01-01

    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 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)

  14. Evaluation of molybdenum and its alloys

    International Nuclear Information System (INIS)

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is critically examined. Pure molybdenum's high ductile-brittle transition temperature appears to be its major disadvantage. The candidate materials examined in detail for this application include low carbon arc-cast molybdenum, TZM-molybdenum alloy, and molybdenum-rhenium alloys. Published engineering properties are collected and compared, and it appears that Mo-Re alloys with 10 to 15% rhenium offer the best combination. Hardware is presently being made from electron beam melted Mo-13Re to test this conclusion

  15. APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Blavette, D.

    1996-01-01

    Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hee Joon [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Edwards, Dan J., E-mail: dan.edwards@pnnl.gov [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Kurtz, Richard J. [Energy and Environment Directorate, Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Yamamoto, Takuya; Wu, Yuan [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Odette, G. Robert [Department of Mechanical Engineering, University of California, Santa Barbara, CA 93106 (United States); Materials Department, University of California, Santa Barbara, CA 93106 (United States)

    2017-02-15

    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 {sup 58}Ni(n{sub th},γ) {sup 59}Ni(n{sub th},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 Al{sub 2}YO{sub 3} 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).

  17. Study of short range order in alloy of glassy metals and effect of neutron irradiation on them

    International Nuclear Information System (INIS)

    Habibi, S.; Banaee, N.; Salman, M.; Gupta, A.; Principi, G.

    2000-04-01

    In this paper, we have studied a series of glassy metals with composition Fe 78-x Ni x Si 8 B 14 with x=0, 15, 25,38,53, 58. We have used Moessbauer spectroscopy to get information about short range order and local structure in these alloys. The specimens are exposed to neutron irradiation to perturb local structure and their short range order. The hyperfine parameters obtained from spectra before and after n-irradiation and are compared

  18. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B N; Edwards, D J; Horsewell, A; Toft, P

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

  19. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    International Nuclear Information System (INIS)

    Singh, B.N.; Edwards, D.J.; Horsewell, A.; Toft, P.

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10 17 n/m 2 s (E > 1 MeV, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

  20. The effects of thermal-neutron irradiation on platinum and dilute platinum-gold alloys

    International Nuclear Information System (INIS)

    Piani, C.S.B.

    1978-12-01

    The effect of varying defect concentrations on the recovery spectrum of thermal-neutron-irradiated pure platinum after isochronal anneals was investigated. The dose-independence of substages I(A), I(B) and I(C), and the dose dependence of substage I(D) and I(E), were observed to be in agreement with electron-irradiated studies. The 120 K substage in pure platinum was shown not to be due to interstitial-interstitial reactions, but could possibly be accounted for in terms of detrapping of interstitials from impurities or intrinsic immobile defects. The 360 K stage was shown to shift and was suppressed with increasing defect concentration. The possible conversion of the crowdion to a dumbbell near 160 K in Stage ll in platinum, as predicted by the two-interstitial model, was investigated by consideration of the initial slopes of the production curves between 80 K and 300 K. A minimum in these slopes was observed near 160 K and could be interpreted as due to the conversion of the highly mobile crowdion to an immobile dumbbell at this temperature. The influence of varying gold concentrations on the recovery spectrum of platinum was investigated in dilute platinum-gold alloys. The characteristics of several additional substages in Stage ll, due to the gold alloying were comparable to the results of electron-irradiation experiments. The observations made with regard to the impurity (gold) dependence of these substages could be interpreted in terms of the concentrations of the interstitials, vacancies and impurities present in the material. The interpretation of these substages was found to be consistent, if the recovery spectrum was investigated as a function of defect concentration [af

  1. EL2-related defects in neutron irradiated GaAs1/sub -x/P/sub x/ alloys

    International Nuclear Information System (INIS)

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

    1985-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Vacancies supersaturation induced by fast neutron irradiation in FeNi alloys

    International Nuclear Information System (INIS)

    Lucki, G.; Chambron, W.; Watanabe, S.; Verdone, J.

    1975-01-01

    The void formation in metals and alloys during irradiation with high-energy particles is a problem of interest in physics and of paramount importance in nuclear technology. Voids are formed as a consequence of vacancy supersaturation and result in swelling as well as in changes of mechanical, electrical and magnetic properties of materials used in power reactors. Isothermal annealings were performed between 400 and 500 0 C with and without fast-neutron (1 MeV) irradiation. Pure Fe--Ni (50--50 at. percent) was irradiated in the Melousine reactor in Grenoble, and Fe--Ni(Mo) (50--50 at. percent + 50 ppM), in the IEAR-1 reactor at the Instituto de Energia Atomica in Sao Paulo. The toroidal-shaped specimens were fabricated from Johnson Mathey zone-refined ingots, and were initially annealed at 800 0 C during 1 h in hydrogen atmosphere and then slowly cooled (4 h) inside the furnace. Magnetic After Effect measurements (MAE) permitted the evaluation of activation energies during fast-neutron irradiation (1.54 eV) and without irradiation (3.14 eV) for pure Fe--Ni and respectively (1.36 eV) and (2.32 eV) for Fe--Ni(Mo). Since the time constants of the relaxation process are inversely proportional to the vacancy concentration, a quantitative evaluation of vacancy supersaturation was made; it decreases from the value 700 at 410 0 C to the value 40 at 490 0 C for pure Fe--Ni and from 765 to 121 for Fe--Ni(Mo) in the same temperature range. 3 figures, 5 tables

  4. Microstructure and tensile properties of neutron-irradiated (FE061Ni039)3V ordered alloy

    International Nuclear Information System (INIS)

    Braski, D.N.

    1982-01-01

    Small tensile specimens of the (Fe 0 61 Ni 0 39 ) 3 V long-range-ordered alloy were irradiated in the ORR to 4 dpa at 523, 623, and 823 K and subsequently tested at the same respective temperatueres. The alloy remained ordered after irradiation at all three temperatures. Irradiation at 523 and 623 K increased the yield strength of the material by producing Frank loops in the microstructure and reduced the total elongation. The low strain hardening observed was attributed to planar slip and the absence of cross slip. Irradiation at 823 K embrittled the alloy. Premature failure was apparently initiated by helium bubbles on sigma phase boundaries which grew rapidly during the test to form microcracks. Fracture occurred after a microcrack propagated across grain boundaries that were weakened by helium and possible sulfur. New LRO alloys without sigma phase should perform better under neutron irradiation

  5. Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

    Science.gov (United States)

    Takahashi, H.; Garner, F. A.

    1992-10-01

    Additions of silicon and aluminum suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminum on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminum segregates away from these boundaries. Whereas the formation of the Ni 3Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation.

  6. Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

    International Nuclear Information System (INIS)

    Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

    2012-01-01

    The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

  7. Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

    Science.gov (United States)

    Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

    2017-11-01

    Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

  8. 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.

  9. Neutron irradiation effects on magnetic properties of iron-nickel Invar alloys

    International Nuclear Information System (INIS)

    Morita, H.; Tanji, Y.; Hiroyoshi, H.; Nakagawa, Y.

    1983-01-01

    The Curie temperature of fcc Fe-Ni containing 30-50% Ni is reaised by neutron irradiation, although no appreciable change is detected in the X-ray diffraction pattern. These results are related to a tendency to two-phase separation of the fcc phase. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, A.V., E-mail: Alexey.V.Subbotin@gmail.com [Scientific and Production Complex Atomtechnoprom, Moscow 119180 (Russian Federation); Panyukov, S.V., E-mail: panyukov@lpi.ru [PN Lebedev Physics Institute, Russian Academy of Sciences, Moscow 117924 (Russian Federation)

    2016-08-15

    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.

  11. Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

    International Nuclear Information System (INIS)

    Takahashi, H.; Garner, F.A.

    1992-01-01

    Additions of silicon and aluminium suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminium on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminium segregates away from these boundaries. Whereas the formation of the Ni 3 Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Harte, Allan, E-mail: allan.harte@manchester.ac.uk [The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Topping, M.; Frankel, P. [The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Jädernäs, D. [Studsvik Nuclear AB, SE 611 82, Nyköping (Sweden); Romero, J. [Westinghouse Electric Company, Columbia, SC (United States); Hallstadius, L. [Westinghouse Electric Sweden AB, SE 72163 Västerås (Sweden); Darby, E.C. [Rolls Royce Plc., Nuclear Materials, Derby (United Kingdom); Preuss, M. [The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom)

    2017-04-15

    Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr){sub 2} and Zr{sub 2}(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){sub 2}, predominantly from the edge region, and homogeneously in the case of Zr{sub 2}(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr{sub 2}(Fe,Ni) SPP with respect to the Zr(Fe,Cr){sub 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. - Highlights: •Protons emulate the effects of neutron irradiation in the evolution of chemistry and morphology of second phase particles. •Detailed energy-dispersive X-ray spectroscopy reveals heterogeneity in Zr-Fe-Cr SPPs both before and after irradiation. •Zr-Fe-Ni SPPs are delayed in irradiation-induced dissolution due to their better self-solubility with respect to Zr-Fe-Cr.

  13. Characterisation of neutron irradiation damage in zirconium alloys - a 'Round Robin' experiment

    International Nuclear Information System (INIS)

    Kelly, P.M.; Blake, R.G.; Jostsons, A.

    1977-01-01

    The nature of the damage structure in the neutron-irradiated zirconium specimens supplied as part of an international 'Round Robin' experiment has been studied using transmission electron microscopy. The damage structure consists entirely of a/3 dislocation loops and no evidence has been found for c component loops. Both vacancy and interstitial loops were found in specimens where inside/outside contrast analysis was possible. Quantitative measurements of loop size distributions and loop concentrations are reported. All specimens exhibited corduroy contrast to varying degress. (author)

  14. Concerning the problem of the plastic deformation mechanism changeover in neutron-irradiated metals and alloys

    International Nuclear Information System (INIS)

    Kolesnikov, A.N.; Krasnoselov, V.A.; Prokhorov, V.I.

    1982-01-01

    With a phenomenological model of plastic deformation instability as a basis, an analysis was made of the neutron irradition effects on the characteristics of strength and plasticity vs. structural parameters and radiation damage morphology. It was demonstrated that the enchanced plasticity in the initial stage of neutron irradiation has to do with the solid solution disintegration. Introduction of indestructible strengthening barriers enhances the stress-resistance of the neck-formation by 1.22 times. The ''big grain'' effect is observable during the deformation channel production only. Both the deformation twinning and deformation-induced martensite transformation raise the plastic flow stability

  15. Investigation of the magnetic aftereffect in dilute Fe-Ni alloys after low-temperature neutron irradiation

    International Nuclear Information System (INIS)

    Blythe, H.J.; Walz, F.; Kronmueller, H.

    1982-01-01

    Dilute Fe alloys containing up to 0.5 at% Ni, neutron-irradiated at 77 K, exhibit a very complicated relaxation spectrum during anneal in the temperature range 30 to 350 K. This behaviour, in which individual peaks transform from one into another, is investigated in detail. All maxima occurring in the temperature range 30 to 140 K are found to be of Debye-type with relaxation times obeying an Arrhenius equation tau = tau 0 exp (Q/kT). The major processes of these spectra are computer-analysed in order to determine their activation parameters Q and tau 0 . The complicated peak genealogy, as observed on anneal, is attributed to the presence of two configurations of reorientating Fe interstitial atoms which form small clusters together with substitutionally and interstitially dissolved Ni atoms. (author)

  16. Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

    Science.gov (United States)

    Gussev, M. N.; Cakmak, E.; Field, K. G.

    2018-06-01

    Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

  17. 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

    The phenomenon of plastic flow localization in the form of "cleared" channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels areinitiated during post-irradiation deformation...... 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...

  18. Spectrographic analysis of uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Roca, M.

    1967-01-01

    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 3 . Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO 3 . (Author) 5 refs

  19. A comparative study of the effects of thermal- and fast-neutron irradiation on some selected dilute face centered cubic alloys

    International Nuclear Information System (INIS)

    Piani, C.S.B.

    1981-12-01

    Point defect reactions in Pt and Cu and certain dilute alloys were investigated using a resistivity method following either fast-neutron or thermal-neutron irradiation at 4 K. An enhanced irradiation-induced resistivity in certain of the alloys could be attributed to actual enhanced defect production. This was related to a mechanism involving defocussing of replacement collision chains at impurities, together with possible nucleation of interstitial clusters at impurities. The close-pair recovery substages I(A), I(B) and I(C), strongly evident in thermal-neutron-irradiated materials, were suppressed by fast-neutron irradiation. This could be related to the higher energy transfers during irradiation and to significant amounts of irradiation annealing (spontaneous recombination). Fast-neutron cascades favoured interstitial clustering and reduced recovery of the interstitial migration substages I(D) and I(E). Interstitial trapping at impurities during I(D) and I(E) although evident, was less effective in fast-neutron irradiation. Higher concentrations of impurities reduced close-pair recovery as well. Stage II detrapping was related to the trapping efficiency of impurities, as well as to the effective defect concentration. Oversized impurities (Au in Pt or Cu) acted as weak traps, while undersized impurities (Cu or Ni in Pt) appeared to from deeper 'mixed-dumbbell' traps. The 120 K substage in Pt had a unique activation energy approximately 0,37 plus minus 0,03 eV, but did not seem to be due to an impurity detrapping process. It was not possible to attribute the 360 K stage in Pt with a unique activation energy in fast-neutron irradiation

  20. Application of a three-feature dispersed-barrier hardening model to neutron-irradiated Fe-Cr model alloys

    Science.gov (United States)

    Bergner, F.; Pareige, C.; Hernández-Mayoral, M.; Malerba, L.; Heintze, C.

    2014-05-01

    An attempt is made to quantify the contributions of different types of defect-solute clusters to the total irradiation-induced yield stress increase in neutron-irradiated (300 °C, 0.6 dpa), industrial-purity Fe-Cr model alloys (target Cr contents of 2.5, 5, 9 and 12 at.% Cr). Former work based on the application of transmission electron microscopy, atom probe tomography, and small-angle neutron scattering revealed the formation of dislocation loops, NiSiPCr-enriched clusters and α‧-phase particles, which act as obstacles to dislocation glide. The values of the dimensionless obstacle strength are estimated in the framework of a three-feature dispersed-barrier hardening model. Special attention is paid to the effect of measuring errors, experimental details and model details on the estimates. The three families of obstacles and the hardening model are well capable of reproducing the observed yield stress increase as a function of Cr content, suggesting that the nanostructural features identified experimentally are the main, if not the only, causes of irradiation hardening in these model alloys.

  1. Comparison of the effect of neutron irradiation on high purity vanadium and vanadium oxygen alloys

    International Nuclear Information System (INIS)

    Arsenault, R.J.; Bressers, J.

    1977-01-01

    An investigation of the effect of neutron damage on the low temperature deformation characteristics of high purity vanadium (R/sub 300K//R/sub 4.2K/ = 1100) was undertaken for two purposes. One purpose was to determine if reducing the purity interstitial content to a lower level would result in a large difference in the effective stress between irradiated and non-irradiated samples. The present data along with previously obtained data does indicate that the difference increases as the impurity interstitial content is reduced. The explanation of this observation is based on the rapid increase of the non-irradiated yield stress at 77 0 K due to small increases in the oxygen content; however, the increase of the yield stress of the irradiated samples is much less with the same increase in oxygen content. A second purpose of this investigation was to determine the size and density of observable neutron produced defects as a function of oxygen content by transmission electron microscopy, and to relate the changes in density with changes in the yield stress. It was found that the density decreases and the size increases as the oxygen content decreases. There is qualitative agreement between the increase in yield stress at 300 0 K and the observable defect density. However, the change in the yield stress at 77 0 K due to neutron irradiation cannot be related to defect density and size

  2. Spectral effects in low-dose fission and fusion neutron irradiated metals and alloys

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Atkin, S.D.; Martinez, C.

    1986-04-01

    Flat miniature tensile specimens were irradiated to neutron fluences up to 9 x 10 22 n/m 2 in the RTNS-II and in the Omega West Reactor. Specimen temperatures were the same in both environments, with runs being made at both 90 0 C and 290 0 C. The results of tensile tests on AISI 316 stainless steel, A302B pressure vessel steel and pure copper are reported here. The radiation-induced changes in yield strength as a function of neutron dose in each spectrum are compared. The data for 316 stainless steel correlate well on the basis of displacements per atom (dpa), while those for copper and A302B do not. In copper the ratio of fission dpa to 14 MeV neutron dpa for a given yield stress change is about three to one. In A302B pressure vessel steel this ratio is more than three at lower fluences, but the yield stress data for fission and 14 MeV neutron-irradiated A302B steel appears to coalesce or intersect at the higher fluences

  3. Effect of fast neutron irradiation upon the omega transformation process in zirconium--niobium alloys

    International Nuclear Information System (INIS)

    Bremer, B.W.

    1974-01-01

    The effect of fast neutrons (E greater than 1 MeV) upon the beta-omega transformation was investigated. Irradiation-produced vacancies promoted the omega transformation by migrating to the regions of high compressive stress associated with the one-dimensional omega embryos, thereby allowing rearrangement of the strain fields. This rearrangement allows the omega embryos to attain large sizes. Growth of these embryos is diffusion controlled. However, irradiation produced no increase in growth rate. It is concluded the vacancies are effectively trapped by these strain fields even at the aging temperature, 400 0 C. The omega hardening mechanism is shown to be related solely to lattice misfit, independent of irradiation, and to saturate when the magnitude of the strain causes a breakdown of the coherent interface, thereby creating one or two interfacial dislocations. Aging at 400 0 C results in alpha growth into the interomega beta phase, producing an additional hardness increase additive to that resulting from the omega phase. At higher aging temperatures 500 0 C, the omega is rapidly replaced by alpha. The alpha microstructure consists of ultra-fine grains, 1000 A, each composed of one 12 interrelated crystallographic variants. Fast neutron irradiation has no effect upon the omega metastable equilibrium phase diagram

  4. 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...... results illustrating initiation and propagation of channels during post-irradiation deformation of neutron irradiated copper and a copper alloy. The observations strongly suggest that the channels are initiated at boundaries, large inclusions and even at previously formed cleared channels. Some...... of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these results are discussed with specific reference to the origin and consequences of plastic flow localization....

  5. Effects of heat treatments and neutron irradiation on the physical and mechanical properties of copper alloys at 100 deg. C

    International Nuclear Information System (INIS)

    Singh, B.N.; Eldrup, M.; Toft, P.; Edwards, D.J.

    1998-05-01

    The final irradiation experiment in a series of screening experiments aimed at investigating the effects of bonding and bakeout thermal cycles on irradiated copper alloys is described herein. Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment. Additional specimens were reaged and given a reactor bakeout treatment at 350 deg. C for 100 h. GlidCop TM CuAl-15 (previously referred to as CuAl-25) was given a heat treatment corresponding to a bonding thermal cycle only. Specimens were neutron irradiated at 100 deg. C to a dose level of ∼0.3 dpa. Post-irradiation tensile tests at (100 deg. C), electrical resistivity measurements (at 23 deg. C), and microstructural examinations were performed. The post-irradiation tests at 100 deg. C revealed that the greatest loss of ductility occurred in the CuCrZr alloys irradiated at 100 deg. C, irrespective of the pre-irradiation heat treatment, with the uniform elongation dropping to levels of less than 1.5%. The yield and ultimate strengths for all of the individual heat treated samples increased substantially after irradiation. The same trend was observed for the CuNiBe alloys, which exhibited much higher uniform elongation and strength after irradiation than that observed in the case of CuCrZr alloys. In both alloys irradiation-induced precipitation lead to a large increase in the strength of the solution annealed specimens with a noticeable decrease in uniform elongation. The CuAl-25 alloy also experienced an increase in the overall strength of the alloy after irradiation, accompanied by approximately a 50% decrease in the uniform and total elongation. The additional bakeout treatments given to the CuCrZr and CuNiBe before irradiation served to increase the strength, but in terms of the ductility no improvement or degradation resulted from the additional thermal exposure

  6. Amorphous-to-crystalline phase transformation by neutron irradiation of the alloy Fe83B17

    International Nuclear Information System (INIS)

    Weis, J.; Gabris, F.; Cerven, I.; Sitek, J.

    1984-01-01

    The purpose of the present work is to investigate the structural changes of amorphous Fe 83 B 17 alloy after irradiation with fast neutrons ( > 1 MeV) and to compare with the crystallization behaviour of the amorphous Fe 83 B 17 alloy after annealing. The structural changes were studied by Moessbauer spectroscopy and X-ray diffraction with the usual Fourier analysis. (author)

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

    Science.gov (United States)

    Yamakawa, K.; Shimomura, Y.

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  9. Mechanical property and conductivity changes in several copper alloys after 13.5 dpa neutron irradiation

    International Nuclear Information System (INIS)

    Ames, M.; Kohse, G.; Lee, T.S.; Grant, N.J.; Harling, O.K.

    1986-01-01

    A scoping experiment in which 25 different copper materials of 17 alloy compositions were irradiated to approx.13.5 dpa approx.400 0 C in a fast reactor is described. The materials include rapidly solidified (RS) alloys, with and without oxide dispersion strengthening, as well as conventionally processed alloys. Immersion density (swelling), electrical conductivity (which can be related to thermal conductivity), and yield stress and ductility by miniature disk bend testing have been measured before and after irradiation. It was found, in general, that the Rs alloys are stable under irradiation to 13.5 dpa, showing small conductivity changes and little or no swelling. Reduction of strength and ductility, in post-irradiation tests at the irradiation temperature, are not generally observed. Some conventionally processed alloys also performed well, although irradiation softening and swelling of several percent were observed in some cases, and pure copper swelled in excess of 5%. It is concluded that a number of copper alloys should receive further study, and that higher dose irradiations will be required to establish the limits of swelling suppression in these alloys

  10. Effects of neutron irradiation on microstructure in experimental and commercial ferritic alloys

    International Nuclear Information System (INIS)

    Gelles, D.S.; Thomas, L.E.

    1983-05-01

    A series of microstructural studies have been undertaken on fast-reactor-irradiated specimens of experimental ferritic alloys and ferritic/martensitic commercial alloys covering a broad range of compositions and starting microstructures. It is found that voids do indeed form in ferritic alloys and that dislocation loops and tangles are created during irradiation at temperatures below 500 0 C. Swelling rates as high as 0.25% per 10 22 n/cm 2 have been measured. However, the major effect of irradiation is precipitation and precipitation can suppress void swelling completely and/or be responsible for degradation of mechanical properties

  11. High strength tungsten heavy alloys with molybdenum additions

    International Nuclear Information System (INIS)

    Bose, A.; Sims, D.M.; German, R.M.

    1987-01-01

    Tungsten heavy alloys are candidates for numerous applications based on the unique combination of high density, high strength, and high ductility coupled with excellent machinability. Though there has been considerable research on heavy alloys, the primary focus has been on the ductility. These alloys are well suited for ballistic uses due to their high densities and it is expected that for superior ballistic performance, a high hardness, high strength and moderate ductility alloy would be ideal. The major goal of this investigation was to obtain heavy alloys with hardness greater than HRA 72. It is evident from the phase diagrams that molybdenum, which goes into solution in tungsten, nickel and iron, could act as a potential strengthening addition. With this in view, tungsten heavy alloys with molybdenum additions were fabricated from mixed elemental powders. A baseline composition of 90W-7Ni-3Fe was chosen to its good elongation and moderate strength. The molybdenum additions were made by replacing the tungsten. Compared to the baseline properties with no molybdenum addition, the strength and hardness showed a continuous increase with molybdenum addition. The ductility of the alloy continued to decrease with increasing molybdenum content, but even with 16% wt. % molybdenum of the elongation was still around 6%. An interesting facet of these alloying additions is the grain refinement that is brought about by adding to molybdenum to the system. The grain refinement is related to the lower solubility of tunbsten in the matrix due to partial displacement by molybdenum

  12. 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.

  13. The recovery between 30K and 400K of copper and copper alloyed with gold after thermal-neutron irradiation

    International Nuclear Information System (INIS)

    Aspeling, J.C.

    1977-08-01

    The point defect reactions responsible for the recovery in resistivity between substages I(D) and I(E) in Stage I after thermal-neutron irradiation were investigated, using a new method termed the ideal isochronal method. Another substage was observed between I(D) and I(E). Whereas the peak temperatures of substages I(D) and I(E) are dependent on applied holding times in a well-known way, the main and very unusual characteristic of the new substage is that its peak temperature does not shift with a change in holding time. Using the ideal isochronal method, it was confirmed that substages I(D) and I(E) have a uniquely thermally activated energy. The new substage was attributed to a process whereby the interstitial has to overcome an energy barrier before recombining with a vacancy. In pure copper several recovery stages are observed between substage I(E) (about 52K) and Stage III (about 300K), whereas with the alloying of gold as a substitutional impurity, additional recovery stages are observed. In the alloyed copper a dose dependence contrary to that normally expected for interstitial-impurity reactions was observed. This phenomenon can, however, be explained consistently when the concentrations of interstitials, vacancies, impurities and impurity agglomerates are considered. No other free migrating point defect stage was observed between substage I(E) and Stage III. High-resolution measurements of the recovery stage at 225K, previously believed to shift in temperature with dose, showed that this stage actually consists of two stages and that the latter do not shift with dose. The two stages were explained in terms of close-pair recovery, in contrast to the previous explanation [af

  14. Microstructures of neutron-irradiated Fe-12Cr-XMn (X=15-30) ternary alloys

    International Nuclear Information System (INIS)

    Miyahara, K.; Hosoi, Y.; Garner, F.A.

    1992-01-01

    The objective of this effort is to determine the factors which control the stability of irradiated alloys proposed for reduced activation applications. The Fe-Cr-Mn alloy system is being studied as an alternative to the Fe-Cr-Ni system because of the need to reduce long-term radioactivation in fusion-power devices. In this study, four Fe-12Cr-XMn (X =15, 20, 25, 30 wt%) alloys were irradiated in the Fast Flux Test Facility to 20 dpa at 643K and 40 dpa at 679, 793, and 873K to investigate the influence of manganese content on void swelling and phase stability. The results confirm and expand the results of earlier studies that indicate that the Fe-Cr-Mn system is relatively unstable compared to that of the Fe-Cr-Ni system, with alpha and sigma phases forming as a consequence of thermal aging or high temperature irradiation

  15. The microstructure of neutron-irradiated Fe-Cr alloys: A small-angle neutron scattering study

    International Nuclear Information System (INIS)

    Heintze, C.; Bergner, F.; Ulbricht, A.; Eckerlebe, H.

    2009-01-01

    Ferritic-martensitic chromium steels are candidate materials for future applications in both Gen-IV fission and fusion technology. Experimental investigation of neutron-irradiated Fe-Cr model alloys is important in order to gain a better understanding of the interplay of chromium content and irradiation behaviour. Small-angle neutron scattering (SANS) is particularly well suited to unfold the size distribution of non-planar irradiation-induced nanoscale features such as defect-solute clusters, nanovoids and α'- particles. This size distribution represents a statistically reliable average over a macroscopic volume. Assumptions on the dominant type of features can be checked against the ratio of magnetic and nuclear scattering. The materials investigated in this work are commercial-purity Fe-Cr alloys of nominal compositions Fe-2.5Cr, Fe-5Cr, Fe- 9Cr and Fe-12.5Cr (at %). Neutron irradiation was performed in the reactor BR2 at Mol (Belgium) at a temperature of 300 deg. C and neutron flux of 9 x 10 13 cm -2 s -1 (E > 1 MeV) [Matijasevic, JNM 377 (2008) 147]. The neutron exposures expressed in units of displacements per atom correspond to 0.6 and 1.5 dpa. A wavelength of 0.58 nm and three detector-sample distances of 1, 4 and 16 m were used in the SANS experiments carried out at the SANS-2 facility of GKSS Geesthacht (Germany). The samples were placed in a saturation magnetic field in order to separate magnetic and nuclear contributions. The scattering curves obtained for the unirradiated conditions of the four Fe-Cr alloys were taken as reference. We have found that the volume fraction of scatterers slightly increases with neutron exposure (Fe-9Cr) or exhibits a saturation-like behaviour (Fe-2.5Cr, Fe-5Cr and Fe-12.5Cr) and that the volume fraction at 1.5 dpa is an increasing function of the chromium level with a slight increase up to 9 at%Cr and a steep increase between 9 and 12.5 at%Cr. The radii of irradiation-induced scatterers are essentially less than 8 nm and

  16. Surface properties of a nanocrystalline Fe-Ni-Nb-B alloy after neutron irradiation

    International Nuclear Information System (INIS)

    Pavuk, M.; Sitek, J.; Sedlackova, K.

    2014-01-01

    In this work, we studied the impact of a neutron radiation on the surface properties of the nanocrystalline (Fe_0_._2_5Ni_0_._7_5)_8_1Nb_7B_1_2 alloy. Changes in topography and domain structure were observed by means of magnetic force microscopy (MFM). (authors)

  17. Effect of solute atom concentration on vacancy cluster formation in neutron-irradiated Ni alloys

    Science.gov (United States)

    Sato, Koichi; Itoh, Daiki; Yoshiie, Toshimasa; Xu, Qiu; Taniguchi, Akihiro; Toyama, Takeshi

    2011-10-01

    The dependence of microstructural evolution on solute atom concentration in Ni alloys was investigated by positron annihilation lifetime measurements. The positron annihilation lifetimes in pure Ni, Ni-0.05 at.%Si, Ni-0.05 at.%Sn, Ni-Cu, and Ni-Ge alloys were about 400 ps even at a low irradiation dose of 3 × 10 -4 dpa, indicating the presence of microvoids in these alloys. The size of vacancy clusters in Ni-Si and Ni-Sn alloys decreased with an increase in the solute atom concentration at irradiation doses less than 0.1 dpa; vacancy clusters started to grow at an irradiation dose of about 0.1 dpa. In Ni-2 at.%Si, irradiation-induced segregation was detected by positron annihilation coincidence Doppler broadening measurements. This segregation suppressed one-dimensional (1-D) motion of the interstitial clusters and promoted mutual annihilation of point defects. The frequency and mean free path of the 1-D motion depended on the solute atom concentration and the amount of segregation.

  18. Surface Properties of a Nanocrystalline Fe-Ni-Nb-B Alloy After Neutron Irradiation

    Science.gov (United States)

    Pavùk, Milan; Sitek, Jozef; Sedlačková, Katarína

    2014-09-01

    The effect of neutron radiation on the surface properties of the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy was studied. Firstly, amorphous (Fe0.25Ni0.75)81Nb7B12 ribbon was brought by controlled annealing to the nanocrystalline state. After annealing, the samples of the nanocrystalline ribbon were irradiated in a nuclear reactor with neutron fluences of 1×1016cm-2 and 1 × 1017cm-2 . By utilizing the magnetic force microscopy (MFM), topography and a magnetic domain structure were recorded at the surface of the ribbon-shaped samples before and after irradiation with neutrons. The results indicate that in terms of surface the nanocrystalline (Fe0.25Ni0.75)81Nb7B12 alloy is radiation-resistant up to a neutron fluence of 1 × 1017cm-2 . The changes in topography observed for both irradiated samples are discussed

  19. Minaturized disk bend tests of neutron-irradiated path A type alloys

    International Nuclear Information System (INIS)

    Lee, M.; Sohn, D.S.; Grant, N.J.; Harling, O.K.

    1983-01-01

    Path A Prime Candidate Alloy (PCA) has been rapidly solidified and consoliated by extrusion. Twenty percent CW samples, precision TEM disks, 3 phi x 0.254 mm, were irradiated in the mixed flux of the Oak Ridge HFIR reactor up to approx. 8.5 dpa (360 appm He) and approx. 34 dpa (3100 appm He) at 300, 400, 500 and 600 0 C. Similar samples of conventionally processed PCA were also irradiated for comparison. Mechanical properties were characterized using a minaturized disk bend test (MDBT) developed at MIT. These tests indicate major decreases in strength and ductility especially for the 500 and 600 0 C irradiations. No major differences were found between this first version of a rapidly solidified and extruded PCA type alloy and conventionally processed PCA

  20. 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.

  1. Rapid analysis of molybdenum contents in molybdenum master alloys by X-ray fluorescence technique

    International Nuclear Information System (INIS)

    Tongkong, P.

    1985-01-01

    Determination of molybdenum contents in molybdenum master alloy had been performed using energy dispersive x-ray fluorescence (EDX) technique where analysis were made via standard additions and calibration curves. Comparison of EDX technique with other analyzing techniques, i.e., wavelength dispersive x-ray fluorescence, neutron activation analysis and inductive coupled plasma spectrometry, showed consistency in the results. This technique was found to yield reliable results when molybdenum contents in master alloys were in the range of 13 to 50 percent using HPGe detector or proportional counter. When the required error was set at 1%, the minimum analyzing time was found to be 30 and 60 seconds for Fe-Mo master alloys with molybdenum content of 13.54 and 49.09 percent respectively. For Al-Mo master alloys, the minimum times required were 120 and 300 seconds with molybdenum content of 15.22 and 47.26 percent respectively

  2. Tensile and fracture toughness properties of copper alloys and their HIP joints with austenitic stainless steel in unirradiated and neutron irradiated condition

    International Nuclear Information System (INIS)

    Taehtinen, S.; Pyykkoenen, M.; Singh, B.N.; Toft, P.

    1998-03-01

    The tensile strength and ductility of unirradiated CuAl25 IG0 and CuCrZr alloys decreased continuously with increasing temperature up to 350 deg C. Fracture toughness of unirradiated CuAl25 IG0 alloy decreased continuously with increasing temperature from 20 deg C to 350 deg C whereas the fracture toughness of unirradiated CuCrZr alloy remained almost constant at temperatures up to 100 deg C, was decreased significantly at 200 deg C and slightly increased at 350 deg C. Fracture toughness of HIP joints were lower than that of corresponding copper alloy and fracture path in HIP joint specimen was always within copper alloy side of the joint. Neutron irradiation to a dose level of 0.3 dpa resulted in hardening and reduction in uniform elongation to about 2-4% at 200 deg C in both copper alloys. At higher temperatures softening was observed and uniform elongation increased to about 5% and 16% for CuAl25 IG0 and CuCrZr alloys, respectively. Fracture toughness of CuAl25 IG0 alloy reduced markedly due to neutron irradiation in the temperature range from 20 deg C to 350 deg C. The fracture toughness of the irradiated CuCrZr alloy also decreased in the range from 20 deg C to 350 deg C, although it remained almost unaffected at temperatures below 200 deg C and decreased significantly at 350 deg C when compared with that of unirradiated CuCrZr alloy. (orig.)

  3. On the ductile-to-brittle transition behavior of martensitic alloys neutron irradiated to 26 dpa

    International Nuclear Information System (INIS)

    Hu, W.L.; Gelles, D.S.

    1987-01-01

    Charpy impact tests were conducted on specimens made of HT-9 and 9Cr-1Mo in various heat treatment conditions which were irradiated in EBR-II to 26 dpa at 390 to 500 0 C. The results are compared with previous results on specimens irradiated to 13 dpa. HT-9 base metal irradiated at low temperatures showed a small additional increase in ductile brittle transition temperature and a decrease in upper shelf energy from 13 to 26 dpa. No fluence effect was observed in 9Cr-1Mo base metal. The 9Cr-1Mo weldment showed degraded DBTT but improved USE response compared to base metal, contrary to previous findings on HT-9. Significant differences were observed in HT-9 base metal between mill annealed material and normalized and tempered material. The highest DBTT for HT-9 alloys was 50 0 C higher than for the worst case in 9Cr-1Mo alloys. Fractography and hardness measurements were also obtained. Significant differences in fracture appearance were observed in different product forms, although no dependence on fluence was observed. Failure was controlled by the preirradiation microstructure

  4. Swelling of austenitic iron-nickelchromium ternary alloys during fast neutron irradiation

    International Nuclear Information System (INIS)

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

    1984-01-01

    Swelling data are now available for 15 iron-nickel-chromium ternary alloys irradiated to exposures as high as 110 displacements per atom (dpa) in Experimental Breeder Reactor-II (EBR-II) between 400 and 650 0 C. These data confirm trends observed at lower exposure levels and extend the generality of earlier conclusions to cover a broader range of composition and temperature. It appears that all austenitic iron-nickel-chromium ternary alloys eventually approach an intrinsic swelling rate of about1%/dpa over a range of temperature even wider than studied in this experiment. The duration of the transient regime that precedes the attainment of this rate is quite sensitive to nickel and chromium content, however. At nickel and chromium levels typical of 300 series steels, swelling does not saturate at engineering-relevant levels. However, there appears to be a tendency toward saturation that increases with declining temperature, increasing nickel and decreasing chromium levels. Comparisons of these results are made with those of similar studies conducted with charged particles. Conclusions are then drawn concerning the validity of charged particle simulation studies to determine the compositional and temperature dependence of swelling

  5. Phase instability of alloys caused by transmutation effects during neutron irradiation

    International Nuclear Information System (INIS)

    Platov, Yu.M.; Pletnev, M.N.

    1994-01-01

    A theory of the phase changes in a two-phase binary A-B alloy in the coarsening condition caused by burnout of solute B due to nuclear reactions is presented. It is shown that this burnout process introduces diffusion redistribution of solute between second phase precipitates and solid solution. The burnout induced solute flux away from second phase precipitates to solid solution maintaining the concentration of element B in the vicinity to its solubility limit and stimulates, thus, the second phase particle dissolution. This occurs in addition to a process decreasing their sizes as a result of direct burnout of atoms B in the precipitates. In the framework of the theory developed here, analytical expressions describing time evolution of the precipitate size distributions, changes of mean radius and number density of the precipitates, and second phase dissolution times are obtained. On the basis of these results and numerical calculations for aluminium-scandium alloy, it is shown that the burnout processes can induce essential phase changes, and thus cause significant changes of the properties of irradiated materials at high neutron fluences. ((orig.))

  6. Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

    International Nuclear Information System (INIS)

    Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

    1979-01-01

    An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n.cm -2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg.mm -2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

  7. Effect of thermal neutron irradiation on mechanical properties of alloys for HTR core applications

    International Nuclear Information System (INIS)

    Ogawa, Yutaka; Kondo, Tatsuo; Ishimoto, Kiyoshi; Ohtsuka, Tamotsu

    1979-02-01

    An industrial heat of Hastelloy-X containing 2.3 ppm boron was creep-tested at 900 0 C after irradiating thermal neutrons by 6.6 x 10 20 n/cm 2 at temperatures 670 to 880 0 C in JMTR. Significant reduction in rupture life and ductility was observed, and large shift of accelerated deformation stage to short time side was also apparent at comparatively high stresses. Below about 2.2 kg/mm 2 , apparent relief from the degradation was seen. The elongation, however, was found to be due to the formation of numerous intergranular cracks in the premature stage of deformation. Based on the post irradiation tensile properties of several industrial alloys the degree of the ductility loss was found to be nearly dependent on the boron content of the alloys. The post irradiation tensile tests for a special low boron grade heat revealed the means of protecting materials from the effect to be feasible. (author)

  8. Solid solutions of hydrogen in niobium, molybdenum and their alloys

    International Nuclear Information System (INIS)

    Ishikawa, T.T.

    1981-01-01

    The solubility of hydrogen in niobium, molybdenum and niobium-molybdenum alloys with varying atomic fraction of molybdenum from 0.15 to 0.75 was measured on the temperature range of 673 0 K to 1273 0 k for one atmosphere hydrogen pressure. The experimental technique involved the saturation of the solvent metal or alloy with hydrogen, followed by quenching and analysis of the solid solution. The results obtained of hydrogen solubility are consistent with the quasi-regular model for the dilute interstitial solid solutions. The partial molar enthalpy and partial molar entropy in excess of the dissolved hydrogen atoms were calculated from data of solubility versus reciprocal doping temperature. The variation of the relative partial molar enthalpy of hydrogen dissolved in niobium-molybdenum alloys, with the increase of molybdenum content of the alloy was analized. (Author) [pt

  9. The Primary Origin of Dose Rate Effects on Microstructural Evolution of Austenitic Alloys During Neutron Irradiation

    International Nuclear Information System (INIS)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.

    2002-01-01

    The effect of dose rate on neutron-induced microstructural evolution was experimentally estimated. Solution-annealed austenitic model alloys were irradiated at approximately 400 degrees C with fast neutrons at seven different dose rates that vary more than two orders difference in magnitude, and two different doses were achieved at each dose rate. Both cavity nucleation and growth were found to be enhanced at lower dose rate. The net vacancy flux is calculated from the growth rate of cavities that had already nucleated during the first cycle of irradiation and grown during the second cycle. The net vacancy flux was found to be proportional to (dpa/sec) exp (1/2) up to 28.8 dpa and 8.4 x 10 exp (-7) dpa/sec. This implies that mutual recombination dominates point defect annihilation, in this experiment even though point defect sinks such as cavities and dislocations were well developed. Thus, mutual recombination is thought to be the primary origin of the effect of dose rate on microstructural evolution

  10. Internal friction of Fe-B alloys neutron irradiated at low temperature

    International Nuclear Information System (INIS)

    Kitajima, Kazunori; Futagami, Koji; Abe, Hironobu; Yoshida, Hiroyuki.

    1975-01-01

    Measurements were made on the internal friction of Fe-B alloys irradiated by neutron at 16 0 K to the dose of 3x10 16 nvt (>1 MeV) and 6x10 17 nvt (thermal). Boron was used to enhance the production of defects by the nuclear transformation B 10 (n,α)Li 7 . Relaxation peaks were found in specimens containing dispersed fine precipitates of NbB 2 in range of B 500--7200 wt ppm and Nb 2000--30000 wt ppm. The most prominent peak is the one with the peak temperature of 169 0 K at the frequency of 264 c/sec. Activation energy determined from the peak shift is 0.28+-0.01 eV, which is nearly equal to that of migration of self-interstitial reported on pure iron. However activation energy of the decay of peaks by annealing is about 0.7 eV. Interpretation was presented that the peak may be attributed to re-orientation of self-interstitials loosely bound to a boron atom. (auth.)

  11. Molybdenum-A Key Component of Metal Alloys

    Science.gov (United States)

    Kropschot, S.J.

    2010-01-01

    Molybdenum, whose chemical symbol is Mo, was first recognized as an element in 1778. Until that time, the mineral molybdenite-the most important source of molybdenum-was believed to be a lead mineral because of its metallic gray color, greasy feel, and softness. In the late 19th century, French metallurgists discovered that molybdenum, when alloyed (mixed) with steel in small quantities, creates a substance that is remarkably tougher than steel alone and is highly resistant to heat. The alloy was found to be ideal for making tools and armor plate. Today, the most common use of molybdenum is as an alloying agent in stainless steel, alloy steels, and superalloys to enhance hardness, strength, and resistance to corrosion.

  12. X-ray target with substrate of molybdenum alloy

    International Nuclear Information System (INIS)

    Hirsch, H.H.

    1980-01-01

    Rotary targets for x-ray tubes are provided comprising a molybdenum base body alloyed with a stabilizing proportion of iron, silicon, cobalt, tantalum, niobium, hafnium, stable metal oxide, or a mixture of the preceding

  13. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects

    International Nuclear Information System (INIS)

    Meslin-Chiffon, E.

    2007-11-01

    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)

  14. Development of Silicide Coating on Molybdenum Alloy Cladding

    International Nuclear Information System (INIS)

    Lim, Woojin; Ryu, Ho Jin

    2015-01-01

    The molybdenum alloy is considered as one of the accident tolerant fuel (ATF) cladding materials due to its high temperature mechanical properties. However, molybdenum has a weak oxidation resistance at elevated temperatures. To modify the oxidation resistance of molybdenum cladding, silicide coating on the cladding is considered. Molybdenum silicide layers are oxidized to SiO 2 in an oxidation atmosphere. The SiO 2 protective layer isolates the substrate from the oxidizing atmosphere. Pack cementation deposition technique is widely adopted for silicide coating for molybdenum alloys due to its simple procedure, homogeneous coating quality and chemical compatibility. In this study, the pack cementation method was conducted to develop molybdenum silicide layers on molybdenum alloys. It was found that the Mo 3 Si layer was deposited on substrate instead of MoSi 2 because of short holding time. It means that through the extension of holding time, MoSi 2 layer can be formed on molybdenum substrate to enhance the oxidation resistance of molybdenum. The accident tolerant fuel (ATF) concept is to delay the process following an accident by reducing the oxidation rate at high temperatures and to delay swelling and rupture of fuel claddings. The current research for Atf can be categorized into three groups: First, modification of existing zirconium-based alloy cladding by improving the high temperature oxidation resistance and strength. Second, replacing Zirconium based alloys with alternative metallic materials such as refractory elements with high temperature oxidation resistance and strength. Third, designing alternative fuel structures using ceramic and composite systems

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

    International Nuclear Information System (INIS)

    Singh, B.N.; Edwards, D.J.; Bilde-Soerensen, J.B.

    2004-10-01

    The phenomenon of plastic flow localization in the form of 'cleared' channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels are initiated during post-irradiation deformation 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 are formed already in the elastic regime and their density increases with increasing plastic strain. The channels appear to have been initiated at grain boundaries, twin boundaries, at relatively large inclusions and even at the previously formed cleared channels. Even though the channels are produced throughout the whole tensile test, no clear evidence has 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 through the opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimens activate dislocation sources at the sites of stress concentration at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these

  16. Characterization of the uranium--2 weight percent molybdenum alloy

    International Nuclear Information System (INIS)

    Hemperly, V.C.

    1976-01-01

    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

  17. A review of chromium, molybdenum, and tungsten alloys

    International Nuclear Information System (INIS)

    Klopp, W.D.

    1975-01-01

    The mechanical properties of chromium, molybdenum, and tungsten alloys are reviewed, with particular emphasis on high-temperature strength and low-temperature ductility. Precipitate strengthening is highly effective at 0.4-0.8 Tsub(m) in these metals, with HfC being most effective in tungsten and molybdenum, and Ta(B,C) most effective in chromium. Low-temperature ductility can be improved by alloying to promote rhenium ductilizing or solution softening. The low-temperature mechanical properties of these alloys appear related to electronic interactions rather than to the usual metallurgical considerations. (Auth.)

  18. Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu; Schneider, H.-C., E-mail: hans-christian.schneider@kit.edu; Möslang, A., E-mail: anton.moeslang@kit.edu

    2014-12-15

    Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects.

  19. Tensile behavior of RAFM alloys after neutron irradiation of up to 16.3 dpa between 250 and 450 °C

    International Nuclear Information System (INIS)

    Materna-Morris, E.; Schneider, H.-C.; Möslang, A.

    2014-01-01

    Tensile specimen of steel EUROFER97 and other alloys on the basis of RAFM steels such, as OPTIFER and F82H alloys, and Ga3X were irradiated and post-examined during a neutron irradiation program of up to 16.3 dpa between 250 and 450 °C in the HFR (High Flux Reactor) in the Netherlands. These tensile results were compared with former irradiation programs, with lower neutron doses of up to 0.8 and 2.4 dpa to quantify the difference in tensile strengthening. The average increase of tensile strength was in a range of 300 MPa between 0.8 and 16.3 dpa at temperatures of 250–300 °C. This behavior can be correlated with irradiation-induced changes in the microstructure. Most of the hardening can be attributed to dislocation loops, point defects or small precipitates as observed in boron-free alloys as F82H mod. and EUROFER97. Whereas the hardening in boron-containing alloys OPTIFER alloys and Ga3X can be correlated in addition with the combination of helium bubbles. At the highest irradiation and test temperature at 450 °C, all tensile data of all investigated materials were in the range of those of non-irradiated and irradiated material due to thermal aging effects

  20. Enhanced atomic mobility due to low dose neutron irradiation as measured in a 30Ag at % Zn alloy by Zener relaxation methods

    International Nuclear Information System (INIS)

    Halbwachs, M.; Hillairet, J.; Gonzalez, H.; Cost, J.

    1975-01-01

    Mean atomic jump rates were measured during and after short-time neutron irradiations of a fcc Ag-30 at. percent Zn alloy by monitoring the rate of the Zener relaxation. Measurements made at 40 0 C immediately after rapid suppression of the flux showed significant enhancement of the atomic mobility for a flux of approximately 10 11 fast n/cm 2 .sec and an irradiation time of only 30 sec. The subsequent decrease of the atomic mobility to the thermal equilibrium value was studied and found to follow simple exponential decay. The results are discussed in terms of the nature and concentration of mobile defects created by cascades, the mechanism for defect annihilation and the concentration of sinks

  1. Atomic-resolution study of homogeneous radiation-induced precipitation in a neutron-irradiated W-10 at. % Re alloy. MSC report No. 5014

    International Nuclear Information System (INIS)

    Herschitz, R.; Seidman, D.N.

    1983-06-01

    The phenomenon of radiation-induced precipitation has been investigated in a W-10 at. % Re alloy using the atom-probe field-ion microscope. Results show a significant alteration of the microstructure of this alloy as a result of the fast-neutron irradiation. Precipitates with the composition approx. WRe (sigma phase) were detected at a density of 10 16 cm -3 . Coherent, semicoherent and possibly incoherent precipitates of the sigma phase have been observed. They were not associated with either linear or planar defects, or with any impurity atoms; i.e. a true homogeneous radiation-induced precipitation occurs in this alloy. A physical argument is presented for the nucleation of the sigma phase precipitates in the vicinity of displacement cascades produced by primary knock-on atoms. It is suggested that the nucleation of the sigma phase is due to the formation of tightly-bound mobil mixed dumbells which react to form an immobile rhenium cluster. The growth of this cluster into a precipitate is most likely driven by the irreversible vacancy: self-interstitial atom (SIA) annihilation reaction, as suggested recently by Cauvin and Martin. A mechanism for the suppression of voids, in this alloy, is presented which is self-consistent with the homogeneous radiation-induced precipitation mechanism

  2. Investigation of the thermo-mechanical behavior of neutron-irradiated Fe-Cr alloys by self-consistent plasticity theory

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Xiazi [State Key Laboratory for Turbulence and Complex System, Department of Mechanics and Engineering Science, College of Engineering, Peking University, Beijing 100871 (China); CAPT, HEDPS and IFSA Collaborative Innovation Center of MoE, BIC-ESAT, Peking University, Beijing 100871 (China); Terentyev, Dmitry [Structural Material Group, Institute of Nuclear Materials Science, SCK CEN, Mol (Belgium); Yu, Long [State Key Laboratory for Turbulence and Complex System, Department of Mechanics and Engineering Science, College of Engineering, Peking University, Beijing 100871 (China); Bakaev, A. [Structural Material Group, Institute of Nuclear Materials Science, SCK CEN, Mol (Belgium); Jin, Zhaohui [School of Materials Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240 (China); Duan, Huiling, E-mail: hlduan@pku.edu.cn [State Key Laboratory for Turbulence and Complex System, Department of Mechanics and Engineering Science, College of Engineering, Peking University, Beijing 100871 (China); CAPT, HEDPS and IFSA Collaborative Innovation Center of MoE, BIC-ESAT, Peking University, Beijing 100871 (China)

    2016-08-15

    The thermo-mechanical behavior of non-irradiated (at 223 K, 302 K and 573 K) and neutron irradiated (at 573 K) Fe-2.5Cr, Fe-5Cr and Fe-9Cr alloys is studied by a self-consistent plasticity theory, which consists of constitutive equations describing the contribution of radiation defects at grain level, and the elastic-viscoplastic self-consistent method to obtain polycrystalline behaviors. Attention is paid to two types of radiation-induced defects: interstitial dislocation loops and solute rich clusters, which are believed to be the main sources of hardening in Fe-Cr alloys at medium irradiation doses. Both the hardening mechanism and microstructural evolution are investigated by using available experimental data on microstructures, and implementing hardening rules derived from atomistic data. Good agreement with experimental data is achieved for both the yield stress and strain hardening of non-irradiated and irradiated Fe-Cr alloys by treating dislocation loops as strong thermally activated obstacles and solute rich clusters as weak shearable ones. - Highlights: • A self-consistent plasticity theory is proposed for irradiated Fe-Cr alloys. • Both the irradiation-induced hardening and plastic flow evolution are studied. • Dislocation loops and solute rich clusters are considered as the main defects. • Numerical results of the proposed model match with corresponding experimental data.

  3. 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.

  4. Influence of neutron irradiation on the magnetic properties of the Fe Ni pure alloy and with impurities of Si and Mo

    International Nuclear Information System (INIS)

    Lucki, George

    1971-01-01

    Hysteresis loop, Initial permeability and Curie Temperature measurements were conducted on several pure and polluted (with Si and Mo) Fe Ni 50-50% at. alloys. Isochronal annealings were performed between 25 deg 65 deg C, on each composition in three different ways: quenched (anisotropic) samples; quenched and irradiated samples; quenched irradiated samples annealed with saturating magnetic field. The experiment showed a sharp decrease in all parameters of the polluted alloys. Fast neutron irradiation results indicated that the magnetic properties are affected by the defects created during irradiation. The effect of thermal treatment, magnetic annealing and irradiation is greatest in anisotropic alloys. It is considered that magnetic annealing introduces a uniaxial anisotropy that tends to increase the remanence and hence the squareness of the hysteresis loop; but an increase in both remanence and coercivity was measured even in absence of the magnetic field. Magnetic after effect has been detected and a simple model for the diffusion of defects is presented. Many models have been proposed to explain the resultant properties, the most feasible being that based upon short-range ordering, proposed by Neel and Taniguchi, together with the interesting hypothesis of Heidenreich and Nesbitt. (author)

  5. High-resolution electron microscopy studies of the precipitation of copper under neutron irradiation in an Fe-1.3WT% Cu alloy

    International Nuclear Information System (INIS)

    Nicol, A. C.

    1998-01-01

    We have studied by electron microscopy the copper-rich precipitates in an Fe-1.3wt%Cu model alloy irradiated with neutrons to doses of 8.61 x 10 -3 dpa and 6.3 x 10 -2 dpa at a temperature of ∼270 C. In the lower dose material a majority (ca. 60%)of the precipitates visible in high-resolution electron microscopy were timed 9R precipitates of size ∼2-4 nm, while ca. 40% were untwinned. In the higher dose material, a majority (ca. 75%) of visible precipitates were untwinned although many still seemed to have a 9R structure. The average angle α between the herring-bone fringes in the twin variants was measured as 125 degree, not the 129 degree characteristic of precipitates in thermally-aged and electron-irradiated material immediately after the bcc->9R martensitic transformation. We argue that these results imply that the bcc->9R transformation of small (<4 nm) precipitates under neutron irradiation takes place at the irradiation temperature of 270 C rather than after subsequent cooling. Preliminary measurements showed that precipitate sizes did not depend strongly on dose, with a mean diameter of 3.4 ± 0.7 nm for the lower dose material, and 3.0 ± 0.5 nm for the higher dose material. This result agrees with the previous assumption that the lack of coarsening in precipitates formed under neutron irradiation is a consequence of the partial dissolution of larger precipitates by high-energy cascades

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

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N.; Edwards, D.J.; Bilde-Soerensen, J.B

    2004-10-01

    The phenomenon of plastic flow localization in the form of 'cleared' channels has been frequently observed in neutron irradiated metals and alloys for more than 40 years. So far, however, no experimental evidence as to how and where these channels are initiated during post-irradiation deformation 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 are formed already in the elastic regime and their density increases with increasing plastic strain. The channels appear to have been initiated at grain boundaries, twin boundaries, at relatively large inclusions and even at the previously formed cleared channels. Even though the channels are produced throughout the whole tensile test, no clear evidence has 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 through the opposite twin boundary and continuing further into the grain. In some cases channels have been found to penetrate through grain boundaries too. It is suggested that the high stress levels reached during deformation of the irradiated specimens activate dislocation sources at the sites of stress concentration at the boundaries and inclusions. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications

  7. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C. (ITER R and D Task no. T213)

    International Nuclear Information System (INIS)

    Singh, B.N.; Stubbins, J.F.; Toft, P.

    2000-03-01

    The fatigue behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of ∼2.5 x 10 17 n/m 2 s (E> 1 MeV) to influence levels of 1.0 - 1.5 x 10 24 n/m 2 (E> 1 MeV) at 250 and 350 deg. C. These irradiations were carried out in temperature controlled rigs where the irradiation temperature was monitored and controlled continuously throughout the whole irradiation experiment. Both unirradiated and irradiated specimens were fatigue tested in vacuum at the irradiation temperatures of 250 and 350 deg. C in a strain controlled mode with a loading frequency of 0.5Hz. Post-fatigue microstructures were examined using transmission electron microscopy and the fracture surfaces were investigated using scanning electron microscope. The present investigations demonstrated that the fatigue life decreases with increasing temperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures and the fracture surface morphology. Finally, the main conclusions and their implications are summarised. (au)

  8. The response of dispersion-strengthened copper alloys to high fluence neutron irradiation at 415 degrees C

    International Nuclear Information System (INIS)

    Edwards, D.J.; Newkirk, J.W.; Garner, F.A.; Hamilton, M.L.; Nadkarni, A.; Samal, P.

    1993-01-01

    Various oxide-dispersion-strengthened copper alloys have been irradiated to 150 dpa at 415 degrees C in the Fast Flux Test Facility (FFTF). The Al 2 O 3 -strengthened GlidCop TM alloys, followed closely by a HfO 2 -strengthened alloy, displayed the best swelling resistance, electrical conductivity, and tensile properties. The conductivity of the HfO 2 -strengthened alloy reached a plateau at the higher levels of irradiation, instead of exhibiting the steady decrease in conductivity observed in the other alloys. A high initial oxygen content results in significantly higher swelling for a series of castable oxide-dispersion-strengthened alloys, while a Cr 2 O 3 -strengthened alloy showed poor resistance to radiation

  9. A review of the effect of neutron irradiation on the deformation behaviour of copper and copper alloys

    International Nuclear Information System (INIS)

    Higgy, H.R.

    1976-08-01

    The basic mechanisms of irradiation hardening are described. The effects of neutron dose, alloying and pre-irradiation deformation on the deformation behaviour of neutron-irradiatied copper and its alloys are considered. The discrepancy in the reported data is discussed. Substitutional and interstitial additions are found to influence the rate of irradiation hardening, while pre-irradiation deformation has no influence. The deformation behaviour of copper is found to alter as a result of irradiation and alloying. (author)

  10. Evaluation of oxide dispersion strengthened (ODS) molybdenum and molybdenum-rhenium alloys

    International Nuclear Information System (INIS)

    Mueller, A.J.; Bianco, R.; Buckman, R.W. Jr.

    1999-01-01

    Oxide dispersion strengthened (ODS) molybdenum alloys being developed for high temperature applications possess excellent high temperature strength and creep resistance. In addition they exhibit a ductile-to-brittle transition temperature (DBIT) in the worked and stress-relieved condition under longitudinal tensile load well below room temperature. However, in the recrystallized condition, the DBTT maybe near or above room temperature, depending on the volume fraction of oxide dispersion and the amount of prior work. Dilute rhenium additions (7 and 14 wt.%) to ODS molybdenum were evaluated to determine their effect on low temperature ductility. The addition of 7 wt.% rhenium to the ODS molybdenum did not significantly enhance the mechanical properties. However, the addition of 14 wt.% rhenium to the ODS molybdenum resulted in a DBTT well below room temperature in both the stress-relieved and recrystallized condition. Additionally, the tensile strength of ODS Mo-14Re is greater than the base ODS molybdenum at 1,000 to 1,250 C

  11. Study and characterization of FeNi and NiCr(80-20) % w alloys, during and after neutron irradiation, using the resistivity method

    International Nuclear Information System (INIS)

    Otero, Mauro Pereira

    1978-01-01

    We have used the resistivity method with and without neutron irradiation to study the parameters that appear in the Order-Disorder Transitions of Fe Ni(50-50)% at. and Ni Cr( 80 - 20) % w. alloys. The results obtained with Fe Ni are in agreement with those obtained by Marchand at the University of Grenoble. Several isothermal annealings were made in the range 400 - 302 deg C in which T c (Order-Disorder Transition Critical Temperature) was determined between 327 and 310 deg C. The activation energy obtained was E a = 0,49 eV and is in agreement with works of Marchand, Dienes and Damask. As for Ni Cr(80-20)% the following has been done: a) Electrical Properties characterizations, having in mind the technological applications; b) Linear and isothermal annealings were performed to determine the Order-Disorder Transition Critical Temperature (I ) supported by hypothesis made, taking into account the Yano's and Taylor's marks. The-result is T c = (536 +- 4) deg C; c) determination of activation energy E a = (1,36 +- 0.14) eV. The resistivity measurements mere performed by means of the classical 4-wire method. An anisotropy of electrical resistivity was found to exist depending on the sense of the applied electrical field. (author)

  12. Impact of neutron irradiation on thermal helium desorption from iron

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

    2017-06-15

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  13. Effects of neutron irradiation at 4500C and 16 dpa on the properties of various commercial copper alloys

    International Nuclear Information System (INIS)

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

    1985-01-01

    High-purity copper and eight copper alloys were irradiated to approx.16 dpa at approx.450 0 C in the MOTA experiment in FFTF. These alloys were also examined after aging at 400 0 C for 1000 hours. The radiation-induced changes in the electrical conductivity, tensile properties, and density were measured and compared to those of the aged materials. The changes in conductivity can be either positive or negative depending on the alloy. Changes in tensile properties of most, but not all, of the alloys seem to be primarily dependent on thermal effects rather than the effect of atomic displacements. Radiation at 450 0 C induced changes in density varying from 0.66% densification to 16.6% swelling. The latter occurred in Cu-O.1% Ag and implies a swelling rate of at least 1%/dpa. 6 references, 3 figures, 2 tables

  14. Amorphous-to-crystalline phase transformation by neutron irradiation of the alloy Fe/sub 83/B/sub 17/

    Energy Technology Data Exchange (ETDEWEB)

    Weis, J.; Gabris, F.; Cerven, I.; Sitek, J. (Slovenska Vysoka Skola Technicka, Bratislava (Czechoslovakia))

    1984-03-01

    The purpose of the present work is to investigate the structural changes of amorphous Fe/sub 83/B/sub 17/ alloy after irradiation with fast neutrons ( > 1 MeV) and to compare with the crystallization behaviour of the amorphous Fe/sub 83/B/sub 17/ alloy after annealing. The structural changes were studied by Moessbauer spectroscopy and X-ray diffraction with the usual Fourier analysis.

  15. Damage rates in neutron irradiated FeCo and FeCo2V ordered and disordered alloys

    International Nuclear Information System (INIS)

    Riviere, J.P.; Dinhut, J.F.

    1979-01-01

    Ordered and disordered samples of FeCo and FeCo2V alloys have been irradiated at liquid hydrogen temperature with fission neutrons up to an integrated dose of about 7.2 x 10 17 n/cm 2 (E > 1 MeV). During the irradiation, the resistivity increases continuously due to point defect production. (author)

  16. Microstructure and mechanical properties of multi-components rare earth oxide-doped molybdenum alloys

    International Nuclear Information System (INIS)

    Zhang Guojun; Sun Yuanjun; Zuo Chao; Wei Jianfeng; Sun Jun

    2008-01-01

    Pure molybdenum and molybdenum alloys doped with two- or three-components rare earth oxide particles were prepared by powder metallurgy. Both the tensile property and fracture toughness of the pure molybdenum and multi-components rare earth oxide-doped molybdenum alloys were determined at room temperature. The multi-components rare earth oxide-doped molybdenum alloys are fine grained and contain a homogeneous distribution of fine particles in the submicron and nanometer size ranges, which is why the molybdenum alloys have higher strength and fracture toughness than pure molybdenum. Quantitative analysis is used to explain the increase in yield strength with respect to grain size and second phase strengthening. Furthermore, the relationship between the tensile properties and microstructural parameters is quantitatively established

  17. Applications of neutron irradiation

    International Nuclear Information System (INIS)

    Ito, Yasuo

    1999-01-01

    The present state of art of applications of neutron irradiation is overviewed taking neutron activation analysis, prompt gamma-ray analysis, fission/alpha track methods, boron neutron capture therapy as examples. What is common among them is that the technologies are nearly matured for wide use by non- nuclear scientists. But the environment around research reactors is not prospective. These applications should be encouraged by incorporating in the neutron science society. (author)

  18. Direct observation of solute segregation to voids in a fast-neutron irradiated (Mo/1.0 at. % Ti alloy

    International Nuclear Information System (INIS)

    Wagner, A.; Seidman, D.N.

    1978-11-01

    The atom-probe field-ion microscope was used to study segregation effects to voids in a Mo--Ti alloy which had been irradiated with fast neutrons. The Ti does not segregate significantly to voids, concentration of Ti in solid solution and the spacial distribution of Ti was not affected by irradiation, carbon was not detected, resolution of TiC or MoC precipitates did not occur

  19. Dendrite tungsten liquation in molybdenum alloys

    International Nuclear Information System (INIS)

    Kantor, M.M.; Ageeva, E.N.; Kolotinskij, V.N.

    1992-01-01

    A study was made on primary crystallization structure of ingots of Mo-W-B system alloys with electron microscopy were used to establish, that cells and cellular dendrites were the main elements of primary crystallization structure. Method of local X-ray spectral analysis enabled to establish, that intracrystallite liquation at cellular growth developed more intensively, as compared to the case of cellular dendrite formation. Change of boron content in alloys didn't practically affect the degree of development of intracrystallite W liquation in Mo

  20. Weldability of molybdenum and its alloy sheet, 1

    International Nuclear Information System (INIS)

    Matsuda, Fukuhisa; Ushio, Masao; Nakata, Kazuhiro; Edo, Yoshiaki

    1979-01-01

    Basic weldability of electron-beam melted pure molybdenum has been examined in electron-beam welding in high vacuum and GTA welding in pure and air mixed argon atmospheres by paying attention to weld defects such as hot cracking and porosity in weld metal and also mechanical properties of welded joint in comparison with conventional TZM alloys. The main conclusions obtained were as follows; (1) The weld metals of electron-beam melted pure molybdenum with electron-beam and GTA weldings in pure and air mixed argon atmosphere up to about 1% were almost porosity free. However, large amount of oxygen content of 200 ppm in powder-metallurgy TZM alloy made very porous weld bead in electron-beam welding in high vacuum. Therefore, oxygen content in base metal should be lowered to the minimum, that is, less than 10 ppm, especially in electron-beam welding in high vacuum. (2) Hot cracking occurred in the weld metal of GTA welding when air content in argon atmosphere exceeded about 0.6% for electron-beam melted pure molybdenum and powder metallurgy TZM alloy. In less than 0.26% air, no hot cracking were observed in this experiment. Moreover, in electron-beam welding, no hot cracking was observed in weld metals for both materials. In order to prevent the formation of hot cracking, the purity of welding atmosphere should be kept as high as possible. (3) Joint efficiency of the welded joint of electron-beam melted pure molybdenum with electron-beam welding was 50 to 60% to base metal at room temperature and 500 0 C and almost 100% at 1000 0 C. Those of GTA welds in pure and 0.13% air mixed argon atmospheres were fairly lower than those in electron-beam welding for each testing temperature. (author)

  1. 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.

  2. Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios*1

    Science.gov (United States)

    Mansur, L. K.; Grossbeck, M. L.

    1988-07-01

    Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

  3. Uranium-molybdenum alloys containing 0,5 to 3 per cent by weight of molybdenum

    International Nuclear Information System (INIS)

    Lehmann, J.

    1959-01-01

    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 dilatation 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 α' and α'' and the alloys containing 2 and 3 per cent Mo give one martensite with a band structure. (author) [fr

  4. Evaluation of ferritic alloy Fe-2 1/4Cr-1Mo after neutron irradiation: Microstructural development

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1986-10-01

    As part of a program to provide a data base on the bainitic alloy Fe-2-1/4-1Mo for fusion energy applications, microstructural examinations are reported for nine specimen conditions for 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510 0 C. Void swelling is found following irradiation at 400 0 C to 480 0 C. Concurrently dislocation structure and precipitation developed. Peak void swelling, void density, dislocation density and precipitate number density formed at the lowest temperature, approximately 400 0 C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior

  5. Neutron irradiation therapy machine

    International Nuclear Information System (INIS)

    1980-01-01

    Conventional neutron irradiation therapy machines, based on the use of cyclotrons for producing neutron beams, use a superconducting magnet for the cyclotron's magnetic field. This necessitates complex liquid He equipment and presents problems in general hospital use. If conventional magnets are used, the weight of the magnet poles considerably complicates the design of the rotating gantry. Such a therapy machine, gantry and target facilities are described in detail. The use of protons and deuterons to produce the neutron beams is compared and contrasted. (U.K.)

  6. Strengthening and elongation mechanism of Lanthanum-doped Titanium-Zirconium-Molybdenum alloy

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Ping, E-mail: huping1985@126.com [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Hu, Bo-liang; Wang, Kuai-she; Song, Rui; Yang, Fan [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Yu, Zhi-tao [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Tan, Jiang-fei [School of Metallurgy Engineering, Xi’an University of Architecture and Technology, Xi’an 710055 (China); Cao, Wei-cheng; Liu, Dong-xin; An, Geng [Jinduicheng Molybdenum Co., Ltd., Xi’an 710068 (China); Guo, Lei [Ruifulai Tungsten & Molybdenum Co., Ltd., Xi’an 721914 (China); Yu, Hai-liang [School of Mechanical, Materials and Mechatronics Engineering, University of Wollongong, NSW 2522 (Australia)

    2016-12-15

    The microstructural contributes to understand the strengthening and elongation mechanism in Lanthanum-doped Titanium-Zirconium-Molybdenum alloy. Lanthanum oxide particles not only act as heterogeneous nucleation core, but also act as the second phase to hinder the grain growth during sintering crystallization. The molybdenum substrate formed sub-grain under the effect of second phase when the alloy rolled to plate.

  7. Special features of nickel-molybdenum alloy electrodeposition onto screen-type cathodes

    International Nuclear Information System (INIS)

    Aleksandrova, G.S.; Varypaev, V.N.

    1982-01-01

    Electrolytic nickel-molybdenum alloy, which has a rather low hydrogen overpotential and high corrosion resistance, is of interest as cathode material in industrial electrolysis. Screen-type electrodes with a nickel-molybdenum coating can be used as nonconsumable cathodes in water-activated magnesium-alloy batteries

  8. Manufacture and properties of molybdenum-rhenium alloys

    International Nuclear Information System (INIS)

    Fischer, B.; Freund, D.

    2001-01-01

    It is necessary to measure strength and creep behavior to guarantee the safe and reliable usage of refractory alloys at extremely high temperatures. In the literature there is very little information available about the properties of Mo-Re alloys at temperatures higher than 1000 C. A special test facility has been designed and built for stress-rupture testing at very high temperatures (up to 3000 C) of refractory metals and alloys in inert atmospheres. - The stress-rupture strength as well as the creep behavior of molybdenum-rhenium alloys with rhenium contents between 41 and 51 wt.% have been determined at temperatures ranging from 1200 to 2000 C, and rupture times of up to 10 hours using this facility. Previous measurements of stress-rupture strength and creep behavior of pure rhenium have been compared with the measurement results of Mo-Re alloys. - The discussion of the values measured is based on metallographic test results and scanning electron microscopy (SEM) images of Mo-Re alloy samples after stress-rupture testing. (orig.) [de

  9. Influence of Chromium and Molybdenum on the Corrosion of Nickel Based Alloys

    International Nuclear Information System (INIS)

    Hayes, J R; Gray, J; Szmodis, A W; Orme, C A

    2005-01-01

    The addition of chromium and molybdenum to nickel creates alloys with exceptional corrosion resistance in a diverse range of environments. This study examines the complementary roles of Cr and Mo in Ni alloy passivation. Four nickel alloys with varying amounts of chromium and molybdenum were studied in 1 molar salt solutions over a broad pH range. The passive corrosion and breakdown behavior of the alloys suggests that chromium is the primary element influencing general corrosion resistance. The breakdown potential was nearly independent of molybdenum content, while the repassivation potential is strongly dependant on the molybdenum content. This indicates that chromium plays a strong role in maintaining the passivity of the alloy, while molybdenum acts to stabilize the passive film after a localized breakdown event

  10. Study of internal oxidation kinetics of molybdenum base alloys

    International Nuclear Information System (INIS)

    Krushinskij, Yu.Yu.; Belyakov, B.G.; Belomyttsev, M.Yu.

    1989-01-01

    Metallographic and microdurometric method as well as new technique were used to study kinetics of internal oxidation (IO). It is shown that study of IO kinetics on the base of metallographic measurements of layers depth is not correct because it is related with insufficient sensitivity of the method. IO kinetics under conditions of formation of molybdenum oxide layer on saturated material surface as well as IO of alloy with high carbon content were investigated. Oxide film formation does not affect the IO kinetics; decarburization observed along with oxidation increases the apparent activation energy and K exponent on time dependence of diffusion layer depth

  11. Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content

    International Nuclear Information System (INIS)

    Decours, J.; Fabrique, B.; Peault, O.

    1963-01-01

    We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the γ-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The α grain is fine, the γ-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the α-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the α-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [fr

  12. Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

    International Nuclear Information System (INIS)

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

  13. Wetting of molybdenum with molten Cu-O alloys

    International Nuclear Information System (INIS)

    Yupko, V.L.; Garbuz, V.V.; Kryuchkova, N.I.

    1992-01-01

    The Cu-O alloys were prepared from type MOb copper (GOST 859-78) with an oxygen content of 0.001 wt.% and type ChDA cuprous oxide (MRTU 6-09-1451-64), the powder of which was first pressed into briquettes. The weighted portions of Cu 2 O were weighed on an Elektrobalans scale having an absolute error of ±5 · 10 -7 g. The relative error in weighing an approximately 1 · 10 -4 g weighed portion of Cu 2 O for preparation of the alloy with the minimum oxygen content of 0.002% was, therefore, ± 0.5% and consequently for the alloys with a higher oxygen content the accuracy was higher. The alloys were prepared on a ZrO 2 + 5% Y 2 O 3 ceramic at 1,420 K in a vacuum of 6.7 · 10 -3 Pa,d their weight was 1.0-1.5 g, and the melting time 30 sec. The pure type MOb copper was remelted in the same manner. The time relationships of the angle of wetting of molybdenum by molten Cu-O alloys under conditions of combined heating are given. With an increase in oxygen content from 0.004 to 0.005%, wetting drops sharply

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

    International Nuclear Information System (INIS)

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

    2005-08-01

    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 and 573 K, respectively. The creep-fatigue deformation behaviour of OFHC-copper was also investigated but only in the unirradiated condition and at room temperature. The creep-fatigue interaction was simulated by applying a certain holdtime on both tension and compression sides of the cyclic loading 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 modes of cyclic loading. In addition, a number of 'interrupted' creep-fatigue tests were performed on the prime aged CuCuZr specimens in the strain controlled mode with a strain amplitude of 0.5% and a holdtime of 10 seconds. The lifetimes in terms of the number of cycles to failure were determined at different strain and load amplitudes at each holdtime. Post-deformation microstructures was investigated using a transmission electron microscopy. The main results of these investigations are presented and their implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the case of OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction in the yield strength due to overaging heat treatments causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K

  15. On the mechanism of dispersion hardening in molybdenum-carbide alloy systems

    International Nuclear Information System (INIS)

    Shulepov, V.I.; Yudkovskij, S.I.; Batenina, O.I. et al.

    1975-01-01

    The effect of heat treatment of the forming alloys of the Mo-Ti-C and Mo-Ti-Zr-C systems (at the temperatures below the recrystallization temperature) on the structure, distribution of carbon and mechanical properties of the alloys is studied. It is shown that the dispersion-strengthened state of the molybdenum alloys may be obtained on the account of the deformation ageing effect, rather than through the use of the standard heat-treatment procedure (hardening plus ageing). On the basis of the experimental results a theoretical explanation of strengthening of the high-alloy molybdenum-titanum-carbon system is given

  16. A study on direct alloying with molybdenum oxides by feed wire method

    Directory of Open Access Journals (Sweden)

    Jingjing Zou

    2018-04-01

    Full Text Available Direct alloying with molybdenum oxides has been regarded in years; the main addition methods are adding to the bottom of electric arc furnace (EAF with scrap, adding to the ladle during the converter tapping and mixing molybdenum oxide, lime and reductant to prepare pellet added to basic oxygen furnace (BOF. In this paper, a new method for direct alloying with molybdenum trioxide is proposed, adding molybdenum trioxide molten steel by feeding wire method in ladle furnace (LF refining process. The feasibility of molybdenum oxide reduction, the influence rules of bottom-blown on liquid steel fluidity and the yield of molybdenum by feeding wire method were analyzed. Results show that molybdenum oxide can be reduced by [Al], [Si], [C], and even [Fe] in molten steel. Bottom blowing position has a significant influence on the flow of molten steel when the permeable brick is located in 1/2 radius. The yields of Mo are higher than 97% for the experiments with feed wire method, the implementation of direct alloying with molybdenum trioxide by feed wire method works even better than that uses of ferromolybdenum in the traditional process.

  17. Highlighting micrographic structures of uranium alloys containing 0.5 to 10 per cent wt molybdenum

    International Nuclear Information System (INIS)

    Laniesse, J.; Bouleau, M.

    1959-02-01

    The authors report a study which aimed at determining for different uranium molybdenum alloys and with respect to their molybdenum content a polishing method which allows a relatively simple grain examination in the as-cast condition, an as perfect as possible resolution of eutectic decompositions, and the appropriate conditions to highlight structures (beta-alpha and gamma-alpha martensite transformations, beta phase retention and decomposition, transient structures, eutectoid decomposition, and so on). Alloys differ by their molybdenum content: from 0.5 to 1 per cent wt, 1.5 to 3 per cent wt, 5 to 10 per cent wt

  18. Friction and corrosion resistance of sputter deposited supersaturated metastable aluminium-molybdenum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Abu-Zeid, O.A. [Univ. of the United Arab Emirates, Al-Ain (United Arab Emirates). Dept. of Mech. Eng.; Bates, R.I. [Design, Mfg. and Marketing Research Inst., Univ. of Salford (United Kingdom)

    1996-12-15

    Two closed field unbalanced magnetrons with targets of aluminium and molybdenum have been used for the co-deposition of aluminium-molybdenum coatings with different compositions. A pin on disk machine and a computer controlled potentiostat have been used to evaluate respectively, the tribological and corrosion properties of the deposited alloys. Results have shown that introducing molybdenum into aluminium coatings improves their poor tribological properties. Aluminium-molybdenum coatings with different compositions have shown low wear behaviour and for coatings with high molybdenum contents (> 80%) friction coefficients against steel, as low as 0.18 have been obtained. The addition of molybdenum into aluminium coatings has reduced their corrosion tendency and corrosion current density in a marine environment. (orig.)

  19. 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)

  20. 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.

  1. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

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

    2000-01-01

    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

  2. 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

    results of these investigations are presented andtheir implications are briefly discussed in the present report. The central conclusion emerging from the present work is that the application of holdtime generally reduces the number of cycles to failure. The largest reduction was found to be in the caseof...... OFHC-copper. Surprisingly, the magnitude of this reduction is found to be larger at lower levels of strain or stress amplitudes, particularly when the level of the stress amplitude is below the monotonic yield strength of the material. The reduction inthe yield strength due to overaging heat treatments...... causes a substantial decrease in the number of cycles to failure at all holdtimes investigated. The increase in the yield strength due to neutron irradiation at 333 K, on the other hand, causes anincrease in the number of cycles to failure. The irradiation at 573 K to a dose level of 0.2-0.3 dpa does...

  3. Emulation of neutron irradiation effects with protons: validation of principle

    International Nuclear Information System (INIS)

    Was, G.S.; Busby, J.T.; Allen, T.; Kenik, E.A.; Jensson, A.; Bruemmer, S.M.; Gan, J.; Edwards, A.D.; Scott, P.M.; Andreson, P.L.

    2002-01-01

    denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 deg. C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy

  4. Adhesive wear of iron chromium nickel silicon manganese molybdenum niobium alloys with duplex structure

    International Nuclear Information System (INIS)

    Lugscheider, E.; Deppe, E.; Ambroziak, A.; Melzer, A.

    1991-01-01

    Iron nickel chromium manganese silicon and iron chromium nickel manganese silicon molybdenum niobium alloys have a so-called duplex structure in a wide concentration range. This causes an excellent resistance to wear superior in the case of adhesive stress with optimized concentrations of manganese, silicon, molybdenum and niobium. The materials can be used for welded armouring structures wherever cobalt and boron-containing alloy systems are not permissible, e.g. in nuclear science. Within the framework of pre-investigations for manufacturing of filling wire electrodes, cast test pieces were set up with duplex structure, and their wear behavior was examined. (orig.) [de

  5. Effect of temperature on the crack resistance of a molybdenum alloy with 30% tungsten

    International Nuclear Information System (INIS)

    Uskov, E.I.; Babak, A.V.; D'yachkov, A.P.; Platonov, V.A.

    1986-01-01

    Results are presented for a study of the effect of temperature on the crack resistance of molybdenum alloy with 30% tungsten (Mo - 30% W), and data are presented for the crack resistance of commercial-purity molybdenum and tungsten obtained by power metallurgy in the temperature range 20-1800 C. It was found that the nature of failure for Mo-30% W alloy depends on test temperature; in the temperature range 20 C-T /SUP d/ /SUB br/ (upper boundary for the temperature range of the ductile to brittle transition), failure is unstable in nature, and at temperatures exceeding this transition, it occurs by steady main crack development

  6. Neutron irradiation facility and its characteristics

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji

    1995-01-01

    A neutron irradiation facility utilizing spallation reactions with high energy protons is conceived as one of the facilities in 'Proton Engineering center (PEC)' proposed at JAERI. Characteristics of neutron irradiation field of the facility for material irradiation studies are described in terms of material damage parameters, influence of the pulse irradiation, irradiation environments other than neutronics features, etc., comparing with the other sorts of neutron irradiation facilities. Some perspectives for materials irradiation studies using PEC are presented. (author)

  7. Hydriding and neutron irradiation in zircaloy-4

    International Nuclear Information System (INIS)

    Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

    2003-01-01

    The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

  8. A neutron irradiator applied to cancer treatment

    International Nuclear Information System (INIS)

    Campos, Tarcisio P.R.; Andrade, Ana P. de

    2000-01-01

    Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

  9. Investigation into electrochemical behavior of molybdenum VM-1 alloy at high current density

    Energy Technology Data Exchange (ETDEWEB)

    Tatarinova, O M; Amirkhanova, N A; Akhmadiev, A G

    1975-01-01

    The effect of the composition and concentration of electrolyte on the workability of the molybdenum VM-1 alloy has been studied and a number of anions has been determined relative to their activation capacity. The best workability of the alloy is achieved in a 15% NaOH solution and a composite electrolyte 15% NaNO/sub 3/+5%NaOH. It is shown that in polarization of the VM-1 alloy both in alkali- and salt solutions a film of oxides of different valence molybdenum is formed: Mo/sub 2/O/sub 3/, Mo/sub 4/O/sub 11/, Mo/sub 9/O/sub 26/, MoO/sub 3/, but molybdenum gets dissolved only in a hexavalent form, its content in a solution being in conformity with the polarizing current densities. Using a temperature-kinetic technique it has been found that the concentrational polarization is the limiting stage in the reaction of molybdenum and VM-1 alloy anodic dissolution in 15% NaNO/sub 3/ solution and in the composite electrolyte 15%NaNO/sub 3/+5%NaOH.

  10. 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.

  11. Neutron irradiation of seeds 2

    Energy Technology Data Exchange (ETDEWEB)

    1968-10-01

    The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs.

  12. Neutron irradiation of seeds 2

    International Nuclear Information System (INIS)

    1968-01-01

    The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs

  13. Influence of plastic deformation on nitriding of a molybdenum-hafnium alloy

    International Nuclear Information System (INIS)

    Lakhtin, Yu.M.; Kogan, Ya.D.; Shashkov, D.P.; Likhacheva, T.E.

    1982-01-01

    The influence of a preliminary plastic strain on the structure and properties of molybdenum alloy with 0.2 wt.% Hf upon nitriding in the ammonia medium at 900-1200 deg C during 1-6 h is investigated. The study of microhardness distribution across the nitrided layer thickness has shown that with increase of the degree of preliminary plastic strain up to 50 % the nitrided layer hardness decreases and with further reduction growth up to 90 % - increases. Nitriding sharply (hundred times) increases wear resistance of molybdenum alloy with hafnium addition. At the reduction degree 25 % the wear resistance is less than at other values of percentage reduction in area owing to the minimum thickness of the nitride zone. The alloy strained before nitriding by 25 % has shown the best results during heat resistance testing

  14. An in situ study of the enhancement of atomic mobility under neutron irradiation in a Ag-30at%Zn alloy, using the Zener relaxation

    International Nuclear Information System (INIS)

    Halbwachs, M.; Hillairet, J.; Gonzalez, H.; Cost, J.

    1975-01-01

    An application of the Zener relaxation to the study of fast neutron effects on a Ag-Zn alloy is presented. The temperature dependence of the relaxation time is shown. It can be seen that the measured relaxation time at a given temperature increases with the irradiation doses [fr

  15. 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.

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

    International Nuclear Information System (INIS)

    Diaz Barriga Arceo, L.; Orozco, E.; Mendoza-Leon, H.; Palacios Gonzalez, E.; Leyte Guerrero, F.; Garibay Febles, V.

    2007-01-01

    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 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 μm in length were obtained after heating at 800 o C, by means of this process

  17. 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.

  18. The chemistry on a subnanometer scale of radiation-induced precipitation and segregation in fast-neutron irradiated tungsten-rhenium alloys

    International Nuclear Information System (INIS)

    Hershitz, R.; Seidman, D.N.

    1984-01-01

    The phenomena of radiation-induced precipitation and segregation have been investigated in W-10 at.% Re and W-25 at.% Re alloys, employing the atom-probe field-ion-microscope technique. The specimens had been irradiated to a fast-neutron fluence of approx.4x10 22 neutrons cm -2 (e>0.1 MeV) at 575, 625 and 675 deg C. This corresponds to 8.6 dpa and an average displacement rate, for the two year irradiation time of 1.4x10 -7 dpa s -1 . In the W-10 at.% Re alloy, coherent, semicoherent and possibly incoherent precipitates with the composition approx.WRe and a disc-shaped morphology -- one or two atomic planes thick -- were detected at a number density of approx.10 16 cm -3 , and a mean diameter of approx.57 A. In the W-25 at.% Re alloy the same precipitates with the composition approx.WRe 3 were detected at a number density of approx.10 17 cm -3 and a mean diameter of 40 A. The semicoherent WRe 3 precipitates were associated with 4 He atoms; that is, they may have been heterogeneously nucleated. None of the other precipitates were associated with either line or planar defects or with any impurity atoms. Therefore, a true homogeneous radiation-induced precipitation occurs in these alloys. In the W-25 at.% Re alloy a two dimensional WRe 3 phase has been observed at a grain boundary. The nucleation of both precipitates in the vicinity of displacement cascades might be produced by primary knock-on atoms. In both cases, the first step in the nucleation is due to the formation of tightly-bound mobile mixed dumbbells which react to form an immobile di-rhenium cluster. Point-defect mechanisms for all the other observations are also discussed

  19. Influence of iron and beryllium additions on heat resistance of silicide coatings on TsMB-30 molybdenum alloy

    International Nuclear Information System (INIS)

    Zajtseva, A.L.; Fedorchuk, N.M.; Lazarev, Eh.M.; Korotkov, N.A.

    1985-01-01

    Alloying of titanium modified silicide coatings on TsMB-30 molybdenum alloy with iron or beryllium is stated to improve their protective properties. Coatings with low content of alloying elements have the best protective properties. Service life of coatings is determined by the formed oxide film and phase transformations taking place in the coating

  20. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  1. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  2. Recovery of point defects, created by neutron irradiation at 20 K, in ordered and disordered FeCo and FeCo2V alloys

    International Nuclear Information System (INIS)

    Dinhut, J.F.; Riviere, J.P.

    1978-01-01

    Samples of FeCo and FeCo2V ordered and disordered alloys were irradiated by fission neutrons at liquid hydrogen temperature up to an integrated dose of 7 x 10 17 n/cm 2 (E > 1 MeV), and then annealed. During the two first important recovery stages below 200 K, the annealing of about 60% of the radiation induced resistivity occurs. These two steps are respectively assigned to close pair recombination and rearrangement and detrapping of interstitials. During the annealing of the two other stages observed above 200 K, the migration and elimination of interstitials and vacancies involves ordering phenomena. These results and their interpretations are discussed in relation to those of pure metals and to those previously found in ordered and disordered alloys of the same type. (author)

  3. Processing and production of molybdenum and tungsten alloys

    International Nuclear Information System (INIS)

    Hagel, W.C.; Shields, J.A. Jr.; Tuominen, S.M.

    1984-01-01

    The technological means to produce and process Mo and W alloys are summarized because for many Mo and W alloy systems the mechanical properties can be optimized only by thermomechanical processing requiring production and processing capabilities that are not widely available. First, the producers of commercial Mo and W alloys are presented along with currently available product forms. Second, currently disclosed standard capabilities of producers and processors in the United States are presented. 56 references, 13 figures, 9 tables

  4. 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.

  5. Effect of low-temperature thermomechanical treatment on mechanical properties of low-alloying molybdenum alloys with carbide hardening

    International Nuclear Information System (INIS)

    Bernshtejn, L.M.; Zakharov, A.M.; Veller, M.V.

    1978-01-01

    Presented are results of testing low-temperature thermomechanical treatment of low-alloying molybdenum alloys, including quenching from 2100 deg C, 40% deformation by hydroextrusion and aging at the temperature of 1200-1400 deg C. Tensile tests at room temperature with the following processing of results have shown that low-temperature thermomechanical treatment of low-alloying molybdenum alloys of Mo-Zr-C and Mo-Zr-Nb-C systems leads to a significant increase in low-temperature mechanical properties (strength properties - by 30-35%, ductility - by 30-40%) as compared with conventional heat treatment (aging after quenching). The treatment proposed increases resistance to small, as well as large plastic deformations, and leads to a simultaneous rise of strength and plastic properties at all stages of tensile test. Alloying of the Mo-Zr-C system with niobium increases both strength and plastic characteristics as compared with alloys without niobium when testing samples, subjected to low temperature thermomechanical treatment and conventional heat treatment at room temperature

  6. Matrix composition effects on the tensile properties of tungsten-molybdenum heavy alloys

    International Nuclear Information System (INIS)

    Bose, A.; German, R.N.

    1990-01-01

    Tungsten-base heavy alloys are liquid-phase sintered from mixed tungsten, nickel, and iron powders. The sintered product is a composite consisting of interlaced tungsten and solidified matrix (W-Ni-Fe) phases. These alloys are most useful in applications requiring high density, strength, and toughness. The design of improved tungsten heavy alloys has been the subject of several research investigations. Much success has taken place through improved processing, but parallel compositional studies have resulted in new microstructure-property combinations. As part of these investigations, the Ni/Fe ratio has been varied, with the general conclusion that optimal strength and ductility occur with a ratio between 2 and 4. Brittle intermetallic phases can form outside of this composition range. Historically, a 7/3 Ni/Fe ratio has been selected for processing studies. Recently, others reported higher ductilities and impact energies for 90 and 93 pct W heavy alloys with the 8/2 Ni/Fe ratio. Alternatively, these alloys can be strengthened by both solid solution and grain size refinement through incorporation of molybdenum, tantalum, or rhenium. These additions are soluble in both the tungsten and matrix phases and retard solution-reprecipitation during liquid phase sintering. In this study, the alloy composition was varied in the nickel/iron ratio and molybdenum was partially substituted for tungsten. The sintered tensile properties are assessed vs these compositional variations

  7. Influence of neutron irradiation on the stability of recipitates in zircaloy: a critical review

    International Nuclear Information System (INIS)

    Lobo, Raquel M.; Andrade, Arnaldo H. P.

    2013-01-01

    The realization of RMB enterprise (Brazilian Multipurpose Reactor) will give the country a powerful tool to investigate the behavior materials subjected to irradiation. Among them, zirconium alloys, used as cladding of nuclear fuel in reactors type LWR. It is know that neutron irradiation can affect the stability of precipitates in zircaloys, generating as a result changes in theirs mechanical properties, important application of this alloys. This paper present a critical review of neutron irradiation effects on microstructural stability of zircaloys (2 and 4). (author)

  8. 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.

  9. Estimation of the effect of molybdenum on chemical and electrochemical stability of iron-based alloys

    International Nuclear Information System (INIS)

    Tyurin, A.G.

    2003-01-01

    The E-pH diagram for Mo-H 2 O system is made more precise. It is shown that a passivating film on molybdenum in weakly acid, neutral and alkali solutions may constitute MoO 2 only. In strongly acid solutions at anodic polarization the film should transform according to the following scheme: MoO 2 → Mo 4 O 11 → Mo 9 O 26 → MoO 3 . Sections of a Fe-Mo-O system phase diagram and a Fe-Mo-H 2 O system E-pH diagram at 25 deg C are plotted. MoO 2 is found to be a product of iron-molybdenum alloy oxidation in the air and in water. For the system of alloy Kh17N13M2-H 2 O the section of a E-pH diagram is plotted at 25 deg C [ru

  10. Isotopic tailoring with 59Ni to study the effect of helium on microstructural evolution and mechanical properties of neutron-irradiated Fe-Cr-Ni alloys

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Greenwood, L.R.; Stubbins, J.F.; Oliver, B.M.

    1992-03-01

    Tensile testing on three model Fe-Cr-Ni alloys removed from four discharges of the 59 Ni isotopic doping experiment in FFTF-MOTA indicates that helium/dpa ratios typical of fusion reactors do not produce changes in the yield strength or elongation that are significantly different from those at much lower helium generation rates. It also appears that tensile properties approach a saturation level that is dependent only on the final irradiation temperature, but not prior temperature history or thermomechanical starting condition. The saturation in mechanical properties reflects a similar saturation in microstructure that is independent of starting condition. The successful conduct of an isotopic doping experiment was found to require post-irradiation measurement of the helium levels in order to compensate for uncertainties in the cross sections for burn-out and burn-in of 59 Ni and for uncertainties in neutron flux and spectra in the vicinity of the edge of the core

  11. Phase transformations in neutron-irradiated Zircaloys

    International Nuclear Information System (INIS)

    Chung, H.M.

    1986-04-01

    Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after ∼3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr 3 O and cubic-ZrO 2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ∼4 x 10 21 ncm -2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs

  12. Karakteristik kawat TMA (titanium molybdenum alloy) dan penggunaannya dalam perawatan ortodonti

    OpenAIRE

    Arifiani, Putri; Erwin Siregar, Erwin Siregar

    2016-01-01

    Kawat merupakan salah satu piranti yang penting dalam perawatan ortodonsia. Perkembangan terkini dari kawat ortodonsia menghasilkan beberapa jenis kawat dengan karakteristik yang berbeda-beda. Studi pustaka membahas karakteristik kawat ortodonsi beta titanium atau Titanium Molybdenum Alloy (TMA) dan penggunaannya dalam perawatan ortodonsi. Perbedaan karakteristik tiap kawat menjadi hal yang perlu dipertimbangkan secara klinis. Kawat beta titanium atau sering disebut juga dengan kawat TMA (Tit...

  13. The nature of planar faults in a dilute molybdenum-boron alloy

    International Nuclear Information System (INIS)

    Chervinskii, V.I.; Kantor, M.M.; Novikov, I.I.; Sofronova, R.M.

    1982-01-01

    Planar faults on (100) planes in dilute molybdenum-boron alloys consist of a mono- or a bilayer of boron atoms. The displacement vectors are of the general type and for mono- and bilayer faults, respectively, where the component d is close to 1/6 and normal to the fault plane. The planar faults are probably an intermediate stage of MoB or Mo 2 BC growth. (author)

  14. Kinetic parameters of nitridation of molybdenum and niobium alloys with various structure states

    International Nuclear Information System (INIS)

    Solodkin, G.A.; Bulgach, A.A.; Likhacheva, T.E.

    1985-01-01

    Effect of preliminary plastic strain under rolling on kinetic parameters of nitridation of VN-2AEh, VN-3 niobium alloys and molybdenum alloy with hafnium is investigated. Extreme character of dependence of kinetic parameters of nitridation on the degree of reduction under rolling is determined. Preliminary plastic strain at negligible reduction is shown to accelerate growth of the zone of internal nitridation and decelerates growth of the nitride zone. Nitrogen atom removal from the surface to the centre is retarded at the increase of the degree of reduction up to 50% and higher. The degree of deformations is the higher the lower nitridation temperature is

  15. Change in mechanical properties of low-alloyed molybdenum alloys at two-stage strengthening during aging

    International Nuclear Information System (INIS)

    Bernshtejn, L.M.; Zakharov, A.M.; Arbuzov, V.K.

    1977-01-01

    Change in mechanical properties of hardened low-alloyed molybdenum alloys (Mo-Zr-C and Mo-Zr-Nb-C) at two-stage strengthening during ageing at 1400 deg C is studied. The initial strengthening maximum following ageing for 5 hr is caused by separation of dispersed ZrC particles and is accompanied by worsened plasticity, a development characteristic of precipitation hardening processes. The second increase in strength after a 10-hr ageing is not accompanied by reduced plasticity, this being characteristic of strengthening as a result of reconstruction of the dislocation structure. Niobium (0.16 wt.%) added to Mo-Zr-C alloys simultaneously increases their plastic and strength properties. The said effect is caused by prevention of premature decomposition of alloys on account of increased low-temperature plasticity, which permits to obtain high resistance to plastic deformation

  16. Localized corrosion of molybdenum-bearing nickel alloys in chloride solutions

    International Nuclear Information System (INIS)

    Postlethwaite, J.; Scoular, R.J.; Dobbin, M.H.

    1988-01-01

    Electrochemical and immersion tests have been applied to a study of the localized corrosion resistance of two molybdenum-bearing nickel alloys. Alloys C-276 and 6y25, in neutral chloride solutions in the temperature range of 25 to 200 C as part of the container materials evaluation screening tests for the Canadian Nuclear Fuel Waste Management Program. Cyclic polarization studies show that the passivation breakdown potentials move rapidly to more active values with increasing temperatures, indicating a reduced resistance to localized corrosion. The results of immersion tests show that both alloys do suffer crevice corrosion in neutral aerated sodium chloride solutions at elevated temperatures, but that in both cases there is a limiting temperature > 100C, below which, the alloys are not attacked, regardless of the chloride concentration

  17. Study of the oxidation kinetics of the nickel-molybdenum alloy

    International Nuclear Information System (INIS)

    Gouillon, Marie-Josephe

    1974-01-01

    This research thesis reports the study of the oxidation of a nickel-molybdenum alloy in the high-nickel-content part of this alloy. After a bibliographical study on the both metals, the author proposes a physical model based on observed phenomena and based on experimental results. Based on a thermodynamic study, the author compares the stability of the different oxides which may be formed, and reports a prediction of oxides obtained on the alloy during oxidation. Qualitative and quantitative studies have been performed by scanning electron microscopy coupled with electronic microprobe analysis to investigate morphological characteristics on oxidation films. A kinetic study by thermogravimetry shows a decrease of the alloy oxidation rate with respect to that of pure nickel at temperatures lower than 800 degrees C. This result is interpreted by the intervention of two opposed diffusion phenomena which act against each other [fr

  18. Crevice corrosion kinetics of nickel alloys bearing chromium and molybdenum

    International Nuclear Information System (INIS)

    Zadorozne, N.S.; Giordano, C.M.; Rodríguez, M.A.; Carranza, R.M.; Rebak, R.B.

    2012-01-01

    Highlights: ► The crevice corrosion resistance of the tested alloys increased with PREN, which is mainly affected by their Mo content. ► Crevice corrosion kinetics was controlled by ohmic drop only in the more dilute chloride solutions. ► Charge transfer control was observed in concentrated chloride solutions. ► A critical ohmic drop was not necessary for crevice corrosion to occur. ► Ohmic drop was a consequence of the crevice corrosion process in certain conditions. - Abstract: The crevice corrosion kinetics of alloys C-22, C-22HS and HYBRID-BC1 was studied in several chloride solutions at 90 °C. The crevice corrosion resistance of the alloys increased with PREN (Pitting Resistance Equivalent Number), which is mainly affected by the Mo content in the alloys. The crevice corrosion kinetics of the three alloys was analyzed at potentials slightly higher than the repassivation potential. Crevice propagation was controlled by ohmic drop in the more dilute chloride solutions, and by charge transfer in the more concentrated chloride solutions. Ohmic drop was not a necessary condition for crevice corrosion to occur.

  19. Development of silicide coating over molybdenum based refractory alloy and its characterization

    International Nuclear Information System (INIS)

    Chakraborty, S.P.; Banerjee, S.; Sharma, I.G.; Suri, A.K.

    2010-01-01

    Molybdenum based refractory alloys are potential candidate materials for structural applications in high temperature compact nuclear reactors and fusion reactors. However, these alloys being highly susceptible to oxidation in air or oxygen at elevated temperature, undergoes severe losses from highly volatile molybdenum trioxide species. Present investigation, therefore, examines the feasibility of development of silicide type of coating over molybdenum base TZM alloy shape (Mo > 99 wt.%) using pack cementation coating technique. TZM alloy was synthesized in this laboratory from oxide intermediates of MoO 2 , TiO 2 and ZrO 2 in presence of requisite amount of carbon, by alumino-thermic reduction smelting technique. The arc melted and homogenized samples of TZM alloy substrate was then embedded in the chosen and intimately mixed pack composition consisting of inert matrix (Al 2 O 3 ), coating powder (Si) and activator (NH 4 Cl) taken in the judicious proportion. The sealed charge packs contained in an alumina crucible were heated at temperatures of 1000 o C for 8-16 h heating cycle to develop the coating. The coating phase was confirmed to be of made of MoSi 2 by XRD analysis. The morphology of the coating was studied by SEM characterization. It had revealed that the coating was diffusion bonded where Si from coating diffused inward and Mo from TZM substrate diffused outward to form the coating. The coating was found to be resistant to oxidation when tested in air up to 1200 o C. A maximum 100 μm of coating thickness was achieved on each side of the substrate.

  20. Neutron irradiation of bacteriophage λ

    International Nuclear Information System (INIS)

    Bozin, D.; Milosevic, M. . E-mail address of corresponding author: bozinde@vin.bg.ac.yu

    2005-01-01

    Double strand breaks (DSB) are the most dangerous lesions in DNA caused by irradiation, but many other lesions, usually called mutations, have not been clearly identified. These lesions, like DSB, can be the source of serious chromosomal damages and finally - cell death. Growing interest in heavy particles for radiotherapy and radioprotection encourages the search of the molecular basis of their action. In this respect, we chose bacteriophage λ1390 as the model system for the study of consequences of neutron irradiation. This derivative of λ phage possesses an unique ability to reversibly reorganize their genome in response to various selective pressures. The phages were irradiated with 13 Gy of mixed neutrons (7.5 Gy from fast and 5.6 Gy from thermal neutrons) and phages genomes were tested to DSB and mutations. Additionally, the stability of λ capsid proteins were tested. After all tests, we can conclude that, under our conditions, low flux of neutrons does not induce neither DNA strand break or DNA mutation nor the stability of λ capsid proteins. (author)

  1. Hair dosimetry following neutron irradiation.

    Science.gov (United States)

    Lebaron-Jacobs, L; Gaillard-Lecanu, E; Briot, F; Distinguin, S; Boisson, P; Exmelin, L; Racine, Y; Berard, P; Flüry-Herard, A; Miele, A; Fottorino, R

    2007-05-01

    Use of hair as a biological dosimeter of neutron exposure was proposed a few years ago. To date, the (32)S(n,p)(32)P reaction in hair with a threshold of 2.5 MeV is the best choice to determine the fast neutron dose using body activation. This information is essential with regards to the heterogeneity of the neutron transfer to the organism. This is a very important parameter for individual dose reconstruction from the surface to the deeper tissues. This evaluation is essential to the adapted management of irradiated victims by specialized medical staff. Comparison exercises between clinical biochemistry laboratories from French sites (the CEA and COGEMA) and from the IRSN were carried out to validate the measurement of (32)P activity in hair and to improve the techniques used to perform this examination. Hair was placed on a phantom and was irradiated at different doses in the SILENE reactor (Valduc, France). Different parameters were tested: variation of hair type, minimum weight of hair sample, hair wash before measurement, delivery period of results, and different irradiation configurations. The results obtained in these comparison exercises by the different laboratories showed an excellent correlation. This allowed the assessment of a dose-activity relationship and confirmed the feasibility and the interest of (32)P measurement in hair following fast neutron irradiation.

  2. 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.

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

    International Nuclear Information System (INIS)

    Nazir, Rabia; Ahmed, Sohail; Mazhar, Muhammad; Akhtar, Muhammad Javed; Siddique, Muhammad; Khan, Nawazish Ali; Shah, Muhammad Raza; Nadeem, Muhammad

    2013-01-01

    Graphical abstract: - Highlights: • Low temperature pyrolysis of [Fe(bipy) 3 ]Cl 2 and [Mo(bipy)Cl 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, 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) 3 ]Cl 2 , and bipyridinemolybedenum(IV) chloride, [Mo(bipy)Cl 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, 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

  4. High-energy neutron irradiation of superconducting compounds

    International Nuclear Information System (INIS)

    Sweedler, A.R.; Snead, C.L.; Newkirk, L.; Valencia, F.; Geballe, T.H.; Schwall, R.H.; Matthias, B.T.; Corenswit, E.

    1975-01-01

    The effect of high-energy neutron irradiation (E greater than 1 MeV) at ambient reactor temperatures on the superconducting properties of a variety of superconducting compounds is reported. The materials studied include the A-15 compounds Nb 3 Sn, Nb 3 Al, Nb 3 Ga, Nb 3 Ge and V 3 Si, the C-15 Laves phase HfV 2 , the ternary molybdenum sulfide Mo 3 Pb 0 . 5 S 4 and the layered dichalcogenide NbSe 2 . The superconducting transition temperature has been measured for all of the above materials for neutron fluences up to 5 x 10 19 n/cm 2 . The critical current for multifilamentary Nb 3 Sn has also been determined for fields up to 16 T and fluences between 3 x 10 17 n/cm 2 and 1.1 x 10 19 n/cm 2

  5. Mechanical properties of molybdenum alloyed liquid phase-sintered tungsten-based composites

    International Nuclear Information System (INIS)

    Kemp, P.B.; German, R.M.

    1995-01-01

    Tungsten-based composites are fabricated from mixed elemental powders using liquid phase sintering, usually with a nickel-iron matrix. During sintering, the tungsten undergoes grain growth, leading to microstructure coarsening that lowers strength but increases ductility. Often the desire is to increase strength at the sacrifice of ductility, and historically, this has been performed by postsintering deformation. There has been considerable research on alloying to adjust the as-sintered mechanical properties to match those of swaged alloys. Prior reports cover many additions, seemingly including much of the periodic table. Unfortunately, many of the modified alloys proved disappointing, largely due to degraded strength at the tungsten-matrix interface. Of these modified alloys, the molybdenum-containing systems exhibit a promising combination of properties, cost, and processing ease. For example, the 82W-8Mo-7Ni-3Fe alloy gives a yield strength that is 34% higher than the equivalent 90W-7Ni-3Fe alloy (from 535 to 715 MPa) but with a 33% decrease in fracture elongation (from 30 to 20% elongation). This article reports on experiments geared to promoting improved properties in the W-Mo-Ni-Fe alloys. However, unlike the prior research which maintained a constant Ni + Fe content and varied the W:Mo ratio, this study considers the Mo:(Ni + Fe) ratio effect for 82, 90, and 93 wt pct W

  6. Mechanical and wear properties of pre-alloyed molybdenum P/M steels with nickel addition

    Directory of Open Access Journals (Sweden)

    Yamanoglu R.

    2012-01-01

    Full Text Available The aim of this study is to understand the effect of nickel addition on mechanical and wear properties of molybdenum and copper alloyed P/M steel. Specimens with three different nickel contents were pressed under 400 MPa and sintered at 1120ºC for 30 minutes then rapidly cooled. Microstructures and mechanical properties (bending strength, hardness and wear properties of the sintered specimens were investigated in detail. Metallographical investigations showed that the microstructures of consolidated specimens consist of tempered martensite, bainite, retained austenite and pores. It is also reported that the amount of pores varies depending on the nickel concentration of the alloys. Hardness of the alloys increases with increasing nickel content. Specimens containing 2% nickel showed minimum pore quantity and maximum wear resistance. The wear mechanism changed from abrasive wear at low nickel content to adhesive wear at higher nickel content.

  7. Passivation and corrosion behaviours of cobalt and cobalt-chromium-molybdenum alloy

    International Nuclear Information System (INIS)

    Metikos-Hukovic, M.; Babic, R.

    2007-01-01

    Passivation and corrosion behaviour of the cobalt and cobalt-base alloy Co30Cr6Mo was studied in a simulated physiological solution containing chloride and bicarbonate ions and with pH of 6.8. The oxido-reduction processes included solid state transformations occurring at the cobalt/electrolyte interface are interpreted using theories of surface electrochemistry. The dissolution of cobalt is significantly suppressed by alloying it with chromium and molybdenum, since the alloy exhibited 'chromium like' passivity. The structural and protective properties of passive oxide films formed spontaneously at the open circuit potential or during the anodic polarization were studied using electrochemical impedance spectroscopy in the wide frequency range

  8. 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.

  9. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-09-01

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

  10. Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels

    International Nuclear Information System (INIS)

    Lehmann, J.; Decours, J.

    1964-01-01

    Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the γ structure, - cooling rate at the transformation points, - whether or not the intermediate γ → β transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram α + γ; β + γ the effects of the morphology (in particular the two types of α pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the γ structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [fr

  11. 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)

  12. The microstructure and mechanical properties of a welded molybdenum alloy

    International Nuclear Information System (INIS)

    Wadsworth, J.; Morse, G.R.; Chewey, P.M.

    1983-01-01

    Wrought Ti-Zr-Mo (TZM) alloy has been welded using electron beam, laser and tungsten-inert gas welding techniques. The microstructure, tensile properties and fracture surfaces of these welded samples have been examined. Although the welds have been found to be defect free, a disparity in grain size leading to large strength differences exists between the weld and parent metal. Tensile tests have revealed that fusion zone strengths are typical of those expected for the grain size in the weld metal. However, brittle behavior is also always observed, with fracture initiating at grain boundaries and propagating by intergranular and cleavage modes. Auger electron spectroscopy analysis has eliminated oxygen or other interstitial elements as sources of grain boundary embrittlement. It is proposed that brittle behavior is a result of local high strain rates in the weld zone. These local high strain rates arise from the strength difference between the wrought parent metal and the weld metal as a result of the strong grain size dependence of TZM. It is shown that, either by reducing the strain rate of testing or by removing the grain size difference between the parent and weld metals by heat treatment, significant ductility can in fact be achieved in tensile-tested butt-welded TZM. Thus, it is proposed that TZM welds are not inherently brittle as had commonly been believed. (Auth.)

  13. Liquid phase sintering of carbides using a nickel-molybdenum alloy

    International Nuclear Information System (INIS)

    Barranco, J.M.; Warenchak, R.A.

    1987-01-01

    Liquid phase vacuum sintering was used to densify four carbide groups. These were titanium carbide, tungsten carbide, vanadium carbide, and zirconium carbide. The liquid phase consisted of nickel with additions of molybdenum of from 6.25 to 50.0 weight percent at doubling increments. The liquid phase or binder comprised 10, 20, and 40 percent by weight of the pressed powders. The specimens were tested using 3 point bending. Tungsten carbide showed the greatest improvement in bend rupture strength, flexural modulus, fracture energy and hardness using 20 percent binder with lesser amounts of molybdenum (6.25 or 12.5 wt %) added to nickel compared to pure nickel. A refinement in the carbide microstructure and/or a reduction in porosity was seen for both the titanium and tungsten carbides when the alloy binder was used compared to using the nickel alone. Curves depicting the above properties are shown for increasing amounts of molybdenum in nickel for each carbide examined. Loss of binder phase due to evaporation was experienced during heating in vacuum at sintering temperatures. In an effort to reduce porosity, identical specimens were HIP processed at 15 ksi and temperatures averaging 110 C below the sintering g temperature. The tungsten carbide and titanium carbide series containing 80 and 90 weight percent carbide phase respectively showed improvement properties after HIP while properties decreased for most other compositions

  14. Theoretical study of the localization-delocalization transition in amorphous molybdenum-germanium alloys

    International Nuclear Information System (INIS)

    Ding, K.; Andersen, H.C.

    1987-01-01

    Electronic structure calculations were performed for amorphous germanium and amorphous alloys of molybdenum and germanium. The calculations used Harrison's universal linear-combination-of-atomic-orbitals parameters to generate one-electron Hamiltonians for structural configurations obtained from molecular-dynamics simulations. The density of states calculated for a model of a-Ge showed a distinct pseudogap, although with an appreciable density of states at the minimum. The states in the pseudogap are localized. As the concentration of Mo atoms increases, the pseudogap of the density of states is gradually filled up. The density of states at the Fermi energy calculated for our model of the alloys agrees quite well with that experimentally determined by Yoshizumi, Geballe, and co-workers. The localization index for the states at the Fermi energy is a decreasing function of Mo concentration in the range of 2--14 at. % Mo and the localization length is an increasing function of molybdenum concentration. These results are consistent with the experimental observation of an insulator-metal transition at about 10 at. % Mo

  15. 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. Copyright © 2016 Elsevier B.V. All rights reserved.

  16. Characterization of phosphorus segregation in neutron-irradiated Russian pressure vessel steel weld

    International Nuclear Information System (INIS)

    Miller, M.K.; Jayaram, R.; Russell, K.F.

    1995-01-01

    An atom probe field ion microscopy characterization of three Russian pressure vessel steels has been performed. Field ion micrographs of several lath boundaries have indicated that they are decorated with a semicontinuous film of discrete brightly-imaging precipitates that were identified as molybdenum carbonitrides. In addition, extremely high phosphorus levels were measured at the lath boundaries. The phosphorus was found to be confined to an extremely narrow region indicative of monolayer type segregation. The phosphorus coverage determined from the atom probe results of the unirradiated materials agree with predictions based on McLean's equilibrium model of grain boundary segregation. The boundary phosphorus coverage of a neutron-irradiated weld material was significantly higher than in the unirradiated material. Ultrafine darkly-imaging copper- and phosphorus-enriched precipitates were also observed in the matrix of the neutron-irradiated material. (orig.)

  17. Determination of hafnium, molybdenum, and vanadium in niobium and niobium-based alloys by atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Ide, Kunikazu; Kobayashi, Takeshi; Sudo, Emiko.

    1985-01-01

    The analytical procedure is as follows: Weigh 1 g of a sample and put it into a 100 cm 3 PTFE beaker. Add 5 ml of distilled water and 5 ml of hydrofluoric acid, and then heat the solution on a hot plate, adding 3 ml of nitric acid dropwise. Dilute the solution to 100 cm 3 with distilled water. When hafnium is determined, add 2 g of diammonium titanium hexafluoride ((NH 4 ) 2 TiF 6 )) before dilution. Working standard solutions are prepared by adding the stock standard solutions of hafnium, molybdenum, and vanadium into niobium solutions. When hafnium is determined, add 2 g of (NH 4 ) 2 TiF 6 and the alloying elements in amounts corresponding to those in sample solutions into the working standard solutions. The tolerable amounts of hydrofluoric acid were 2.9 M, 2.1 M, and 3.1 M and those of nitric acid were 1.0 M, 1.6 M, and 1.6 M for hafnium, molybdenum, and vanadium, respectively. It was found that (NH 4 ) 2 TiF 6 greatly increased the sensitivity for hafnium determination. Niobium showed minus effect for hafnium and plus effect for molybdenum and vanadium. The atomic absorption of molybdenum and vanadium were not influenced by the presence of 20 % of each alloying element, while the atomic absorption of hafnium was given plus effect by 20 % of zirconium, iron, cobalt, nickel, manganese, chromium or vanadium and minus effect by 20 % tungsten. The analytical values of hafnium, molybdenum, and vanadium in niobium-based alloys by this method showed a good agreement with those by X-ray fluorescence analysis. The lower limits of determination (S/N=2) were 0.05, 0.001, and 0.002 % and the relative standard deviation were 3, 1, and 1.5 % for hafnium, molybdenum, and vanadium, respectively. (author)

  18. Effect of molybdenum and chromium additions on the mechanical properties of Fe3Al-based alloys

    International Nuclear Information System (INIS)

    Sun Yangshan; Xue Feng; Mei Jianping; Yu Xingquan; Zhang Lining

    1995-01-01

    Iron aluminides based on Fe 3 Al offer excellent oxidation and sulfidation resistance, with lower material cost and density than stainless steels. However, their potential use as structural material has been hindered by limited ductility and a sharp drop in strength above 600 C. Recent development efforts have indicated that adequate engineering ductility of 10--20% and tensile yield strength of as high as 500 MPa can be achieved through control of composition and microstructure. These improved tensile properties make Fe 3 Al-based alloys more competitive against conventional austenic and ferritic steels. The improvement of high temperature mechanical properties has been achieved mainly by alloying processes. Molybdenum has been found to be one of the most important alloying elements for strengthening Fe 3 Al-based alloys at high temperatures. However, the RT(room temperature) ductility decreases with the increase of a molybdenum addition. On the other hand, a chromium addition to Fe 3 Al-based alloys is very efficient for improving RT ductility but not beneficial to yield strength at temperatures to 800 C. The purpose of the present paper is to report the effects of combined additions of molybdenum and chromium on mechanical properties at ambient temperature and high temperature of 600 C

  19. Effect of neutron irradiation on vitreous carbon

    International Nuclear Information System (INIS)

    Kurolenkin, E.I.; Virgil'ev, Yu.S.; Chugunova, T.K.

    1989-01-01

    The change in mass (m), volume (V), specific electric resistance (ρ), coefficient of linear thermal expansion (α), dynamic elasticity modulus (E), and limit of bending strength (σ) of vitreous carbon are studied upon neutron irradiation. Samples for study were two forms of vitreous carbon obtained by hardening thermally reactive polymers at 900-1,000 degree K. Phenol-formaldehyde (bakelite lacquer A, Bakelite A) and furfural-phenol-formaldehyde (FM-2) resin were used. They were irradiated in the experimental water - water VVR-M reactor between 360-1,030 degree K. The maximal neutron flux was 1.65·10 21 neut/cm 2 . Neutron irradiation of vitreous carbon led to its shrinkage and accompanied weakening. Shrinkage and weakening of vitreous carbon was decreased with an increase of treatment and irradiation temperatures

  20. A process for electrodeposition of layers of niobium, vanadium, molybdenum or tungsten, or of their alloys

    International Nuclear Information System (INIS)

    Diepers, H.; Schmidt, O.

    1977-01-01

    An improvement is proposed for the process for electrodeposition of layers of niobium, vanadium, molybdenum or tungsten or of their alloys from molten-salt electrolytes (fluorid melts) which is to increase the quality of layers in order to obtain regular thickness and smooth surfaces. According to the invention, a pre-separation is executed on an auxiliary cathode before the (preheated) cathode is immersed. The cathode is only charged for separation after the adjustment of a constant anode potential. It is an advantage that the auxiliary cathode is mechanically and electrically connected with the cathode. As an electrolyte, a mixture of niobium fluorides and a eustetic mixture of potassium fluorides, sodium fluorides and lithium fluorides are particularly suitable for the electrodeposition of miobium. (UWI) [de

  1. Influence of structure on the mechanical properties of low-alloyed molybdenum

    International Nuclear Information System (INIS)

    Ivashchenko, R.K.; Maksimovich, G.G.; Mil'man, Yu.K.; Shchepanskij, Ya.S.; AN Ukrainskoj SSR, Lvov. Fiziko-Mekhanicheskij Inst.)

    1978-01-01

    Studied were the effects of grain size and the presence of cellular dislocation structure on strength and plastic characteristies of TSM-6 low-alloyed molybdenum during short-term tests in the temperature range fron 20 to 1300 deg C. Structure refining results in considerable increase in short-term strength within the termperature range studied. Ressitance of a crystal lattice to dislocation motion and the Ky parameter in the Hall-Petch equation decreases with the temperature increase. Specific elongation (delta) with increasing the test temperature changes in a curve with the maximum for recrystallized metal and in a curve with minimim for metal with previously formed cellular structure. During the short-term tests the properties of metal with a cellular duslocation structure with provision for a low value of cold brittleness temperature are considered preferable in the whole temperature range studied

  2. Studies on the reactive melt infiltration of silicon and silicon-molybdenum alloys in porous carbon

    Science.gov (United States)

    Singh, M.; Behrendt, D. R.

    1992-01-01

    Investigations on the reactive melt infiltration of silicon and silicon-1.7 and 3.2 at percent molybdenum alloys into porous carbon preforms have been carried out by process modeling, differential thermal analysis (DTA) and melt infiltration experiments. These results indicate that the initial pore volume fraction of the porous carbon preform is a critical parameter in determining the final composition of the raction-formed silicon carbide and other residual phases. The pore size of the carbon preform is very detrimental to the exotherm temperatures due to liquid silicon-carbon reactions encountered during the reactive melt infiltration process. A possible mechanism for the liquid silicon-porous (glassy) carbon reaction has been proposed. The composition and microstructure of the reaction-formed silicon carbide has been discussed in terms of carbon preform microstructures, infiltration materials, and temperatures.

  3. Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

    Science.gov (United States)

    Jiao, Z.; Michalicka, J.; Was, G. S.

    2018-04-01

    Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

  4. 57Fe Moessbauer study of amorphous and nanocrystalline Fe73.5Nb3Cu1Si13.5B9 after neutron irradiation

    International Nuclear Information System (INIS)

    Miglierini, M.; Sitek, J.; Szasz, Z.; Vitazek, K.

    1994-01-01

    57 Fe Moessbauer spectroscopy is used to study neutron irradiation induced changes in the short-range order of Fe 73.5 Nb 3 Cu 1 Si 13.5 B 9 alloy. The samples are investigated in both amorphous and nanocrystalline states. Neutron irradiation leads to an increase of the standard deviation of a hyperfine field distribution (HFD), implying rearrangement of the atoms towards disordering. Simultaneously, changes in the average value of the hyperfine field and a net magnetic moment position occur as a consequence of a spin reorientation, atom mixing and microscopic stress centres which are introduced by neutron irradiation. (orig.)

  5. Chemical reactions induced by fast neutron irradiation

    International Nuclear Information System (INIS)

    Katsumura, Y.

    1989-01-01

    Here, several studies on fast neutron irradiation effects carried out at the reactor 'YAYOI' are presented. Some indicate a significant difference in the effect from those by γ-ray irradiation but others do not, and the difference changes from subject to subject which we observed. In general, chemical reactions induced by fast neutron irradiation expand in space and time, and there are many aspects. In the time region just after the deposition of neutron energy in the system, intermediates are formed densely and locally reflecting high LET of fast neutrons and, with time, successive reactions proceed parallel to dissipation of localized energy and to diffusion of the intermediates. Finally the reactions are completed in longer time region. If we pick up the effects which reserve the locality of the initial processes, a significant different effect between in fast neutron radiolysis and in γ-ray radiolysis would be derived. If we observe the products generated after dissipation and diffusion in longer time region, a clear difference would not be observed. Therefore, in order to understand the fast neutron irradiation effects, it is necessary to know the fundamental processes of the reactions induced by radiations. (author)

  6. Basic design of a rotating disk centrifugal atomizer for uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Alzari, Silvio

    2001-01-01

    One of the most used techniques to produce metallic powders is the centrifugal atomization with a rotating disk. This process is employ to fabricate ductile metallic particles of uranium-molybdenum alloys (typically U- 7 % Mo, by weight) for nuclear fuel elements for research and testing reactors. These alloys exhibit a face-centered cubic structure (γ phase) which is stable above 700 C degrees and can be retained at room temperature. The rotating disk centrifugal atomization allows a rapid solidification of spherical metallic droplets of about 40 to 100 μm, considered adequate to manufacture nuclear fuel elements. Besides the thermo-physical properties of both the alloy and the cooling gas, the main parameters of the process are the radius of the disk (R), the diameter of the atomization chamber (D), the disk rotation speed (ω), the liquid volume flow rate (Q) and the superheating of the liquid (ΔT). In this work, they were applied approximate analytical models to estimate the optimal geometrical and operative parameters to obtain spherical metallic powder of U- 7 % Mo alloy. Three physical phenomena were considerate: the liquid metal flow along the surface of the disk, the fragmentation and spheroidization of the droplets and the cooling and solidification of the droplets. The principal results are the more suitable gas is helium; R ≅ 20 mm; D ≥ 1 m; ≅ 20,000 - 50,000 rpm; Q ≅ 4 - 10 cm 3 /s; ΔT ≅ 100 - 200 C degrees. By applying the relevant non-dimensional parameters governing the main physical phenomena, the conclusion is that the more appropriate non-radioactive metal to simulate the atomization of U- 7 % Mo is gold [es

  7. Interfacial microstructure and joining properties of Titanium–Zirconium–Molybdenum alloy joints brazed using Ti–28Ni eutectic brazing alloy

    Energy Technology Data Exchange (ETDEWEB)

    Song, X.G., E-mail: songxg@hitwh.edu.cn [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Shandong Provincial Key Lab of Special Welding Technology, Harbin Institute of Technology at Weihai, Weihai 264209 (China); Tian, X. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Shandong Provincial Key Lab of Special Welding Technology, Harbin Institute of Technology at Weihai, Weihai 264209 (China); Zhao, H.Y. [Shandong Provincial Key Lab of Special Welding Technology, Harbin Institute of Technology at Weihai, Weihai 264209 (China); Si, X.Q.; Han, G.H.; Feng, J.C. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Shandong Provincial Key Lab of Special Welding Technology, Harbin Institute of Technology at Weihai, Weihai 264209 (China)

    2016-01-20

    Vacuum brazing of Titanium–Zirconium–Molybdenum (Abbreviated as TZM) alloy using Ti–28Ni (wt%) eutectic brazing alloy was successfully achieved in this study. Reliable TZM brazed joints were obtained at the temperatures ranging from 1000 °C to 1160 °C for 600 s. The interfacial microstructure of TZM joints was characterized by employing SEM, EDS and XRD. The effects of brazing temperature on interfacial microstructure and joining properties were investigated in details. TZM brazed joints mainly consisted of δ-Ti{sub 2}Ni phase and Ti-based solid solution (Ti(s,s)). The interfacial microstructure of TZM joints was influenced obviously by brazing temperature. Both the thickness of brazing seam and the amount of δ-Ti{sub 2}Ni phase was reduced with the increasing brazing temperature, while the Ti(s,s) layer did not change significantly. The maximum average shear strength of TZM joints reached 107 MPa when brazed at 1080 °C. The presence of δ-Ti{sub 2}Ni intermetallic phase and crack-like structure in joints deteriorated the joining properties, which resulted in the formation of brittle fracture after shear test. In addition, fracture locations were related to the brazing temperature. When the brazing temperature was relatively low, cracks initiated and propagated in the continuous δ-Ti{sub 2}Ni layer. However, the fracture locations preferred to locating at the interface between TZM substrate and brazing seam when brazing temperature exceeded 1080 °C.

  8. High temperature fatigue behaviour of TZM molybdenum alloy under mechanical and thermomechanical cyclic loads

    International Nuclear Information System (INIS)

    Shi, H.J.; Niu, L.S.; Korn, C.; Pluvinage, G.

    2000-01-01

    High temperature isothermal mechanical fatigue and in-phase thermomechanical fatigue (TMF) tests in load control were carried out on a molybdenum-based alloy, one of the best known of the refractory alloys, TZM. The stress-strain response and the cyclic life of the material were measured during the tests. The fatigue lives obtained in the in-phase TMF tests are lower than those obtained in the isothermal mechanical tests at the same load amplitude. It appears that an additional damage is produced by the reaction of mechanical stress cycles and temperature cycles in TMF situation. Ratcheting phenomenon occurred during the tests with an increasing creep rate and it was dependent on temperature and load amplitude. A model of lifetime prediction, based on the Woehler-Miner law, was discussed. Damage coefficients that are functions of the maximum temperature and the variation of temperature are introduced in the model so as to evaluate TMF lives in load control. With this method the lifetime prediction gives results corresponding well to experimental data

  9. Mechanical properties of depleted uranium-2 w/o molybdenum alloy

    International Nuclear Information System (INIS)

    Deel, O.L.; Burian, R.J.

    1979-01-01

    The primary objective of this program is to develop data and techniques for determining the dynamic impact response of radioactive-material shipping-container systems for environmental control and safety overview and assessment. One phase of this program is the dynamic testing of 1/8-, 1/4-, and 1/2-scale models of uranium-shielded truck casks. These linearly scaled models are fabricated from the same materials typically used in full-size prototype casks. In order to analytically evaluate the results of dynamic tests, it is necessary to know the mechanical properties of the materials of construction. Since the properties of cast uranium--molybdenum alloys vary significantly with casting and heat-treating techniques, it is necessary to fully characterize the mechanical properties of the uranium used in the model tests. This report presents the results of these studies. The uranium alloy exhibited a tensile strength equal to or greater than that reported by others. As indicated by the percentage of elongation and reduction in area, the ductility was lower. Comparative data for the other mechanical properties measured were not found in the literature

  10. Gaseous oxygen and hydrogen embrittlements of the uranium-10 weight % molybdenum alloy

    International Nuclear Information System (INIS)

    Corcos, Jean.

    1979-07-01

    The stress corrosion of an Uranium-10 weight % Molybdenum alloy in high purity gaseous oxygen and hydrogen was studied. Tests were performed with fracture-mechanic specimens, fatigue precracked and carried out in tension with a constant sustained load. The experimental procedure enabled to determine the S.C. morphology during the test, and its kinetics. Tests in gaseous oxygen were performed with p02=0.15 MPa from 0 0 C to 100 0 C, and at 20 0 C for p02=0.15, 0.15.10 -2 and 0.15.10 -4 MPa. Two kinetic laws are proposed. Cracking is transgranular with a quasi-clivage type, and occurs on the (1 1 1) planes of the matrix. Tests in gaseous hydrogen were performed with pH2=0.15 MPa from - 50 0 C to + 135 0 C; for all the tests, even those under no exterior load, there is a failure by S.C. and macroscopic hydruration occurs. We propose a kinetic law, which may display that the hydruration phenomenon rules the S.C. propagation. We have performed the identification of the hydride, as well as the study of the precipitation. These phenomena don't occur with pH2=0.15.10 -2 MPa. The embrittlement is thought to be due to a formation-failure cycle of an hydride precipitate at the crack tip [fr

  11. Orientational relationships between phases in the γ→α transformations for uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    Brun, G.

    1966-04-01

    A crystallographic study has been made of the γ → α + γ transformation in the alloy containing 3 per cent by weight of molybdenum using electronic micro-diffraction; it has been possible to establish the orientational relationships governing the germination of the α phase in the γ phase. One finds: (111)γ // (100) α, (112-bar)γ // (010) α, (11-bar 0)γ // (001)α. By choosing a monoclinic lattice containing the same number of atoms as the orthorhombic lattice for defining the γ mother phase, the change in structure has been explained by adding a homogeneous (112-bar)γ [111]γ shearing deformation to a heterogeneous deformation brought about by slipping of the atoms which are not situated at the nodes of this lattice. The identity of the orientation relationships γ/α and γ/α''b and the loss of coherence γ /α as a function of temperature or of time lead to the conclusion that, in the range studied, the γ → α transformation begins with a martensitic process and continues by germination and growth. (author) [fr

  12. Effects of exposure to high-temperature helium containing oxygen on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature

    International Nuclear Information System (INIS)

    Noda, T.; Okada, M.; Watanabe, R.

    1980-01-01

    The effects of exposure to helium containing oxygen of 0.1-115 vpm at 1000 0 C on the mechanical properties of molybdenum and TZM-Mo alloy at room temperature were studied. The stress-relieved molybdenum specimen which was not recrystallized at test temperature showed the ductility after exposure to helium containing oxygen. The recrystallized molybdenum and TZM lost ductility after exposure to helium containing oxygen of 0.1-13 vpm in a few hours. The embrittlement of molybdenum was considered to be due to the grain boundary weakening. Molybdenum to which carbon was added seemed to hinder the grain boundary weakening by the oxygen contamination. Both stress-relieved and recrystallized TZM specimens picked up oxygen linearly with time of exposure to helium. The increase in oxygen content of TZM, which was considered to be caused by the internal oxidation of titanium and zirconium, results in the embrittlement of TZM. (orig.)

  13. Neutron irradiation damage in transition metal carbides

    International Nuclear Information System (INIS)

    Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

    1991-01-01

    Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

  14. Neutron irradiation induced amorphization of silicon carbide

    International Nuclear Information System (INIS)

    Snead, L.L.; Hay, J.C.

    1998-01-01

    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 x 10 25 n/m 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 (-10.8%), elastic modulus as measured using a nanoindentation technique (-45%), hardness as measured by nanoindentation (-45%), and standard Vickers hardness (-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

  15. 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.

  16. Neutron irradiation effects in advanced superconductors

    International Nuclear Information System (INIS)

    Yoshida, H.; Kodaka, H.; Miyata, K.; Hayashi, Y.; Atobe, K.

    1988-01-01

    This paper reports the effects of neutron irradiation on superconducting transitions studied by susceptibility and resistivity measurements for A15 type compounds, Laves-phase compounds and oxide superconductors. For A15 superconductors, the transition temperature (T c ) decreased with increasing neutron fluence and showed large drop started at about 5 x 10 18 n/cm 2 (E > 0.1 MeV). Post-irradiation annealing gave recovery of T c , but the behaviors were different for the materials with different composition and microstructure. The Laves-phase compounds showed less degradation than the A15 superconductors. For oxide superconductors very sensitive transition change was observed, including the radiation-induced superconductivity

  17. A standard fission neutron irradiation facility

    International Nuclear Information System (INIS)

    Sahasrabudhe, S.G.; Chakraborty, P.P.; Iyer, M.R.; Kirthi, K.N.; Soman, S.D.

    1979-01-01

    A fission neutron irradiation facility (FISNIF) has been set up at the thermal column of the CIRUS reactor at BARC. The spectrum and the flux have been measured using threshold detectors. The paper describes the setting up of the facility, measurement and application. A concentric cylinder containing UO 2 powder sealed inside surrounds the irradiation point of a pneumatic sample transfer system located in the thermal column of the reactor. Samples are loaded in a standard aluminium capsule with cadmium lining and transported pneumatically. A sample transfer time of 1 s can be achieved in the facility. Typical applications of the facility for studying activation of iron and sodium in fission neutrons are also discussed. (Auth.)

  18. Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

  19. Transpassive dissolution of alloy 625, chromium, nickel, and molybdenum in high-temperature solutions containing hydrochloric acid and oxygen

    International Nuclear Information System (INIS)

    Kritzer, P.; Boukis, N.; Dinjus, E.

    2000-01-01

    Coupons of nickel, molybdenum, chromium, and the nickel-based Alloy 625 (UNS 06625) were corroded in strongly oxidizing hydrochloric acid (HCl) solutions at 350 C and a pressure (p) of 24 MPa, with reaction times between 0.75 h and 50 h. For Alloy 625, the effect of surface roughness also was investigated. Nickel and molybdenum showed strong material loss after only 5 h of reaction as a result of the instability of the solid oxides formed under experimental conditions. The attack on chromium started at the grain boundaries. At longer reaction times, thick, spalling oxide layers formed on the surface. The attack on Alloy 625 also started at the grain boundaries and at inclusions leading to the formation of small pits. On polished surfaces, the growth of these pits occurred faster than on nonpolished surfaces, but fewer pits grew. Corrosion products formed at the surface consisted of oxygen and chromium. On isolated spots, nickel- and chlorine-containing products also were found

  20. Radiation damage of pixelated photon detector by neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Isamu [KEK, 1-1 Oho Tsukuba 305-0801 (Japan)], E-mail: isamu.nakamura@kek.jp

    2009-10-21

    Radiation Damage of Pixelated Photon Detector by neutron irradiation is reported. MPPC, one of PPD or Geiger-mode APD, developed by Hamamatsu Photonics, is planned to be used in many high energy physics experiments. In such experiments radiation damage is a serious issue. A series of neutron irradiation tests is performed at the Reactor YAYOI of the University of Tokyo. MPPCs were irradiated at the reactor up to 10{sup 12}neutron/cm{sup 2}. In this paper, the effect of neutron irradiation on the basic characteristics of PPD including gain, noise rate, photon detection efficiency is presented.

  1. Influence of neutron irradiation on ferromagnetic metallic glasses

    International Nuclear Information System (INIS)

    Miglierini, M.; Nasu, Saburo; Sitek, J.

    1992-01-01

    Transmission 57 Fe Moessbauer spectroscopy is used to study effects of neutron irradiation on magnetic properties of Fe-based ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transformation from ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. Presence of Ni in the samples reduces the effects of radiation damage. (orig.)

  2. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  3. Radiation damage in stainless steel under varying temperature neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Naoaki [Kyushu Univ., Kasuga, Fukuoka (Japan). Research Inst. for Applied Mechanics

    1998-03-01

    Microstructural evolution of model alloys of 316SS was examined by neutron irradiation at JMTR under cyclic temperature varying condition. In the case of Fe-16Cr-17Ni, formation of interstitial loops and voids are strongly suppressed by varying the temperature from 473K to 673K. By adding Ti as miner element (0.25wt%), however, abnormal accumulation of vacancies (void swelling of 11%dpa at 0.1dpa) was observed. Theoretical analysis standing on the rate theory of defect clustering and simulation irradiation experiments with heavy ions indicates that the vacancy-rich condition which appears temporally during and after changing the temperature from low to high brings these results. It was also shown that only 1 dpa pre-irradiation at low temperature changes swelling behavior at high temperature above several 10 dpa. The understanding of non-steady-state defect processes under temperature varying irradiation is very important to estimate the radiation damage under fusion environment where short-term and long-term temperature variation is expected. (author)

  4. Experimental investigation and thermodynamic modeling of molybdenum and vanadium-containing carbide hardened iron-based alloys

    International Nuclear Information System (INIS)

    Cabrol, E.; Bellot, C.; Lamesle, P.; Delagnes, D.; Povoden-Karadeniz, E.

    2013-01-01

    Highlights: ► Improvement of a carbide selective extraction method. ► Determination of experimental data on the Fe–C–Cr–Mo–V system for carbides above 900 °C: crystallographic structures and compositions of precipitates, matrix composition. ► High molybdenum solubility in FCC carbides. ► Improvement of thermodynamic databases from experimental results. ► Validation of the optimized database with different compositions steels. -- Abstract: A technique for the microstructural study of steels, based on the use of matrix dissolution to collect the very low number density precipitates formed in martensitic steels, has been considerably improved. This technique was applied to two different grades of alloy, characterized by high nickel and cobalt contents and varying chromium, molybdenum and vanadium contents. The technique was implemented at temperatures ranging between 900 °C and 1000 °C, in order to accurately determine experimental data including the crystallographic structure and chemical composition of the carbides, the carbide solvus temperatures, and variations in the chemical composition of the matrix. These experimental investigations reveal that the solubility of molybdenum in FCC carbides can be very high. These results have been compared with the behavior predicted by computational thermodynamics, and used to evaluate and improve the thermodynamic Matcalc steel database. This upgraded database has been validated on three other steels with different chemical compositions, characterized by the same Fe–Cr–Mo–V–C system

  5. Experimental investigation and thermodynamic modeling of molybdenum and vanadium-containing carbide hardened iron-based alloys

    Energy Technology Data Exchange (ETDEWEB)

    Cabrol, E., E-mail: ecabrol@mines-albi.fr [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Aubert and Duval, BP1 F-63770 Les Ancizes (France); Bellot, C. [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Aubert and Duval, BP1 F-63770 Les Ancizes (France); Lamesle, P.; Delagnes, D. [Institut Clément Ader, Mines Albi, Campus Jarlard, F-81013 Albi Cedex 09 (France); Povoden-Karadeniz, E. [Christian Doppler Laboratory for Early Stages of Precipitation, Vienna University of Technology, Favoritenstrasse 9-11, A-1040 Vienna (Austria)

    2013-04-15

    Highlights: ► Improvement of a carbide selective extraction method. ► Determination of experimental data on the Fe–C–Cr–Mo–V system for carbides above 900 °C: crystallographic structures and compositions of precipitates, matrix composition. ► High molybdenum solubility in FCC carbides. ► Improvement of thermodynamic databases from experimental results. ► Validation of the optimized database with different compositions steels. -- Abstract: A technique for the microstructural study of steels, based on the use of matrix dissolution to collect the very low number density precipitates formed in martensitic steels, has been considerably improved. This technique was applied to two different grades of alloy, characterized by high nickel and cobalt contents and varying chromium, molybdenum and vanadium contents. The technique was implemented at temperatures ranging between 900 °C and 1000 °C, in order to accurately determine experimental data including the crystallographic structure and chemical composition of the carbides, the carbide solvus temperatures, and variations in the chemical composition of the matrix. These experimental investigations reveal that the solubility of molybdenum in FCC carbides can be very high. These results have been compared with the behavior predicted by computational thermodynamics, and used to evaluate and improve the thermodynamic Matcalc steel database. This upgraded database has been validated on three other steels with different chemical compositions, characterized by the same Fe–Cr–Mo–V–C system.

  6. Induced defects in neutron irradiated GaN single crystals

    International Nuclear Information System (INIS)

    Park, I. W.; Koh, E. K.; Kim, Y. M.; Choh, S. H.; Park, S. S.; Kim, B. G.; Sohn, J. M.

    2005-01-01

    The local structure of defects in undoped, Si-doped, and neutron irradiated free standing GaN bulk crystals, grown by hydride vapor phase epitaxy, has been investigated by employing Raman scattering and cathodoluminescence. The GaN samples were irradiated to a dose of 2 x 10 17 neutrons in an atomic reactor at Korea Atomic Energy Research Institute. There was no appreciable change in the Raman spectra for undoped GaN samples before and after neutron irradiation. However, a forbidden transition, A 1 (TO) mode, appeared for a neutron irradiated Si-doped GaN crystal. Cathodoluminescence spectrum for the neutron irradiated Si-doped GaN crystal became much more broadened than that for the unirradiated one. The experimental results reveal the generation of defects with locally deformed structure in the wurtzite Si-doped GaN single crystal

  7. 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

  8. F-type centers in neutron-irradiated AIN

    International Nuclear Information System (INIS)

    Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

    1990-01-01

    The production of point defects by neutron irradiation and thermal decay in sintered AIN polycrystal are investigated. The absorption band at 370 nm is observed after reactor neutron irradiation to a dose of 10 16 n/cm 2 (E > 0.1 MeV). The defect corresponding to the band is tentatively assigned as an F-type center from the optical absorption and electron spin resonance. (author)

  9. Simulation analysis of radiation fields inside phantoms for neutron irradiation

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.

    2007-01-01

    Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)

  10. Strengthening of the brazed joint for single-crystalline molybdenum by using Mo-40%Ru-B alloys

    Energy Technology Data Exchange (ETDEWEB)

    Hiraoka, Y. [Okayama Univ. of Science (Japan). Department of Applied Physics; Igarashi, T. [Tokyo Tungsten Co. Ltd., Toyama (Japan). Research and Development Division

    1998-12-01

    In this study, the bend properties of the single-crystalline molybdenum brazed by using Mo-40%Ru alloys containing boron of 1-6 mass%Ru alloy for the improvement of the joint strength was determined. (orig.) [Deutsch] Durchgefuehrt wurde die Herstellung von Verbindungen aus einkristallinem Molybdaen. Hierbei kamen Mo-40%Ru-Legierungen mit 1 bis 6 Gew.-% Bor als Lotmaterialien zum Einsatz. Festigkeit und Duktilitaet der Verbindungen wurden mittels 3-Punkt-Biegepruefung bei Raumtemperatur und unter fluessigem Stickstoff ermittelt. Die Bruchflaechen der Proben wurden mit Hilfe eines Rasterelektronenmikroskopes untersucht. Die Ergebnisse lassen sich wie folgt zusammenfassen: Der optimale Borgehalt bezueglich Festigkeit und Duktilitaet der geloeteten Verbindung liegt bei 2 Gew.-%. Die entsprechende Probe hat bei einem Biegewinkel von 100 bei Raumtemperatur nicht versagt. Auch unter fluessigem Stickstoff zeigte diese Probe eine Festigkeit in der Groessenordnung des einkristallinen Vollmaterials. (orig.)

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

    International Nuclear Information System (INIS)

    James, L.A.

    1977-01-01

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

  12. Uses of AES and RGA to study neutron-irradiation-enhanced segregation to internal surfaces

    International Nuclear Information System (INIS)

    Gessel, G.R.; White, C.L.

    1980-01-01

    The high flux of point defects to sinks during neutron irradiation can result in segregation of impurity or alloy additions to metals. Such segregants can be preexisting or produced by neutron-induced transmutations. This segregation is known to strongly influence swelling and mechanical properties. Over a period of years, facilities have been developed at ORNL incorporating AES and RGA to examine irradiated materials. Capabilities of this system include in situ tensile fracture at elevated temperatures under ultrahigh vacuum 10 -10 torr and helium release monitoring. AES and normal incidence inert ion sputtering are exploited to examine segregation at the fracture surface and chemical gradients near the surface

  13. A neutron irradiator to perform nuclear activation

    International Nuclear Information System (INIS)

    Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

    2001-01-01

    The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

  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. Microstructure and ductility of Fe28Cr16Co alloy with additions of silicon, molybdenum, titanium and aluminium

    International Nuclear Information System (INIS)

    Vodopivec, F.; Zvokelj, J.; Breskvar, B.; Gnidovec, D.; Rodic, A.; Torkar, M.

    1994-01-01

    The microstructure of several alloys with base composition Fe28Cr16Co and addition up to 1.5% silicon, 0.32% titanium, 2.34% molybdenum and 1% aluminium was investigated in the temperature range 500 to 1250 C by optical microscopy, hardness measurements and dilatometry. Also, ductility and wire drawing tests were carried out on some alloys. The addition of silicon, titanium, molybdenum or 0.13% Al does not prevent the formation of γ phase up to the temperature 1250 C and the formation of phase σ in the temperature range 700 to approximately 1000 C. The addition of 1% Al prevents the formation of phase σ and shifts the temperature of formation of phase γ to 1158 C. The addition of different elements does not affect significantly the spinodal decomposition of phase α. At increased temperature an interval of sufficient ductility for deformation by wire drawing was established. The ductility was greatly improved if the microstructure consisted of small inclusions of phase γ in a matrix of phase α, probably because the deformation by twinning was hindered. However, insufficient magnetic properties were obtained also after 80% of deformation. (orig.)

  16. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    Oliveira, Fabio Branco Vaz de

    2008-01-01

    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, time and

  17. Analytic chemistry of molybdenum

    International Nuclear Information System (INIS)

    Parker, G.A.

    1983-01-01

    Electrochemical, colorimetric, gravimetric, spectroscopic, and radiochemical methods for the determination of molybdenum are summarized in this book. Some laboratory procedures are described in detail while literature citations are given for others. The reader is also referred to older comprehensive reviews of the analytical chemistry of molybdenum. Contents, abridged: Gravimetric methods. Titrimetric methods. Colorimetric methods. X-ray fluorescence. Voltammetry. Catalytic methods. Molybdenum in non-ferrous alloys. Molydbenum compounds

  18. Changes in grain boundary composition induced by neutron irradiation of austenitic stainless steels

    International Nuclear Information System (INIS)

    Asano, K.; Nakata, K.; Fukuya, K.; Kodama, M.

    1992-01-01

    The radiation induced segregation of solutes to the grain boundary in austenitic stainless steels were studied. Type 304 and type 316 steel samples neutron irradiated at 561K up to 9.2x10 25 n/m 2 were obtained and minute compositional profiles across grain boundaries were examined using an analytical scanning transmission electron microscope equipped with a field emission electron gun. Chromium was slightly enriched at grain boundaries at the lowest irradiation dose but decreased with increasing fluence. Higher fluence irradiation resulted in depletion in chromium and molybdenum, and enrichment in nickel, silicon and phosphorus. These changes in grain boundary chemistry were limited within about 5nm of the boundary. Significant depletion of chromium and enrichment of impurities on the grain boundary occurred at fluences roughly coincidental with that of SCC susceptibility change obtained from another project

  19. Neutron irradiation of rat embryos in utero

    International Nuclear Information System (INIS)

    Vogel, H.H. Jr.

    1978-01-01

    In the rat radiation is most effective in producing congenital anomalies during the organ-forming period (days 9 to 13), which is approximately equivalent to the 14th to 50th days of human pregnancy. We have exposed female Sprague--Dawley rats on the 18th day of pregnancy to single whole-body doses of fission neutrons (20 to 150 rads). After 20 rads there was a small decrease in body weight which lasted from birth to weaning. During this period 9% of the irradiated rats died compared with 4% of the controls. After 50 rads, 65/275 (23.6%) of the rats died between birth and weaning, and the body-weight loss of the survivors was increased. After 100 rads, 62/133 (47%) died at birth or day 1 and 103/133 (77.4%) died before weaning. A large and significant decrease in body weight persisted in the survivors. After 150 rads of fission neutrons, all 95 rats died within 48 hr of birth. From cross-fostering experiments, we believe this is a direct effect of radiation on the embryos and not an indirect action through the mother or her milk. The LD 50 for the period from birth to weaning is approximately 75 rads of fission neutrons. Studies of organ weight were conducted daily for the first week after birth in an attempt to find the cause of radiation mortality. Body weight of the irradiated animals averaged only about one-half that of the controls. The liver, kidney, brain, and testes of the neutron-irradiated rats weighed significantly less than those of the controls. The weights of the spleen, lungs, duodenum, and stomach were decreased but not significantly. The bone marrow appeared depleted in the irradiated long bones, but the spleen maintained active hematopoiesis 1 to 2 months after neutron exposure

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-15

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

  1. Effect of neutron irradiation on mechanical properties of ferritic steels

    International Nuclear Information System (INIS)

    Kass, S.B.; Murty, K.L.

    1995-01-01

    Effect of neutron radiation exposure was investigated in various ferritic steels with the main emphasis being the effects of thermal neutrons on radiation hardening. Pure iron of varied grain sizes was also used for characterizing the grain size effects on the source hardening before and after neutron irradiation. While many steels are considered in the overall study, the results on 1020, A516 and A588 steels are emphasized. Radiation hardening due to fast neutrons was seen to be sensitive to the composition of the steels with A354 being the least resistant and A490 the least sensitive. Majority of the radiation hardening stems from friction hardening, and source hardening term decreased with exposure to neutron radiation apparently due to the interaction of interstitial impurities with radiation produced defects. Inclusion of thermal neutrons along with fast resulted in further decrease in the source hardening with a slight increase in the friction hardening which revealed a critical grain size below which exposure to total (fast and thermal) neutron spectrum resulted in a slight reduction in the yield stress compared to the exposure to only fast neutrons. This is the first time such a grain size effect is reported and this is shown to be consistent with known radiation effects on friction and source hardening terms along with the observation that low energy neutrons have a nonnegligible effect on the mechanical properties of steels. In ferritic steels, however, despite their small grain size, exposure to total neutron spectrum yielded higher strengths than exposure to only fast neutrons. This behavior is consistent with the fact that the source hardening is small in these alloys and radiation effect is due only to friction stress

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

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

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

  3. Study of supersaturation of defects under neutron irradiation by Zener relaxation

    International Nuclear Information System (INIS)

    Gonzalez, Hector C.; Justus, Francisco J.W.

    2004-01-01

    Vacancy supersaturation in dynamic equilibrium under fast neutron irradiation could be determined by anelastic relaxation. This phenomenon is particularly noticeable in some substitutional binary alloys. Relaxation is due to the reordering of atoms pairs under a stress, being a local reordering at the atomic scale. Relaxation time (τ) is inversely proportional to the vacancy concentration (Cv) and decreases under irradiation because a dynamical equilibrium of vacancy concentration, higher than thermodynamic equilibrium, is established. Theoretical models allow estimating the magnitude of that supersaturation. Determinations of τ at different temperatures, with and without fast neutron irradiations, were made with an 'in situ' device placed in the high temperature loop in the RA1 CAC-CNEA reactor. An alloy Au-30% Ni was used, since it presents an appreciable Zener effect. The measurements were performed in a spring-shaped specimen in order to minimize temperature and flux gradients. An Arrhenius plot of τ was obtained, and it was observed that for temperatures lower than 220 C degrees a vacancy supersaturation exists. The lowest temperature of our experiments was 190 C degrees. A value of τ at this temperature was three times lower under irradiations. A plot of τ vs. fast neutron fluence (φ f t) at the irradiation temperature T= 203 C degrees was obtained. An increase of τ was observed. After an annealing at T = 280 C degrees, the value of τ recovers the value corresponding to the unirradiated case. This fact suggests that the loops produced by irradiation act as defect sinks. (author) [es

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

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    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.

  5. 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...

  6. 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

  7. 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

  8. 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

  9. Mechanical properties of molybdenum-titanium alloys micro-structurally controlled by multi-step internal nitriding

    International Nuclear Information System (INIS)

    Nagae, M.; Yoshio, T.; Takemoto, Y.; Takada, J.; Hiraoka, Y.

    2001-01-01

    Internally nitrided dilute Mo-Ti alloys having a heavily deformed microstructure near the specimen surface were prepared by a novel two-step nitriding process at 1173 to 1773 K in N 2 gas. For the nitrided specimens three-point bend tests were performed at temperatures from 77 to 298 K in order to investigate the effect of microstructure control by internal nitriding on the ductile-to-brittle transition temperature (DBTT) of the alloy Yield strength obtained at 243 K of the specimen maintaining the deformed microstructure by the two-step nitriding was about 1.7 times as much as recrystallized specimen. The specimen subjected to the two-step nitriding was bent more than 90 degree at 243 K, whereas recrystallized specimen was fractured after showing a slight ductility at 243 K. DBTT of the specimen subjected to the two-step nitriding and recrystallized specimen was about 153 K and 203 K, respectively. These results indicate that multi-step internal nitriding is very effective to the improvement in the embrittlement by the recrystallization of molybdenum alloys. (author)

  10. 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)

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

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-01-01

    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 60 Co;7.4 MBq/day

  12. Studying titanium-molybdenum-zirconium alloys of increased corrosion resistance in acid solutions

    International Nuclear Information System (INIS)

    Tomashov, N.D.; Kazarin, V.I.; Mikheev, V.S.; Goncharenko, B.A.; Sigalovskaya, T.M.; Kalyanova, M.P.

    1977-01-01

    New promising Ti-Mo-Nb-Zr system alloys, possessing good workability and a high corrosion resistance in non-oxidizing solutions of acids, have been developed. The alloys may be recommended as structural materials for equipment operating in severely agressive acid media, such as hydrochloric, sulphuric and phosphoric acids. The corrosion resistance of alloys of the above system in solutions of H 2 SO 4 , HCl and H 3 PO 4 acids may be maximized by increasing the overall alloying to 42% (keeping the ratio of the alloying components Mo/Nb/Zr=4/1/1 unchanged), while retaining sufficiently good plasticity and workability

  13. A comparison of corrosion resistance of cobalt-chromium-molybdenum metal ceramic alloy fabricated with selective laser melting and traditional processing.

    Science.gov (United States)

    Zeng, Li; Xiang, Nan; Wei, Bin

    2014-11-01

    A cobalt-chromium-molybdenum alloy fabricated by selective laser melting is a promising material; however, there are concerns about the change in its corrosion behavior. The purpose of this study was to evaluate the changes in corrosion behavior of a cobalt-chromium-molybdenum alloy fabricated by the selective laser melting technique before and after ceramic firing, with traditional processing of cobalt-chromium-molybdenum alloy serving as a control. Two groups of specimens were designated as group selective laser melting and group traditional. For each group, 20 specimens with a cylindrical shape were prepared and divided into 4 cells: selective laser melting as-cast, selective laser melting fired in pH 5.0 and 2.5, traditional as-cast, and traditional fired in pH 5.0 and 2.5. Specimens were prepared with a selective laser melting system for a selective laser melting alloy and the conventional lost wax technique for traditional cast alloy. After all specimen surfaces had been wet ground with silicon carbide paper (1200 grit), each group of 10 specimens was put through a series of ceramic firing cycles. Microstructure, Vickers microhardness, surface composition, oxide film thickness, and corrosion behavior were examined for specimens before and after ceramic firing. Three-way ANOVA was used to evaluate the effect of porcelain firing and pH values on the corrosion behavior of the 2 alloys (α=.05). Student t tests were used to compare the Vickers hardness. Although porcelain firing changed the microstructure, microhardness, and x-ray photoelectron spectroscopy results, it showed no significant influence on the corrosion behavior of the selective laser melting alloy and traditional cast alloy (P>.05). No statistically significant influence was found on the corrosion behavior of the 2 alloys in different pH value solutions (P>.05). The porcelain firing process had no significant influence on the corrosion resistance results of the 2 alloys. Compared with traditional

  14. Neutron irradiation effects in pressure vessel steels and weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  15. Behavior of fluorine 18 in neutron irradiated zeolites

    International Nuclear Information System (INIS)

    Estevez Lopez, D.R.

    1992-01-01

    The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

  16. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  17. The effects of oxide evolution on mechanical properties in proton- and neutron-irradiated Fe-9%Cr ODS steel

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, M.J., E-mail: matthewswenson1@u.boisestate.edu [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Dolph, C.K. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Wharry, J.P. [Boise State University, 1910 University Drive, Boise, ID 83725 (United States); Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

    2016-10-15

    The objective of this study is to evaluate the effect of irradiation on the strengthening mechanisms of a model Fe-9%Cr oxide dispersion strengthened steel. The alloy was irradiated with protons or neutrons to a dose of 3 displacements per atoms at 500 °C. Nanoindentation was used to measure strengthening due to irradiation, with neutron irradiation causing a greater increase in yield strength than proton irradiation. The irradiated microstructures were characterized using transmission electron microscopy and atom probe tomography (APT). Cluster analysis reveals solute migration from the Y-Ti-O-rich nanoclusters to the surrounding matrix after both irradiations, though the effect is more pronounced in the neutron-irradiated specimen. Because the dissolved oxygen atoms occupy interstitial sites in the iron matrix, they contribute significantly to solid solution strengthening. The dispersed barrier hardening model relates microstructure evolution to the change in yield strength, but is only accurate if solid solution contributions to strengthening are considered simultaneously.

  18. Damage production by fast electrons in dilute alloys of vanadium, niobium and molybdenum

    International Nuclear Information System (INIS)

    Jung, P.

    1975-01-01

    Vanadium, niobium and molybdenum samples containing about 300 ppm of zirconium were irradiated at helium temperature with electrons of energies between 0,6 and 3.1 MeV. The measured damage rates were analysed in terms of minimum threshold energy, damage function and resistivity per unit concentration of Frenkel pairs. For the minimum threshold energy T(Sub)d, values of 25+-2 eV (V) 28+-2 e V(Nb) and 34+-2 e V(Mo) were obtained. Pronounced differences between the displacement functions of molybdenum and that of niobium and vanadium are found which are explained by different stability of the defects during the irradiation at helium temperature

  19. Boron Steel: An Alternative for Costlier Nickel and Molybdenum Alloyed Steel for Transmission Gears

    Directory of Open Access Journals (Sweden)

    A. Verma

    2010-06-01

    Full Text Available Case Carburized (CC low carbon steels containing Ni, Cr and Mo alloying elements are widely used for transmission gears in automobile, as it possesses desired mechanical properties. In order to cut cost and save scarce materials like Ni and Mo for strategic applications, steel alloyed with Boron has been developed, which gives properties comparable to Ni-Cr-Mo alloyed steel. In the process of steel development, care was taken to ensure precipitation of boron which results in precipitation hardening. The characterization of the developed boron steel had exhibited properties comparable to Ni-Cr-Mo alloyed steel and superior to conventional boron steel.

  20. In-situ radiation response of nickel--molybdenum alloys in an HVEM

    International Nuclear Information System (INIS)

    Carpenter, R.W.; Kenik, E.A.

    1977-01-01

    Experimental research has shown that perturbations of lattice periodicity caused by certain types of discrete precipitation or spinodal decomposition will suppress radiation-induced void swelling. The reported results extend this study to the Ni--Mo alloy system, in which certain alloy compositions contain short- or long-range order (SRO or LRO) and consequently modulations of the lattice periodicity

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

    The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    Highlights: • Titanium alloy β21S was coated with Mo doped DLC. • XRD, XPS and micro Raman show that Mo is present in the form of carbide. • Mo doping facilitates apatite growth on DLC during immersion in Hanks’ solution. • Mo doped DLC sample shows better passivation behavior in Hanks’ solution. - Abstract: 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Anandan, C., E-mail: canandan@nal.res.in; Mohan, L.; Babu, P. Dilli

    2014-03-01

    Highlights: • Titanium alloy β21S was coated with Mo doped DLC. • XRD, XPS and micro Raman show that Mo is present in the form of carbide. • Mo doping facilitates apatite growth on DLC during immersion in Hanks’ solution. • Mo doped DLC sample shows better passivation behavior in Hanks’ solution. - Abstract: 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.

  5. 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.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  7. Review of recent studies on neutron irradiation embrittlement in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sudo, Akira; Miyazono, Shohachiro

    1983-06-01

    Recent studies in foreign countries (USA, France, FRG and UK) on neutron irradiation embrittlement have been reviewed. These studies are classified into four areas, such as 1) effect of chemical composition on irradiation embrittlement sensitivity, 2) postirradiation heat treatment for embrittlement relief, 3) fracture toughness evaluation of irradiated materials based on fracture mechanics analysis, and 4) effect of irradiation on fatigue crack propagation behavior. The first area mainly includes the studies related to the effects of copper, phosphorus impurities and nickel alloying and synergistic effect of these components, and furthermore, evaluation of Regulatory Guide 1.99 Rev.l. Studies in the second area show the effects of annealing condition (temperature and time) and metallugical condition on embrittlement relief, and evaluation of periodic annealing in the period of irradiation as a promising method for embrittlement control. Studies in the third area show the correlation between fracture toughness and Cv notch ductility changes with neutron irradiation, and J-R curves of irradiated materials based on the elasto-plastic fracture mechanics. In the forth area, most of studies are investigated in air condition but a few studies in reactor-grade water at high temperature and pressure. (author)

  8. Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

    Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

  9. 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....... Specimens exposed to periodic temperature variations experienced a low temperature (360 °C) during the initial 10% of accrued dose in each of the eight cycles, and a higher temperature (520 °C) during the remaining 90% of accrued dose in each cycle. The microstructures of the irradiated stainless steel...

  10. Neutron irradiation effects on intermetallic precipitates in Zircaloy as a function of fluence

    International Nuclear Information System (INIS)

    Etoh, Y.; Shimada, S.

    1993-01-01

    Intermetallic precipitates in Zircaloy-2 and -4, recrystallized at the α-phase temperature, have been examined using analytical electron microscopy. The specimens were irradiated in BWRs up to a fast neutron fluence of 1.4x10 26 n/m 2 (E>1 MeV). Neutron irradiation induces a crystalline-to-amorphous transition, depleting Fe in the amorphous phase of Zr(Fe, Cr) 2 precipitates in the alloys. Amorphization starts from the periphery of the precipitates and all of them are totally amorphized at higher fluences than 1.2x10 26 n/m 2 . The width of the Fe-depleted zone increases in proportion to the 0.45 power of fluence. This result indicates that diffusion of Fe is the rate-controlling process for Fe depletion in Zr(Fe, Cr) 2 precipitates. Dissolution of Zr 2 (Fe, Ni) precipitates in Zircaloy-2 occurs during neutron irradiation. At a high fluence, such as 1.2x10 26 n/m 2 , Zr 2 (Fe, Ni) precipitates are almost completely dissolved into the matrix and the dissolution rate of Fe is faster than that of Ni. (orig.)

  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. Relationship between swelling and elastic properties in neutron-irradiated 316 stainless steel

    International Nuclear Information System (INIS)

    Bates, J.F.

    1976-04-01

    The results encompass elastic property measurements on several alloys, which differ in silicon, molybdenum and phosphorus contents but have a nominal 316 stainless steel composition. It is shown that there is a good correlation between the initial shear modulus of the material and the resultant swelling rate of that material. It is also shown that the bias factor concept does not satisfactorily account for the observed compositional sensitivity of swelling in the alloys investigated. 6 fig

  13. Fracture mechanism of a dispersion-hardened molybdenum alloy with strong structural interfaces

    International Nuclear Information System (INIS)

    Vasil'ev, A.D.; Malashenko, I.S.; Moiseev, V.F.; Pechkovskij, Eh.P.; Sul'zhenko, V.K.; Trefilov, V.I.; AN Ukrainskoj SSR, Kiev. Inst. Ehlektrosvarki)

    1978-01-01

    Fracture mechanism in the two-phase Mo-15wt.%Nb-3.5 vol.% TiN alloy known to be of ''brittle matrix-strong interfaces'' type has been investigated depending on tensile test temperature. Several temperature intervals of fracture have been found, each of them having its own peculiarities. A scheme is suggested for fracture mechanism changes in dispersion-hardened alloys with strong interfaces. At low test temperatures brittle cleavage fracture takes place. With temperature increase fracture mechanisms change in the following way: brittle intergranular fracture; fracture of ''microvoid coalescence'' type; fracture typical for reinforced materials with ductile matrix; intergran laru fracture. Particles of strengthening phase have been shown to play different roles depending on the test temperature in the fracture of the alloys studied

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

    International Nuclear Information System (INIS)

    Gdowski, G.E.

    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 2 S and CO 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

  15. High dose effects in neutron irradiated face-centered cubic metals

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors

  16. Structure of compensating centers in neutron irradiated n-type germanium

    International Nuclear Information System (INIS)

    Erchak, D.P.; Kosobutskij, V.S.; Stel'makh, V.F.

    1989-01-01

    Structural model of one of the main compensating defects of Ge-M1, Ge-M5, Ge-M6 in neutron irradiated (10 18 -10 20 cm -2 ) germanium, strongly alloyed (2x10 18 -3x10 19 cm -3 ) with antimony, phosphorus and arsenic respectively, is suggested. The above mentioned compensating centers are paramagnetic in a positive charge state and represent a vacancy, two nearby germanium atoms of which are replaced with two atoms of corresponding fine donor impurity. It is mainly contributed (63%- for Ge-M5 centers, 56% - for Ge-M6 centers) by orbitals of two germanium atoms neighbouring the vacancy. The angle of the bonds of each of two mentioned germanium atoms with its three neighbours and orientation of maximum electron density of hybride orbital, binding both germanium atoms, is approximately by 5 deg greater the tetrahedral one

  17. 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)

  18. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  19. Pressure vessel code construction capabilities for a nickel-chromium-tungsten-molybdenum alloy

    International Nuclear Information System (INIS)

    Rothman, M.F.

    1990-01-01

    HAYNES alloy 230 (UNS NO6230) has achieved wide usage in a variety of high-temperature aerospace, chemical process industry and industrial heating applications since its introduction in 1981. Combining high elevated temperature strength with excellent metallurgical stability, environment-resistance and relatively straight forward fabrication characteristics, this Ni-Cr-W-Mo alloy was an excellent candidate for ASME Pressure vessel Code applications. Coverage under case No. 2063 was granted in July, 1989, for both Section I and Section VIII Division 1 construction. In this paper, the metallurgy of 230 alloy will be described, and its design strength capabilities contrasted with those for more established code materials. Other important performance capabilities, such as long-term thermal stability, oxidation-resistance, fatigue-resistance, and resistance to other forms of environmental degradation will be discussed. It will be shown that the combined properties of 230 alloy offer some significant advantages over other materials for applications such as expansion bellows, heat-exchangers, valves and other components in the fossil energy, nuclear energy and chemical process industries, among others

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

    International Nuclear Information System (INIS)

    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)

  1. Study of boron carbide evolution under neutron irradiation

    International Nuclear Information System (INIS)

    Simeone, D.

    1999-01-01

    Owing to its high neutron efficiency, boron carbide (B 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 10 B(n,α) 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 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 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 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 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 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 4 C. (author)

  2. Separation of Protactinium from Neutron Irradiated Thorium Oxide

    International Nuclear Information System (INIS)

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

    1983-01-01

    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 2 material into ThF 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

  3. Neutron irradiation effects on high Nicalon silicon carbide fibers

    International Nuclear Information System (INIS)

    Osborne, M.C.; Steiner, D.; Snead, L.L.

    1996-01-01

    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

  4. Studies of neutron irradiation effects at IPNS-REF

    International Nuclear Information System (INIS)

    Kirk, M.A.

    1983-09-01

    Neutron irradiation effects studies at the Radiation Effects Facility (REF) at the Intense Pulsed Neutron Source (IPNS) located at Argonne National Laboratory (ANL) are reviewed. A brief history of the development of this user facility is followed by an overview of the scientific program. Experiments unique to a spallation neutron source are covered in more detail. Future direction of research at this facility is suggested

  5. National Low-Temperature Neutron-Irradiation Facility

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Klabunde, C.E.; Young, F.W. Jr.

    1983-08-01

    The Materials Sciences Division of the United States Department of Energy will establish a National Low Temperature Neutron Irradiation Facility (NLTNIF) which will utilize the Bulk Shielding Reactor (BSR) located at Oak Ridge National Laboratory. The facility will provide high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. This report describes the planned experimental capabilities of the new facility

  6. 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.

  7. First results of laser welding of neutron irradiated stainless steel

    International Nuclear Information System (INIS)

    Osch, E.V. van; Hulst, D.S. d'; Laan, J.G. van der.

    1994-10-01

    First results of experimental investigations on the laser reweldability of neutron irradiated material are reported. These experiments include the manufacture of 'heterogeneous' joints, which means joining of irradiated stainless steel of type AISI 316L-SPH to 'fresh' unirradiated material. The newly developed laser welding facility in the ECN Hot Cell Laboratory and experimental procedures are described. Visual inspections of welded joints are reported as well as results of electron microscopy and preliminary metallographic examinations. (orig.)

  8. In-situ field-ion microscope study of the recovery behavior of heavy metal ion-irradiated tungsten, tungsten (rhenium) alloys and molybdenum

    International Nuclear Information System (INIS)

    Nielsen, C.H.

    1977-06-01

    Three field ion microscope (FIM) experiments were carried out to study the annealing behavior of heavy ion irradiated tungsten, tungsten (rhenium) alloys and molybdenum. The first experiment dealt with the stage I long-range migration of tungsten self interstitial atoms (SIAs) in high purity tungsten of resistivity ratio, R = 24,000 (R = rho 300 /rho 4 . 2 , where rho 300 and rho 4 . 2 are the room temperature and 0 0 C resistivities). The FIM specimens were irradiated in situ at 18 K with 30 keV W + ions to an average dose of 5 x 10 12 ions cm -2 and subsequently examined by the pulsed-field evaporation technique. The second experiment dealt with the phenomenon of impurity atom trapping of SIAs during long-range migration. It was shown that rhenium atoms in a tungsten matrix tend to capture tungsten SIAs and remain bound up to temperatures as high as 390 K. The final experiment was concerned with the low temperature annealing kinetics of irradiated molybdenum. High purity molybdenum of resistivity ratio R = 5700 was irradiated at 10 K with 30 keV Mo + ions to a dose of approximately 5 x 10 12 ions cm -2 . The results indicated that the electric field has only a minimal effect on the SIA annealing kinetics. This tends to strengthen the contention that the molybdenum SIA becomes mobile at 32 K

  9. Needs of in-situ materials testing under neutron irradiation

    International Nuclear Information System (INIS)

    Noda, K.; Hishinuma, A.; Kiuchi, K.

    1989-01-01

    Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

  10. Detection and measurement of neutron-irradiated gemstones

    International Nuclear Information System (INIS)

    Bunnak, S.; Jerachanchai, S.; Chinudomsub, K.; Saiyut, K.

    1990-01-01

    Color enhance gemstone, neutron-irradiated topaz, was analyzed by gamma spectrometry for examining characteristic and activity. Topaz was irradiated in the wet-tube facility of the Research Reactor TRR/1 which neutron fluence is 2.52x10 17 neutron per square centimeter. After 100 days of decay, topaz was sampling to the qualitative and quantitative analysis using multichannel analyzer of Nuclear Data Model ND65 and hyper pure germanium detector. Calculation and evaluation were done by microcomputer IBM/PC 640 KB RAM. The qualitative analysis showed that the neutron-irradiated topaz has 2 major isotopes, i.e., Ta-182 and Sc-46. Quantitative activity was compared with reference standard source Eu-152 (NBS) and the results were shown in the table 1. The Health Physics Division, OAEP, inspected on 6240.9 gm of the neutron-irradiated topaz using standard release limit 2 nCi/gm (74 Bq/gm). It was found that only 423.9 gm out of the total amount were over the standard release limit

  11. Structural properties and neutron irradiation effects of ceramics

    International Nuclear Information System (INIS)

    Yano, Toyohiko

    1994-01-01

    In high temperature gas-cooled reactors and nuclear fusion reactors being developed at present, various ceramics are to be used in the environment of neutron irradiation for undertaking important functions. The change of the characteristics of those materials by neutron irradiation must be exactly forecast, but it has been known that the response of the materials is different respectively. The production method of ceramics and the resulted structures of ceramics which control their characteristics are explained. The features of covalent bond and ionic bond, the synthesis of powder and the phase change by heating, sintering and sintering agent, and grain boundary phase are described. The smelling of ceramics by neutron irradiation is caused by the formation of the clusters of Frenkel defects and minute spot defects. Its restoration by annealing is explained. The defects remaining in materials after irradiation are the physical defects by flipping atoms cut due to the collision with high energy particles and the chemical defects by nuclear transformation. Some physical defects can be restored, but chemical defects are never restored. The mechanical properties of ceramics and the effect of irradiation on them, and the thermal properties of ceramics and the effect of irradiation on them are reported. (K.I.)

  12. Moessbauer spectroscopy on amorphous Fe/sub x/Ni/sub 80-x/B20 after neutron irradiation

    International Nuclear Information System (INIS)

    Sitek, J.; Miglierini, M.

    1985-01-01

    Amorphous Fe/sub x/Ni/sub 80-x/B 20 glassy alloys (x = 40, 50, 60, and 70) irradiated with fast neutrons in a fluence range of 10 14 to 10 19 cm -2 were investigated by Moessbauer spectroscopy. There were some significant changes in the Moessbauer spectrum parameters of the 10 19 cm -2 irradiated samples except Fe 40 Ni 40 B 20 . This corresponds to a change in the direction of the easy axis of magnetization. The measurements show that the resistance of the Fe-Ni-B system against neutron irradiation improves with increasing Ni content up to a certain point

  13. Studying VM-1 molybdenum alloy workability at high current density. II

    Energy Technology Data Exchange (ETDEWEB)

    Tatarinova, O M; Amirkhanova, N A; Zaripov, R A

    1976-01-01

    Under galvanostatic conditions, voltampere characteristics have been taken off for VM-1 alloy; determined are also the selective effect of electrolytes and the influence of hydrodynamical conditions on the rate of anodic dissolution in the electrolytes containing 15% NaNO/sub 3/; 15% NaNO/sub 3/ + 5% NaOH, and 15 % NaOH. In a composite electrolyte, the quality of the surface is improved, and higher current densities have been attained as compared with those for pure 15% NaNO/sub 3/. The process of dissolution in the above electrolytes is effected with diffuse limitations. For the electrochemical treatment of the VM-1 alloy under production conditions, a composite electrolyte containing 15% NaNO/sub 3/ and 5% NaOH has been suggested and tested.

  14. 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)

  15. Influence of Molybdenum on the Alloying and Physical Properties of Fe-Al

    Czech Academy of Sciences Publication Activity Database

    Jirásková, Yvonna; Buršík, Jiří; Hapla, Miroslav; Turek, Ilja

    2015-01-01

    Roč. 28, č. 3 (2015), s. 905-910 ISSN 1557-1939 R&D Projects: GA ČR(CZ) GAP108/11/1350; GA MŠk(CZ) ED1.1.00/02.0068 Institutional support: RVO:68081723 Keywords : Nanostructured materials * Mechanical alloying * Magnetic proprties * Grain core * Defects Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.100, year: 2015

  16. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Languille, A.

    2000-01-01

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

  17. Controllable synthesis of molybdenum tungsten disulfide alloy for vertically composition-controlled multilayer

    Science.gov (United States)

    Song, Jeong-Gyu; Ryu, Gyeong Hee; Lee, Su Jeong; Sim, Sangwan; Lee, Chang Wan; Choi, Taejin; Jung, Hanearl; Kim, Youngjun; Lee, Zonghoon; Myoung, Jae-Min; Dussarrat, Christian; Lansalot-Matras, Clement; Park, Jusang; Choi, Hyunyong; Kim, Hyungjun

    2015-01-01

    The effective synthesis of two-dimensional transition metal dichalcogenides alloy is essential for successful application in electronic and optical devices based on a tunable band gap. Here we show a synthesis process for Mo1−xWxS2 alloy using sulfurization of super-cycle atomic layer deposition Mo1−xWxOy. Various spectroscopic and microscopic results indicate that the synthesized Mo1−xWxS2 alloys have complete mixing of Mo and W atoms and tunable band gap by systematically controlled composition and layer number. Based on this, we synthesize a vertically composition-controlled (VCC) Mo1−xWxS2 multilayer using five continuous super-cycles with different cycle ratios for each super-cycle. Angle-resolved X-ray photoemission spectroscopy, Raman and ultraviolet–visible spectrophotometer results reveal that a VCC Mo1−xWxS2 multilayer has different vertical composition and broadband light absorption with strong interlayer coupling within a VCC Mo1−xWxS2 multilayer. Further, we demonstrate that a VCC Mo1−xWxS2 multilayer photodetector generates three to four times greater photocurrent than MoS2- and WS2-based devices, owing to the broadband light absorption. PMID:26204328

  18. The effect of molybdenum content with changes in phase and heat capacity of UMo alloy

    International Nuclear Information System (INIS)

    Aslina Br Ginting; Supardjo; Agoeng Kadarjono; Dian Anggraini

    2011-01-01

    Has done the analysis of phase and heat capacity change of the UMo alloy by variation of 7% Mo, 8% and 9% Mo. Analysis performed using phase change Differential Thermal Analysis (DTA) at a temperature between 30°C until 1400°C with heating rate 10°C/minute and heat capacity analysis carried out using Differential Scanning Calorimetry (DSC) at a temperature between 30°C to 450°C with heating rate 5°C/minute. The purpose of this study was to determine the character of the UMo alloy include phase change and heat capacity variation with Mo content due to higher content of Mo is expected to change both the character U-7% Mo alloy, U-8% Mo and U-9% Mo. The analysis showed that of 7% Mo, 8% Mo and 9% Mo the combination experiencing α+ δ a phase change becomes α + β phase at temperatures of 578.63°C to 580.16°C. At the temperature 606.50°C to 627.58°C having a phase change of α+ β to β + γ be followed by the endothermic reaction in the content of 9% Mo with the enthalpy ΔH = 6.5989 J / g. At temperatures 1075.45°C up to 1160.51°C phase change β + γ into γ phase. The increase in Mo content to heating at a temperature 1100°C not cause a significant phase change. At temperatures above 1177.21°C, the increase in Mo content leads to changes in the γ phase of forming L + γ phase which followed the reaction of uranium with Mo to form γ phase - solid solution. The higher content of Mo, the reaction heat is needed and released the greater. The results of the analysis of the heat capacity is obtained that the increase in Mo content in the U-7% Mo, U-8% Mo, and U-9% Mo alloy does not give a significant difference in heat capacity. This is attested by doing different test (F test) at 95% degree of confidence. This data is expected to be as a first step to study the manufacture of UMo alloy as a fuel of high uranium density for research reactor. (author)

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

    International Nuclear Information System (INIS)

    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.

    2015-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, A.J., E-mail: aclarke@lanl.gov; 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.

    2015-10-15

    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.

  1. 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.

  2. Nanoprecipitates in single-crystal molybdenum-alloy nanopillars detected by TEM and atom probe tomography

    International Nuclear Information System (INIS)

    Oveisi, Emad; Bártová, Barbora; Gerstl, Stephan; Zimmermann, Julien; Marichal, Cécile; Van Swygenhoven, Helena; Hébert, Cécile

    2013-01-01

    Transmission electron microscopy (TEM) supported by various chemical analyses techniques as well as atom probe tomography is applied to characterize newly identified nanosized precipitates in Mo-alloy nanopillars that were prepared by directional solidification. It is shown that the α-Mo matrix contains Al-enriched face-centred cubic precipitates which have a 4.12 Å lattice parameter, and exhibit a Kurdjumov–Sachs crystallographic orientation relationship with the matrix. Such precipitates could be responsible for the unusual behaviour of the pillars during compression tests

  3. Effect of ion-implantation on surface characteristics of nickel titanium and titanium molybdenum alloy arch wires

    Directory of Open Access Journals (Sweden)

    Manu Krishnan

    2013-01-01

    Full Text Available Aim: To evaluate the changes in surface roughness and frictional features of ′ion-implanted nickel titanium (NiTi and titanium molybdenum alloy (TMA arch wires′ from its conventional types in an in-vitro laboratory set up. Materials and Methods: ′Ion-implanted NiTi and low friction TMA arch wires′ were assessed for surface roughness with scanning electron microscopy (SEM and 3 dimensional (3D optical profilometry. Frictional forces were studied in a universal testing machine. Surface roughness of arch wires were determined as Root Mean Square (RMS values in nanometers and Frictional Forces (FF in grams. Statistical Analysis Used: Mean values of RMS and FF were compared by Student′s ′t′ test and one way analysis of variance (ANOVA. Results: SEM images showed a smooth topography for ion-implanted versions. 3D optical profilometry demonstrated reduction of RMS values by 58.43% for ion-implanted NiTi (795.95 to 330.87 nm and 48.90% for TMA groups (463.28 to 236.35 nm from controls. Nonetheless, the corresponding decrease in FF was only 29.18% for NiTi and 22.04% for TMA, suggesting partial correction of surface roughness and disproportionate reduction in frictional forces with ion-implantation. Though the reductions were highly significant at P < 0.001, relations between surface roughness and frictional forces remained non conclusive even after ion-implantation. Conclusion: The study proved that ion-implantation can significantly reduce the surface roughness of NiTi and TMA wires but could not make a similar reduction in frictional forces. This can be attributed to the inherent differences in stiffness and surface reactivity of NiTi and TMA wires when used in combination with stainless steel brackets, which needs further investigations.

  4. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element

    International Nuclear Information System (INIS)

    Delaplace, J.

    1960-09-01

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the γ → β transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the β → α transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form β at ordinary temperatures after quenching from the β and γ regions. The β phase is particularly unstable and changes into needles of the α form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The β phase obtained by quenching from the β phase region is more stable than that obtained by quenching from the γ region. Chromium is a more effective stabiliser of the β phase than is iron. Unfortunately it causes serious surface cracking. The β → α transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct γ → α transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author) [fr

  5. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  6. 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)

  7. Isotropic nuclear graphites; the effect of neutron irradiation

    International Nuclear Information System (INIS)

    Lore, J.; Buscaillon, A.; Mottet, P.; Micaud, G.

    1977-01-01

    Several isotropic graphites have been manufactured using different forming processes and fillers such as needle coke, regular coke, or pitch coke. Their properties are described in this paper. Specimens of these products have been irradiated in the fast reactor Rapsodie between 400 to 1400 0 C, at fluences up to 1,7.10 21 n.cm -2 PHI.FG. The results show an isotropic behavior under neutron irradiation, but the induced dimensional changes are higher than those of isotropic coke graphites although they are lower than those of conventional extruded graphites made with the same coke

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  9. Ultrasonic attenuation measurements and 'glassy' behaviour of neutron irradiated quartz

    International Nuclear Information System (INIS)

    Laermans, C.; Esteves, V.; Vanelstraete, A.

    1986-01-01

    The ultrasonic attenuation of longitudinal acoustic waves in slightly disordered crystalline quartz has been measured over a temperature range from 1.3 to 300 K, using the pulse-echo technique. Neutron irradiation is demonstrated to increase the ultrasonic attenuation at low temperatures indicating the presence of two-level tunneling systems similar to those of glasses. The present low-temperature acoustic results agree with a frequency independence and a T 3 behaviour for the relaxation process predicted by the two-level tunneling TLS-model where the regime ωT 1 >> 1 holds. (author)

  10. Gamma and neutron irradiation tests on commercial IC op amps

    International Nuclear Information System (INIS)

    Kennedy, E.J.; Morris, A.C. Jr.; Su, D.K.

    1985-01-01

    Experimental results of gamma and neutron irradiation tests on 30 types of integrated-circuit operational amplifiers from 11 manufacturers are presented. All units were low-cost, commercial-grade devices. Op amps were evaluated for changes in offset voltage, input bias current, power supply current, open-loop gain, gain-bandwidth product, slew rate, power-supply and common-mode rejection ratios. Bipolar transistor op amps with resistive collector load resistors for the input stage indicated the best radiation hardness

  11. Radiation defects produced by neutron irradiation in germanium single crystals

    International Nuclear Information System (INIS)

    Fukuoka, Noboru; Honda, Makoto; Atobe, Kozo; Yamaji, Hiromichi; Ide, Mutsutoshi; Okada, Moritami.

    1992-01-01

    The nature of defects produced in germanium single crystals by neutron irradiation at 25 K was studied by measuring the electrical resistivity. It was found that two levels located at E c -0.06 eV and E c -0.13 eV were introduced in an arsenic-doped sample. Electron traps at E c -0.10eV were observed in an indium-doped sample. The change in electrical resistivity during irradiation was also studied. (author)

  12. Swelling and fracturing of borides under neutron irradiation

    International Nuclear Information System (INIS)

    Krainy, A.G.; Ogorodnikov, V.V.; Grinik, E.U.; Chirko, L.I.; Shinakov, A.A.

    1994-01-01

    The neutron irradiation of high temperature borides, which are included in boron-containing reactor materials, results in high internal stresses, leading to considerable swelling and micro- and macro-fracturing. Experimental results over a large range of temperature and fluences, show a change of damage mechanism for borides within 400-530 C: the macro-cracking with formation of annular and radial cracks is observed below this temperature zone. The accumulation of micro-fractures and the process of gas swelling take place at irradiation temperatures above 530 C. The effect of the high internal stresses is compared to external pressure. 12 refs., 4 figs

  13. Effects of neutron irradiation on a superconducting metallic glass

    International Nuclear Information System (INIS)

    Kramer, E.A.; Johnson, W.L.; Cline, C.

    1979-06-01

    The effects of fast neutron irradiation on a superconducting metallic glass (Mo 6 Ru 4 ) 82 B 18 have been studied. Following irradiation to a total fluence of 10 19 n/cm 2 , T/sub c/ increases from 6.05 K to 6.19 K, and the width of the transition decreases sharply. The density of the material decreases by 1.5%, and the x-ray scattering intensity maxima are broadened. An improvement in the ductility of the samples is observed which together with the other observations suggests the production of defects having atomic scale dimensions and characterized by excess volume

  14. An integrated approach for microstructure simulation: Application to nickel-aluminum-molybdenum alloys

    Science.gov (United States)

    Wang, Tao

    The properties and performance of a material are strongly dependent on its microstructure. For example, the gamma' precipitate coherently embedded in the gamma matrix is the primary strengthening phase in Ni-base superalloys, and its volume fraction, morphology and size distribution largely determine the strength, fatigue and creep properties of an alloy. In the present study, a multiscale computational approach was proposed to predict the microstructure evolution in Ni-base superalloys. It integrated a quantitative phase-field model with first-principles calculations as well as the CALPHAD (CALculation of PHAse Diagram) technique. Fundamental materials property databases such as lattice parameters and atomic mobility were developed. A phenomenological model was developed to describe the lattice parameter in solid states as a function of temperature and composition, and successfully applied to Ni-Al binary system by evaluating the model parameters using experimental data. An integrated computational approach was also proposed for evaluating the lattice misfit between the matrix and precipitates by combining first-principles calculations, existing experimental data and phenomenological modeling when the experimental data is limited. The lattice parameters and the local lattice distortions around solute atoms in gamma-Ni solutions were studied using first-principles calculations. The solute atoms considered include Al, Co, Cr, Hf, Mo, Nb, Re, Ru, Ta, Ti and W. The effects of the atomic size and the electronic and magnetic interactions on lattice distortion have been discussed. Atomic mobility in disordered gamma and ordered gamma' phases was modeled for the Ni-Al-Mo ternary system, and a kinetic database was developed. The diffusion of Al in gamma' was simulated, and the formation energies of vacancy in different sublattices were calculated by first-principles approach, both of which indicate the anti-site diffusion mechanism being dominant for diffusion of Al. The

  15. The role of molybdenum in suppressing cold dwell fatigue in titanium alloys

    Science.gov (United States)

    Ready, Adam J.; Haynes, Peter D.; Grabowski, Blazej; Rugg, David; Sutton, Adrian P.

    2017-07-01

    We test a hypothesis to explain why Ti-6242 is susceptible to cold dwell fatigue (CDF), whereas Ti-6246 is not. The hypothesis is that, in Ti-6246, substitutional Mo-atoms in α-Ti grains trap vacancies, thereby limiting creep relaxation. In Ti-6242, this creep relaxation enhances the loading of grains unfavourably oriented for slip and they subsequently fracture. Using density functional theory to calculate formation and binding energies between Mo-atoms and vacancies, we find no support for the hypothesis. In the light of this result, and experimental observations of the microstructures in these alloys, we agree with the recent suggestion (Qiu et al. 2014 Metall. Mater. Trans. A 45, 6075-6087. (doi:10.1007/s11661-014-2541-5)) that Ti-6246 has a much smaller susceptibility to CDF because it has a smaller grain size and a more homogeneous distribution of grain orientations. We propose that the reduction of the susceptibility to CDF of Ti-6242 at temperatures above about 200°C is due to the activation of slip in `hard' grains, which reduces the loading of grain boundaries.

  16. [sup 57]Fe Moessbauer study of amorphous and nanocrystalline Fe[sub 73. 5]Nb[sub 3]Cu[sub 1]Si[sub 13. 5]B[sub 9] after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Miglierini, M. (Dept. of Nuclear Physics and Technology, Slovak Technical Univ., Bratislava (Slovakia)); Sitek, J. (Dept. of Nuclear Physics and Technology, Slovak Technical Univ., Bratislava (Slovakia)); Szasz, Z. (Dept. of Nuclear Physics and Technology, Slovak Technical Univ., Bratislava (Slovakia)); Vitazek, K. (Dept. of Nuclear Physics and Technology, Slovak Technical Univ., Bratislava (Slovakia))

    1994-05-01

    [sup 57]Fe Moessbauer spectroscopy is used to study neutron irradiation induced changes in the short-range order of Fe[sub 73.5]Nb[sub 3]Cu[sub 1]Si[sub 13.5]B[sub 9] alloy. The samples are investigated in both amorphous and nanocrystalline states. Neutron irradiation leads to an increase of the standard deviation of a hyperfine field distribution (HFD), implying rearrangement of the atoms towards disordering. Simultaneously, changes in the average value of the hyperfine field and a net magnetic moment position occur as a consequence of a spin reorientation, atom mixing and microscopic stress centres which are introduced by neutron irradiation. (orig.)

  17. Extractive spectrophotometric determination of molybdenum in steels and nickel base high-temperature alloys as a ternary complex with hydroxylamine hydrochloride and 4-(2-pyridylazo) resorcinol

    International Nuclear Information System (INIS)

    Reddy, M.R.P.; Kumar, P.V.S.; Shyamsundar, J.P.; Anjaneyulu, Y.

    1990-01-01

    The red coloured, ternary complex formed by molybdenum(VI) with hydroxylamine hydrchloride and 4-(2-pyridylazo) resorcinol (PAR), on heating for 20 minutes in a hot water bath at pH 6-8, can be extracted with a mixture of n-butanol and benzene (4:1). This complex exhibits absorption maximum at 530nm with a molar absorptivity of 2.8x10 4 lit. mol -1 cm -1 and obeys Beer's law upto 4.5μg/ml of Mo(VI). The composition of the complex is found to be 1:1:1, for Mo(VI): hydroxylamine hydrochloride: PAR. Large amounts of tartaric acid and 2, 2-diamino-cyclohexane tetra acetic acid (CyDTA) can be tolerated in this method. The extraction becomes highly selective in presence of CyDTA (3ml of 5x10 -2 M) and interference of ions like Fe(III)(5mg), Pb(II)(4mg), Zn(II)(2mg), Cu(II)(4mg), Mn(II)(3mg), Ni(II)(4mg), Sn(II)(3mg), Cr(VI)(2mg) and Bi(III)(0.5mg) can be effectively supressed in the determination of 10-45μg of molybdenum. The method can be successfully applied for the determination of molybdenum in steels and nickel base high temperature alloys. (author). 2 tabs., 8 refs

  18. Impurities effect on the swelling of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Donne, M.D.; Scaffidi-Argentina, F.

    1995-01-01

    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

  19. Neutron irradiation test of depleted CMOS pixel detector prototypes

    International Nuclear Information System (INIS)

    Mandić, I.; Cindro, V.; Gorišek, A.; Hiti, B.; Kramberger, G.; Mikuž, M.; Zavrtanik, M.; Hemperek, T.; Daas, M.; Hügging, F.; Krüger, H.; Pohl, D.-L.; Wermes, N.; Gonella, L.

    2017-01-01

    Charge collection properties of depleted CMOS pixel detector prototypes produced on p-type substrate of 2 kΩ cm initial resistivity (by LFoundry 150 nm process) were studied using Edge-TCT method before and after neutron irradiation. The test structures were produced for investigation of CMOS technology in tracking detectors for experiments at HL-LHC upgrade. Measurements were made with passive detector structures in which current pulses induced on charge collecting electrodes could be directly observed. Thickness of depleted layer was estimated and studied as function of neutron irradiation fluence. An increase of depletion thickness was observed after first two irradiation steps to 1 · 10 13 n/cm 2 and 5 · 10 13 n/cm 2 and attributed to initial acceptor removal. At higher fluences the depletion thickness at given voltage decreases with increasing fluence because of radiation induced defects contributing to the effective space charge concentration. The behaviour is consistent with that of high resistivity silicon used for standard particle detectors. The measured thickness of the depleted layer after irradiation with 1 · 10 15 n/cm 2 is more than 50 μm at 100 V bias. This is sufficient to guarantee satisfactory signal/noise performance on outer layers of pixel trackers in HL-LHC experiments.

  20. High Ni austenite stainless steel resistant to neutron irradiation degradation

    International Nuclear Information System (INIS)

    Yonezawa, Toshio; Iwamura, Toshihiko; Kanasaki, Hiroshi; Fujimoto, Koji; Nakata, Shizuo; Ajiki, Kazuhide; Nakamura, Mitsuhiro.

    1997-01-01

    The composition of the stainless steel of the present invention comprises from 0.005 to 0.08% of C, up to 3% of Mn, up to 0.2% of Si+P+S, from 25 to 40% of Ni, from 25 to 40% of Cr, up to 3% of Mo, up to 0.3% of Nb+Ta, up to 0.3% of Ti, up to 0.001% of B and the balance of Fe. A solid solubilization treatment at a temperature of from 1,000 to 1,150degC is applied to the stainless steel having the composition. The stainless steel is excellent in stress corrosion cracking-resistance at a working circumstance of a LWR type reactor (high temperature and high pressure water at from 270 to 350degC/from 70 to 160 atm even after undergoing neutron irradiation of about 1 x 10 22 n/cm 2 (E>1 MeV) which is a maximum neutron irradiation amount undergone till the final stage of the working life of the LWR-type reactor. In addition, the average thermal expansion coefficient at from room temperature to 400degC ranges from 15x10 -6 - 19x10 -6 /K. (I.N.)

  1. Results of neutron irradiation of liquid lithium saturated with deuterium

    International Nuclear Information System (INIS)

    Tazhibayeva, Irina; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna; Lyublinski, Igor; Vertkov, Alexey; Mazzitelli, Giuseppe

    2017-01-01

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(−144/RT). • The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10 −13 cm −2 s −1 . The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(-144/RT). The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

  2. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

  3. 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....

  4. Recovery in stages I and II of thermal and fission neutron irradiated molybdenum

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Klabunde, C.E.; Redman, J.K.

    1975-01-01

    The influence of initial dose and irradiation doping upon the recovery of Mo was studied for the markedly different types of damage produced by thermal and fission neutrons. The features of the Stage I recovery commonly seen for several fcc metals can be identified, including a I/sub D/ peak at 40 0 K. The typical dose-dependent behavior of a I/sub E/ subpeak was observed at approximately 47 0 K, and evidence for free interstitial migration is further supported by irradiation doping results. Stage II shows a first-order peak at 120 0 K in which the population percentage increases with increasing initial dose in opposite fashion to fcc impurity detrapping peaks. (auth)

  5. Preparation of 227Ac by neutron irradiation of 226Ra

    International Nuclear Information System (INIS)

    Kukleva, E.; Kozempel, J.; Vlk, M.; Micolova, P.; Vopalka, D.

    2015-01-01

    Radium-223 is prospective alpha-emitting therapeutic radionuclide for targeted radionuclide therapy. Although 223 Ra is formed naturally by the decay of 235 U, for practical reasons its preparation involves neutron irradiation of 226 Ra. The α-decay of the 227 Ra (T 12 = 43 min.) produced via 226 Ra(n,γ) 227 Ra reaction leads to 227 Ac, a mother nuclide of 227 Th and 223 Ra subsequently. Irradiation target radium material is generally available in multi-gram quantities from historical stock. Main aim of this study was to experimentally and theoretically evaluate and verify available literature data on production of 223 Ra. According to data obtained from γ-spectra, the approximate yield values were determined and effective cross-section for the 223 Ra production was calculated. (authors)

  6. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  7. 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)

  8. 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.)

  9. Damages to gladiolu corm caused by fast neutron irradiation

    International Nuclear Information System (INIS)

    Zhang Zhiwei; Wang Dan; Zhang Dongxue; Zheng Chun

    2007-01-01

    Gladiolus corms were irradiated to 100-500kGy by fast neutrons in the CFBR-II pulsed reactor, Scanning electron microscope images of the irradiated samples revealed significant radiation damages to the gladiolus corms, and the mutagenic effects were studied by SDS-polyacrylamide gel electrophoresis (SDS-PAGE). Within the dose range, radiation damage to the corm increased with the dose, with corm epidermis of the samples irradiated in vertical incidence being more serious than those irradiated in side-incidence to the same dose. Biological characters were investigated via field experiments, and the bands of protein subunit were analyzed by SDS-PAGE. The results showed that the fast neutrons irradiation inhibited growth of M1 generation seedling significantly. Protein expression was obviously inhibited by the irradiation. The study indicates that fast neutron induction is an effective way for gladiolus breeding. And the results may lay a foundation for studies on fast neutron mutation breeding. (authors)

  10. Investigation of radiation damage effects in neutron irradiated CCD

    International Nuclear Information System (INIS)

    Brau, James E.; Igonkina, Olga; Potter, Chris T.; Sinev, Nikolai B.

    2005-01-01

    A Charge Coupled Devices (CCD)-based vertex detector is a leading option for vertex detection at the future linear collider. A major issue for this application is the radiation hardness of such devices. Tests of radiation hardness of CCDs used in the SLD vertex detector, VXD3, have been reported earlier. The first measurements of 1998 involved a spare VXD3 CCD that was irradiated with neutrons from a radioactive source (Pu-Be), and from a nuclear reactor. In 2003, we had the opportunity to disassemble the VXD3 detector and study the nature of the radiation damage it incurred during 3 years of operation at SLC. In the preparation for this study, additional experiments with the spare VXD3 CCD were performed. These included measurements of trapping times in neutron irradiated CCDs. Results, reported here, will help us better understand the mechanism of radiation damage effects and develop techniques to minimize performance degradation due to radiation damage

  11. Effect of neutron irradiation on single crystal V3Si

    International Nuclear Information System (INIS)

    Viswanathan, R.; Caton, R.; Cox, D.E.; Guha, A.; Sarachik, M.P.; Smith, F.W.; Testardi, L.R.

    1977-01-01

    We We have investigated the effect of successive neutron irradiation up to a fluence of approximately 2 x 10 19 n/cm 2 , by measurements of heat capacity, susceptibility, resistivity, acoustic velocity and neutron diffraction in a single crystal V 3 Si. We find that for low level doses (phi t greater than or equal to 3.5 x 10 18 n/cm 2 ) (a) the structural transformation is very sensitive, whereas the suerconducting transition temperature, T/sub c/, is hardly affected, and (b) except for low temperature heat capacity, most of the other measurements show very little change. For the highest fluence of 2 x 10 19 n/cm 2 used to date, the T/sub c/ dropped to 7.5 K with large changes in the linear heat capacity coefficient, magnetic susceptibility and sound velocity. These results are discussed briefly in this paper

  12. 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.

  13. Effect of neutron irradiation on p-type silicon

    International Nuclear Information System (INIS)

    Sopko, B.

    1973-01-01

    The possibilities are discussed of silicon isotope reactions with neutrons of all energies. In the reactions, 30 Si is converted to a stable phosphorus isotope forming n-type impurities in silicon. The above reactions proceed as a result of thermal neutron irradiation. An experiment is reported involving irradiation of two p-type silicon single crystals having a specific resistance of 2000 ohm.cm and 5000 to 20 000 ohm.cm, respectively, which changed as a result of irradiation into n-type silicon with a given specific resistance. The specific resistance may be pre-calculated from the concentration of impurities and the time of irradiation. The effects of irradiation on other silicon parameters and thus on the suitability of silicon for the manufacture of semiconductor elements are discussed. (J.K.)

  14. Point defects in cubic boron nitride after neutron irradiation

    International Nuclear Information System (INIS)

    Atobe, Kozo; Honda, Makoto; Ide, Munetoshi; Yamaji, Hiromichi; Matsukawa, Tokuo; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

    1993-01-01

    The production of point defects induced by reactor neutrons and the thermal behavior of defects in sintered cubic boron nitride are investigated using the optical absorption and electron spin resonance (ESR) methods. A strong structureless absorption over the visible region was observed after fast neutron irradiation to a dose of 5.3 x 10 16 n/cm 2 (E > 0.1 MeV) at 25 K. This specimen also shows an ESR signal with g-value 2.006 ± 0.001, which can be tentatively identified as an electron trapped in a nitrogen vacancy. On examination of the thermal decay of the signal, the activation energy for recovery of the defects was determined to be about 1.79 eV. (author)

  15. Dielectric changes in neutron-irradiated rf window materials

    International Nuclear Information System (INIS)

    Frost, H.M.; Clinard, F.W. Jr.

    1987-01-01

    Ceramics used for windows in ECRH heating systems for magnetically-confined fusion reactors must retain adequate properties during and after intense neutron irradiation. Of particular concern is a decrease in transmissivity, a parameter inversely related to the product of dielectric constant K and loss tangent tanδ. Samples of polycrystalline Al 2 O 3 and BeO were irradiated to 1 x 10 26 n/m 2 at 660K in the EBR-II fission reactor, and the above properties subsequently measured at 95 GHz. It was found that ktanδ for both materials doubled, implying a doubling of thermal stresses and a consequent reduction of time-to-failure from an assumed one year to 20 min for beryllia and 2 s for alumina. In the case of BeO, a large increase in reflectance of the incident millimeter-wave power results from dielectrically uncompensated swelling. This phenomenon could significantly degrade source performance

  16. 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.

  17. Proceedings of neutron irradiation technical meeting on BNCT

    International Nuclear Information System (INIS)

    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.)

  18. Germanium-doped gallium phosphide obtained by neutron irradiation

    Science.gov (United States)

    Goldys, E. M.; Barczynska, J.; Godlewski, M.; Sienkiewicz, A.; Heijmink Liesert, B. J.

    1993-08-01

    Results of electrical, optical, electron spin resonance and optically detected magnetic resonance studies of thermal neutron irradiated and annealed at 800 °C n-type GaP are presented. Evidence is found to support the view that the main dopant introduced via transmutation of GaP, germanium, occupies cation sites and forms neutral donors. This confirms the possibility of neutron transmutation doping of GaP. Simultaneously, it is shown that germanium is absent at cation sites. Presence of other forms of Ge-related defects is deduced from luminescence and absorption data. Some of them are tentatively identified as VGa-GeGa acceptors leading to the self-compensation process. This observation means that the neutron transmutation as a doping method in application to GaP is not as efficient as for Si.

  19. Results of neutron irradiation of liquid lithium saturated with deuterium

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Lyublinski, Igor [JSC «Red Star», Moscow (Russian Federation); NRNU «MEPhI», Moscow (Russian Federation); Vertkov, Alexey [JSC «Red Star», Moscow (Russian Federation); Mazzitelli, Giuseppe [ENEA, RC Frascati, Frascati (Italy)

    2017-04-15

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(−144/RT). • The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10{sup −13} cm{sup −2} s{sup −1}. The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(-144/RT). The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

  20. Features of soldering of molybdenum a lols

    International Nuclear Information System (INIS)

    Grishin, V.L.; Rybkin, B.V.; Cherkasov, A.F.

    1980-01-01

    Soldering features of complex-alloy molybdenum alloys were investigated in comparison with alloys based on solid solutions. Soldering features of heterogeneous molybdenum base alloys were investigated using samples of 0.5-1.O mm sheets with the strain of about 95% made of ingots which had been smelted in arc vacuum furnaces. The soldering of samples was carried out in 5x1O -5 mm Hg vacuum using different sources of heating: radiation, electron-ray and contact. It was shown that heat-resisting soldered joints of heterogeneous molybdenum alloys could be produced using zirconium and niobium base solders containing the most effective hardeners of the parent material (titanum, vanadium, tantalum, molybdenum, tungsten). To preserve high mechanical properties of heterogeneous alloys it was expedient to use for welding local heating sources which permitted to decrease considerably temperature- time conditions of the process

  1. Shielding of a neutron irradiator with {sup 241}Am-Be source

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

  2. Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

    International Nuclear Information System (INIS)

    Byun, T.S.; Farrell, K.; Lee, E.H.; Hunn, J.D.; Mansur, L.K.

    2001-01-01

    Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels

  3. Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

    International Nuclear Information System (INIS)

    Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

    2010-01-01

    An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

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

    International Nuclear Information System (INIS)

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

    1987-08-01

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

  5. Establishment of Wear Resistant HVOF Coatings for 50CrMo4 Chromium Molybdenum Alloy Steel as an Alternative for Hard Chrome Plating

    Science.gov (United States)

    Karuppasamy, S.; Sivan, V.; Natarajan, S.; Kumaresh Babu, S. P.; Duraiselvam, M.; Dhanuskodi, R.

    2018-05-01

    High cost imported components of seamless steel tube manufacturing plants wear frequently and need replacement to ensure the quality of the product. Hard chrome plating, which is time consuming and hazardous, is conventionally used to restore the original dimension of the worn-out surface of the machine components. High Velocity Oxy-Fuel (HVOF) thermal spray coatings with NiCrBSi super alloy powder and Cr3C2 NiCr75/25 alloy powder applied on a 50CrMo4 (DIN-1.7228) chromium molybdenum alloy steel, the material of the wear prone machine component, were evaluated for use as an alternative for hard chrome plating in this present work. The coating characteristics are evaluated using abrasive wear test, sliding wear test and microscopic analysis, hardness test, etc. The study results revealed that the HVOF based NiCrBSi and Cr3C2NiCr75/25 coatings have hardness in the range of 800-900 HV0.3, sliding wear rate in the range of 50-60 µm and surface finish around 5 microns. Cr3C2 NiCr75/25 coating is observed to be a better option out of the two coatings evaluated for the selected application.

  6. 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.

  7. Radiation embrittlement behavior of fine-grained molybdenum alloy with 0.2 wt%TiC addition

    Energy Technology Data Exchange (ETDEWEB)

    Kitsunai, Y. [Tohoku University (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan)]. E-mail: kurishi@imr.tohoku.ac.jp; Kuwabara, T. [Tohoku University (Japan); Narui, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Hasegawa, M. [International Research Center for Nuclear Materials Science, Institute for Materials research (IMR), Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Takida, T. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan); Takebe, K. [A.L.M.T. TECH Inc., 2 Iwasekoshi-machi, Toyama 931-8543 (Japan)

    2005-11-15

    In order to elucidate the effects of pre-irradiation microstructures and irradiation conditions on radiation embrittlement and radiation-induced ductilization (RIDU), fine-grained Mo-0.2 wt%TiC specimens with high and low reduction rates in plastic working, which are designated as MTC-02H and MTC-02L, respectively, were prepared by powder metallurgical methods. The specimens were neutron irradiated to 0.1-0.15 dpa with controlled 1-cycle and 4-cycle heating between 573 and 773 K, and 473 and 673 K, respectively, in JMTR. Vickers microhardness and three-point bending impact tests and TEM microstructural examinations were made. The degree of radiation embrittlement, assessed by DBTT shift due to irradiation, was strongly dependent on the reduction rate and cycle number. The 4-cycle irradiation suppressed the radiation embrittlement compared with the 1-cycle irradiation, and the suppression was much more significant in MTC-02L than in MTC-02H. The observed behavior is discussed in connection with RIDU and microstructural evolution caused by the 4-cycle irradiation.

  8. Helium influence on the microstructure and swelling of 9%Cr ferritic steel after neutron irradiation to 16.3 dpa

    International Nuclear Information System (INIS)

    Klimenkov, M.; Möslang, A.; Materna-Morris, E.

    2014-01-01

    Specially fabricated samples of the European reference 9Cr-WTaV steel EUROFER 97 alloyed with 0.081 mass% natural B and 0.081 and 0.114 mass% pure isotope 10 B were neutron-irradiated with about 16.3 dpa at temperatures in the range from 523 K to 723 K to study the influence of helium produced by 10 B(n,α) 7 Li transmutation reaction on microstructure, swelling and hardness. The spatial and size distributions of helium bubbles or cavities after irradiation at different temperatures were investigated by transmission electron microscopy. Vickers microhardness HV0.1 tests were performed on the as received specimens and specimens after irradiation. The influence of irradiation temperature and helium concentration on the size and density of the bubbles or cavities was analyzed and correlations with the hardness, tensile properties, and the fracture surface were discussed

  9. Moessbauer spectroscopy on amorphous Fe/sub x/Ni/sub 80-x/B/sub 20/ after neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Sitek, J.; Miglierini, M. (Slovenska Vysoka Skola Technicka, Bratislava (Czechoslovakia))

    1985-05-16

    Amorphous Fe/sub x/Ni/sub 80-x/B/sub 20/ glassy alloys (x = 40, 50, 60, and 70) irradiated with fast neutrons in a fluence range of 10/sup 14/ to 10/sup 19/ cm/sup -2/ were investigated by Moessbauer spectroscopy. There were some significant changes in the Moessbauer spectrum parameters of the 10/sup 19/ cm/sup -2/ irradiated samples except Fe/sub 40/Ni/sub 40/B/sub 20/. This corresponds to a change in the direction of the easy axis of magnetization. The measurements show that the resistance of the Fe-Ni-B system against neutron irradiation improves with increasing Ni content up to a certain point.

  10. Uranium doping and neutron irradiation of Bi-2223 superconduction tapes for improved critical current density

    International Nuclear Information System (INIS)

    Moss, S.D.; Wang, W.G.; Dou, S.X.; Weinstein, R.

    1998-01-01

    It is demonstrated that a combination of neutron irradiation with uranium doping introduce fission tracks through a Bi-2223 tape which act as effective pinning centres, leading to a substantial increase in critical current. Preliminary data suggests that the combination of uranium doping and neutron irradiation produces improved flux pinning in Bi-2223 tapes over neutron irradiation alone. Before irradiation, SEM, DTA and XRD analyses were performed on the tapes. Both before and after irradiation the trapped maximum magnetic flux was measured at 77K. Before neutron irradiation, uranium doping has no effect on critical current. Preliminary SEM data suggested that the uranium is homogeneously distributed throughout the oxide core of the tape. The presence of 2212 and other secondary phases in the doped tapes suggest further refinement of the sintering procedure is necessary. (authors)

  11. Metal ion protection of DNA to fast neutron irradiation

    International Nuclear Information System (INIS)

    Constantinescu, B.; Bugoi, R.; Radulescu, I.; Radu, L.

    1998-01-01

    The most important effects of the ionising radiation are the single and double strand breaks (SSB and DBS), modifications of the DNA bases and deoxyribose, as well as the occurrence of alkali and heat labile sites (revealed as strand breaks after alkaline or thermic treatment of irradiated DNA). The ionising particles can have either direct effects on the DNA constituents or indirect effects, mediated by the OH - radicals, produced by water radiolysis. The occurrence of SSB and DSB in the chromatin DNA strands is supposed to hinder the DNA-dye complex formation. Usually, the dyes present different fluorescence parameters in the two possible states, so one can correlate the lifetime or the quantum yield with the extent of the damage. We took into account the protective effect offered both by histones, which behave as 'scavenger molecules' for OH - radicals and by the high compactness of DNA chromatin. Similar protective effects might be the results of the metallic ion addition which triggers some conformational transitions of the chromatin DNA towards a highly compacted structure. In this paper we present a study of the complexes of fast neutron irradiated chromatin with proflavine. Fluorimetric and time resolved spectroscopic determinations (single photon counting method) of chromatin-Pr complexes were realised. Information regarding the chromatin protein damage were obtained by monitoring the fluorescence of Trp. The chromatin was irradiated (20-100 Gy) with fast neutrons, obtained by the reaction of 13.5 MeV deuterons on a thick beryllium target at the IFIN-HH U-120 Cyclotron. The dose mean lineal energy in water at the point of interest was 50 keV/m and the mean dose rate was 1.5 Gy/min. By fluorescence determinations, changes of the Pr intercalation parameters in fast neutron irradiated chromatin DNA have been observed. Fluorescence techniques provide valuable information on the binding equilibrium by considering the radiation deexcitation of the complex. The

  12. 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

  13. Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content; Comportement au cyclage thermique et stabilite thermique des alliaces uranium-molybdene de faibles teneurs en molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Decours, J; Fabrique, B; Peault, O [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the {gamma}-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The {alpha} grain is fine, the {gamma}-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the {alpha}-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the {alpha}-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors) [French] Nous avons etudie le comportement au cyclage thermique des alliages U-Mo, brut de coulee, dont la teneur varie de 0,5 a 3 pour cent de molybdene, les resultats de stabilite du grain au cours de traitements thermiques de longue duree, ainsi que ceux des traitements combines de longue duree et de cyclage. Les

  14. Nuclear graphite based on coal tar pitch; behavior under neutron irradiation between 400 and 14000C

    International Nuclear Information System (INIS)

    Mottet, P.; Fillatre, A.; Schill, R.; Micaud, G.

    1977-01-01

    Two nuclear grades of coal tar pitch coke graphites have been developed and tested under neutron irradiation. The neutron irradiation induced dimensional changes between 400 and 1400 0 C, at fluences up to 1,2.10 22 n.cm -2 PHI.FG show a behavior comparable to anisotropic petroleum coke graphites. Less than 10% variation in thermal expansion, maximum decrease by a factor four in thermal conductivity, and large increase of the Young modulus have been observed

  15. Present status of ESNIT (energy selective neutron irradiation test facility) program

    International Nuclear Information System (INIS)

    Noda, K.; Ohno, H.; Sugimoto, M.; Kato, Y.; Matsuo, H.; Watanabe, K.; Kikuchi, T.; Sawai, T.; Usui, T.; Oyama, Y.; Kondo, T.

    1994-01-01

    The present status of technical studies of a high energy neutron irradiation facility, ESNIT (energy selective neutron irradiation test facility), is summarized. Technological survey and feasibility studies of ESNIT have continued since 1988. The results of technical studies of the accelerator, the target and the experimental systems in ESNIT program were reviewed by an International Advisory Committee in February 1993. Recommendations for future R and D on ESNIT program are also summarized in this paper. ((orig.))

  16. Inhomogeneous strain induced by fast neutron irradiation in NaKSO/sub 4/ crystals

    Energy Technology Data Exchange (ETDEWEB)

    Kandil, S.H.; Kassem, M.E.; El-Khatib, A.; El-Gamal, M.A.; El-Wahidy, E.F.

    1987-11-01

    The paper reports the effect of fast neutron irradiation on the thermal properties of NaKSO/sub 4/ crystals in the temperature range 400-475 K. Results are presented for the thermal expansion, tensile strain and specific heat of NaKSO/sub 4/, as a function of neutron irradiation dose. All these results revealed an inhomogeneous strain induced by the radiation. It is suggested that this induced inhomogeneous strain could be used to detect neutron exposure doses.

  17. Inhomogeneous strain induced by fast neutron irradiation in NaKSO4 crystals

    International Nuclear Information System (INIS)

    Kandil, S.H.; Kassem, M.E.; El-Khatib, A.; El-Gamal, M.A.; El-Wahidy, E.F.

    1987-01-01

    The paper reports the effect of fast neutron irradiation on the thermal properties of NaKSO 4 crystals in the temperature range 400-475 K. Results are presented for the thermal expansion, tensile strain and specific heat of NaKSO 4 , as a function of neutron irradiation dose. All these results revealed an inhomogeneous strain induced by the radiation. It is suggested that this induced inhomogeneous strain could be used to detect neutron exposure doses. (UK)

  18. 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.

  19. Study of natural diamond detector spectrometric properties under neutron irradiation

    CERN Document Server

    Alekseyev, A B; Kaschuck, Y; Krasilnikov, A; Portnov, D; Tugarinov, S

    2002-01-01

    Natural diamond detector (NDD) performance was studied up to a neutron fluence of 10 sup 1 sup 5 neutron/cm sup 2. The variations of the NDD spectrometric response to incident alpha-particles from sup 2 sup 4 sup 1 Am source after exposure to fast neutron fluences up to 3x10 sup 1 sup 6 n/cm sup 2 were examined. No significant variations up to the level of 10 sup 1 sup 4 n/cm sup 2 were observed. Degradation of charge collection efficiency at higher fluences is reported. No remarkable increase of the NDD leakage current and count rate change had been observed up to a neutron fluence of 3x10 sup 1 sup 6 n/cm sup 2. The charge collection efficiency variations of neutron irradiated diamond spectrometer were studied ex situ under gamma-rays, beta-radiation and visible light excitation. Charge collection efficiency restoration up to 75% level and the NDD performance stabilization by extrinsic low-intensity visible light (550 nm

  20. Hyper-thermal neutron irradiation field for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1994-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

  1. Neutron irradiation of sputtered NbN films

    International Nuclear Information System (INIS)

    Weber, H.W.; Gregshammer, P.; Gray, K.E.; Kampwirth, R.T.

    1989-01-01

    In addition to the excellent high-field superconducting properties of NbN, it is the strain and radiation tolerance, measured in bulk NbN, which makes the material so attractive for large high-field magnets, especially for fusion applications. The authors report neutron irradiation experiments on sputtered NbN films, up to a fluence of 10/sup 23/ m/sup -2/ (E > 0.1 MeV), which prove that NbN films are also extremely radiation-hard high-field superconductors. Both the transition temperatures, T/sub c/, and the normal state resistivities show only small changes with neutron fluence. Concerning the critical current densities, j/sub c/, degradations by as much as 30% are observed at low fields, whereas in an intermediate field range (11-15 T) virtually no change of j/sub c/ and at high fields near 20 T even an increase of j/sub c/ are found. The latter observation is ascribed to a radiation-induced increase of the upper critical field, B/sub c2/, and to the occurence of peak effects near B/sub c2/

  2. Tritium extraction from neutron-irradiated lithium aluminate

    International Nuclear Information System (INIS)

    Garcia H, F.

    1995-01-01

    Lithium aluminate is being strongly considered as a breeder material because of its thermophysical, chemical and mechanical stability at high temperatures and its favorable irradiation behavior. Furthermore, it is compatible with other blanket and structural materials. In this work, the effects of calcination temperature during preparation, extraction temperature and sweep gas composition were observed. Lithium aluminate prepared by four different methods, was neutron irradiated for 30 minutes at a flux of 10 12 -10 13 n/cm 2 s in the TRIGA Mark III reactor at Salazar, Mexico; and the tritium extraction rate was measured. Calcination temperature do not affect the tritium extraction rate. However, using high calcination temperature, gamma lithium aluminate was formed. The tritium extraction at 600 Centigrade degrees was lower than at 800 Centigrade degrees and the tritium amount extracted by distillation of the solid sample was higher. The sweep gas composition showed that tritium extraction was less with Ar plus 0.5 % H 2 that with Ar plus 0.1 % H 2 . This result was contrary to expected, where the tritium extraction rate could be higher when hydrogen is added to the sweep gas. Probably this effect could be attributed to the gas purity. (Author)

  3. The effect of neutron irradiation on silicon carbide fibers

    International Nuclear Information System (INIS)

    Newsome, G.A.

    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

  4. ANITA-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications

    International Nuclear Information System (INIS)

    Cepraga, Dan-Gabriel

    2000-01-01

    1 - Description of program or function: ANITA-2000 is a code package for the activation characterisation of materials exposed to neutrons in fusion machines. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation, continuous or stepwise. It provides activity, atomic density, decay heat, biological hazard, clearance index and gamma-ray source spectra at shut down and for different cooling times. An interactive utility module, MODBIN, to produce the neutron activation cross sections libraries in the required binary ANITA-4M Format, is also included. The GRANITA interactive module may plot activation parameters as a function of the cooling time. The main improvements include: -the number of irradiation time intervals has been increased to 2000; -different neutron wall loading can be used for each burn time interval; -the photon source calculation in the 18 energy group Scale structure has been added; -the clearance index can be calculated. In addition the code language has been standardized to Fortran '95 - by maintaining the backward compatibility (except for the time/date routines) - so as the same code package can be compiled and run on Unix environment and on PC, both under DOS-Windows and under Linux. 2 - Methods: The mathematical solution of the problem is given in analytical form using recurrence relations. Double precision arithmetic is used

  5. Effect of neutron irradiation on select MAX phases

    International Nuclear Information System (INIS)

    Tallman, Darin J.; Hoffman, Elizabeth N.; Caspi, El’ad N.; Garcia-Diaz, Brenda L.; Kohse, Gordon; Sindelar, Robert L.; Barsoum, Michel W.

    2015-01-01

    Herein we report on the effect of neutron irradiation – of up to 0.1 displacements per atom at 360(20) °C or 695(25) °C – on polycrystalline samples of Ti 3 AlC 2 , Ti 2 AlC, Ti 3 SiC 2 and Ti 2 AlN. Rietveld refinement of X-ray diffraction patterns of the irradiated samples showed irradiation-enhanced dissociation into TiC of the Ti 3 AlC 2 and Ti 3 SiC 2 phases, most prominently in the former. Ti 2 AlN also showed an increase in TiN content, as well as Ti 4 AlN 3 after irradiation. In contrast, Ti 2 AlC was quite stable under these irradiation conditions. Dislocation loops are seen to form in Ti 2 AlC and Ti 3 AlC 2 after irradiation at 360(20) °C. The room temperature electrical resistivity of all samples increased by an order of magnitude after irradiation at 360(20) °C, but only by 25% after 695(25) °C, providing evidence for the MAX phases’ dynamic recovery at temperatures as low at 695(25) °C. Based on these preliminary results, it appears that Ti 2 AlC and Ti 3 SiC 2 are the more promising materials for high-temperature nuclear applications

  6. 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.

  7. Neutron-Irradiated Samples as Test Materials for MPEX

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Rapp, Juergen

    2015-01-01

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility

  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. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  10. Effect of high concentration of molybdenum on the structure and properties of niobium

    International Nuclear Information System (INIS)

    Trekina, M.I.; Vasil'eva, E.V.

    1989-01-01

    The effect of alloying of 20,25 and 30 % Mo on the structure and mechanical properties of niobium is studied. It is shown that niobium alloying with molybdenum in the studied concentration leads to grain grinding, which increases with the molybdenum content growth in the alloy. The effective energy values of recrystallization activation of the studied niobium and molybdenum alloys are determined. The high hardness level at some plasticity and deformability of niobium alloy with 20 % Mo is established

  11. Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

    International Nuclear Information System (INIS)

    Scibetta, M.; Chaouadi, R.; Puzzolante, J.L.

    1999-10-01

    Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm 2 (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain

  12. Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo-5%Re. Analysis of results performed in the frame of the NET task PDS 1.4

    Energy Technology Data Exchange (ETDEWEB)

    Scibetta, M.; Chaouadi, R.; Puzzolante, J.L

    1999-10-01

    Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are considered amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation. This report aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo-5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40 and 450 degrees Celsius up to a fast neutron fluence of 3.5 1020 n/cm{sup 2} (0.2 dpa). Fracture toughness tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain.

  13. 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.

  14. Effects of molybdenum content on the structure and mechanical properties of as-cast Ti-10Zr-based alloys for biomedical applications

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Wen-Fu, E-mail: fujiiwfho@yahoo.com.tw [Department of Materials Science and Engineering, Da-Yeh University, Changhua, Taiwan (China); Wu, Shih-Ching; Hsu, Shih-Kuang [Department of Dental Technology and Materials Science, Central Taiwan University of Science and Technology, Taiwan (China); Institute of Biomedical Engineering and Materials Science, Central Taiwan University of Science and Technology, Taiwan (China); Li, Yu-Chi [Department of Materials Science and Engineering, Da-Yeh University, Changhua, Taiwan (China); Hsu, Hsueh-Chuan, E-mail: hchsu@ctust.edu.tw [Department of Dental Technology and Materials Science, Central Taiwan University of Science and Technology, Taiwan (China); Institute of Biomedical Engineering and Materials Science, Central Taiwan University of Science and Technology, Taiwan (China)

    2012-04-01

    The effects of molybdenum on the structure and mechanical properties of a Ti-10Zr-based system were studied with an emphasis on improving the strength/modulus ratio. Commercially pure titanium (c.p. Ti) was used as a control. As-cast Ti-10Zr and a series of Ti-10Zr-xMo (x = 1, 3, 5, 7.5, 10, 12.5, 15, 17.5 and 20 wt.%) alloys prepared using a commercial arc-melting vacuum pressure casting system were investigated. X-ray diffraction (XRD) for phase analysis was conducted with a diffractometer. Three-point bending tests were performed to evaluate the mechanical properties of all specimens. The experimental results indicated that these alloys had different structures and mechanical properties when various amounts of Mo were added. The as-cast Ti-10Zr has a hexagonal {alpha} Prime phase, and when 1 wt.% Mo was introduced into the Ti-10Zr alloy, the structure remained essentially unchanged. However, with 3 or 5 wt.%, the martensitic {alpha} Double-Prime structure was found. When increased to 7.5 wt.% or greater, retention of the metastable {beta} phase began. The {omega} phase was observed only in the Ti-10Zr-7.5Mo alloy. Among all Ti-10Zr-xMo alloys, the {alpha} Double-Prime -phase Ti-10Zr-5Mo alloy had the lowest elastic modulus. It is noteworthy that all the Ti-10Zr and Ti-10Zr-xMo alloys had good ductility. In addition, the Ti-10Zr-5Mo and Ti-10Zr-12.5Mo alloys exhibited higher bending strength/modulus ratios at 20.1 and 20.4, respectively. Furthermore, the elastically recoverable angles of these two alloys (26.4 Degree-Sign and 24.6 Degree-Sign , respectively) were much greater than those of c.p. Ti (2.7 Degree-Sign ). Given the importance of these properties for implant materials, the low modulus, excellent elastic recovery capability and high strength/modulus ratio of {alpha} Double-Prime phase Ti-10Zr-5Mo and {beta} phase Ti-10Zr-12.5Mo alloys appear to make them promising candidates. - Highlights: Black-Right-Pointing-Pointer The effects of Mo on the structure

  15. Tensile behavior of EUROFER ODS steel after neutron irradiation up to 16.3 dpa between 250 and 450 °C

    International Nuclear Information System (INIS)

    Materna-Morris, Edeltraud; Lindau, Rainer; Schneider, Hans-Christian; Möslang, Anton

    2015-01-01

    Highlights: • The first 9%CrWVTa steel (0.5% Y_2O_3), EUROFER ODS HIP, have been neutron irradiated up to 16.3 dpa, between 250 and 450 °C, in the High Flux Reactor (HFR). • After post-irradiation tensile tests, there was not any increase of the upper yield strength or strain localization after irradiation which is typical of RAFM steels. • Initially higher yield strength, R_p_0_._2, and distinctive tensile strength, R_m, of EUROFER ODS HIP compared to EUROFER97 steel. • These values increased due to the neutron irradiation at lower irradiation temperatures. - Abstract: During the development of structural material for future fusion reactors, a 50 kg heat of reduced-activation ferritic-martensitic 9%CrWVTa steel with nanoscaled Y_2O_3-particles, EUROFER97 ODS HIP, was produced using powder metallurgy fabrication technology. This first batch of EUROFER97 ODS HIP and, for comparison, the steel EUROFER97 were prepared for a post-irradiation tensile test program. During neutron irradiation in the HFR (High Flux Reactor, The Netherlands), an accumulated dose of up to 16.3 dpa was reached for 771 days at full power, with the irradiation temperature ranging between 250 and 450 °C. During the post-examinations, all specimens showed the highest tensile strength at lower irradiation temperatures between 250 and 350 °C. However, ODS-alloy and steel were found to clearly differ in the mechanical behavior, which could be documented by fully instrumented tensile tests. In the un-irradiated state, tensile strength of the ODS-alloy already was increased considerably by about 60% compared to the steel. Strengthening was further increased by another 20% after neutron irradiation, but with a much better ductility than observed in the steel. The typical irradiation-induced strain localization of EUROFER97 or RAFM steels could not be observed in the EUROFER97 ODS HIP alloy.

  16. Effect of phase instabilities on the correlation of nickel ion and neutron irradiation swelling in solution annealed 316 stainless steel

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.; Lee, E.H.; Sklad, P.S.

    1979-01-01

    Annealed 316 stainless steel specimens were neutron irradiated to establish steady-state microstructures and then subjected to further high temperature irradiations with 4 MeV Ni ions. It is shown that void growth under neutron irradiation is simulated in ion irradiations carried out at approx. 180 0 C above reactor temperature. However, the precipitate microstructure developed during neutron irradiation is unstable during subsequent ion irradiation. As a result, the relative swelling rates at various reactor temperatures are not simulated correctly

  17. 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)

  18. Study of damages by neutron irradiation in lithium aluminates

    International Nuclear Information System (INIS)

    Palacios G, O.

    1999-01-01

    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 (≅ 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 γ -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)

  19. Standard specification for cobalt-chromium-nickel-molybdenum-tungsten alloy (UNS R31233) plate, sheet and strip. ASTM standard

    International Nuclear Information System (INIS)

    1998-09-01

    This specification is under the jurisdiction of ASTM Committee B-2 on Nonferrous Metals and Alloys and is the direct responsibility of Subcommittee B02.07 on Refined Nickel and Cobalt, and Alloys Containing Nickel or Cobalt or Both as Principal Constituents. Current edition approved Apr. 10, 1998 and published September 1998. Originally published as B 818-91. Last previous edition was B 818-93

  20. Application of thermoelectricity to NDE of thermally aged cast duplex stainless steels and neutron irradiated ferritic steels

    International Nuclear Information System (INIS)

    Coste, J.F.; Leborgne, J.M.; Massoud, J.P.; Grisot, O.; Miloudi, S.

    1997-10-01

    The thermoelectric power (TEP) of an alloy depends mainly on its temperature, its chemical composition and its atomic arrangement. The TEP measurement technique is used in order to study and follow two degradation phenomena affecting some components of the primary loop of Pressurized Water Reactors (PWR). The first degradation phenomenon is the thermal aging of cast duplex stainless steel components. The de-mixing of the ferritic Fe-Cr-Ni slid solution is responsible for the decreasing of the mechanical characteristics. Laboratory studies have shown the sensitivity of TEP to the de-mixing phenomenon. TEP increases linearly with the ferrite content and with and Arrhenius-type aging parameter depending on time, temperature and activation energy. TEP is also correlated to mechanic characteristics. The second degradation phenomenon is the aging of ferritic steels due to neutron irradiation at about 290 deg C. In this case, the degradation mechanism is the formation of clusters of solute atoms and/or copper rich precipitates that causes the hardening of the material. As a first approach, a study of binary Fe-Cu alloys irradiated by electrons at 288 deg C has revealed the possibility of following the copper depletion of the ferritic matrix. Moreover, the recovery of the mechanical properties of the alloy by annealing can be monitored. Finally, a correlation between Vickers hardness and TEP has been established. (author)

  1. Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels; Proprietes des alliages uranium-molybdene de faibles teneurs utilisables comme materiaux combustibles

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

    Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the {gamma} structure, - cooling rate at the transformation points, - whether or not the intermediate {gamma} {yields} {beta} transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram {alpha} + {gamma}; {beta} + {gamma} the effects of the morphology (in particular the two types of {alpha} pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the {gamma} structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors) [French] Dans ce rapport

  2. Effects of La2O3 Content and Rolling on Microstructure and Mechanical Properties of ODS Molybdenum Alloys

    Science.gov (United States)

    Ma, Jingling; Li, Wuhui; Wang, Guangxin; Li, Yaqiong; Guo, Hongbo; Zhao, Zeliang; Li, Wei

    2017-10-01

    In order to study the effects of La2O3 content and rolling on microstructure and mechanical properties of Mo-La2O3 alloys, Mo-0.5% (1%) La2O3 alloys were prepared by liquid-solid doping technique, subsequently rolled either by a single-direction rolling or a cross-rolling. As a result, three different materials were prepared for this study. After being annealed at 1800 °C, the single-directionally rolled Mo-1% La2O3 alloy shows the best mechanical properties in terms of strength, hardness, and sagging deformation among the three materials. This is attributed to the observation that the alloy is only recovered with a microstructure of subgrains and dislocations. The single-directionally rolled Mo-0.5% La2O3 exhibits the worst mechanical property among the three materials. In this material, coarse grains, but no subgrains and dislocations, can be observed after annealing, indicating that it is fully recrystallized. For the cross-rolled Mo-1% La2O3 alloy, grains of dispersed sizes, but no dislocations, are visible after annealing, implying that this alloy is partially recrystallized. Accordingly, the mechanical property of this material is in between the other two materials. Thus, the mechanical properties of the three materials can be well understood based on OM, SEM, and TEM results. Overall, the single-directionally rolled Mo-1% La2O3 alloy possesses good mechanical properties and is more suitable for high-temperature applications.

  3. Colony form variation of Bacillus pumilus E601 after cultured and neutron irradiation

    International Nuclear Information System (INIS)

    Chen Xiaoming; Wei Baoli; Zhang Jianguo

    2008-01-01

    The distribution of two colony forms of Bacillus pumilus E601 and the effect of neutron irradiation on the colony form were reported. The translucent and opaque colonies were cultured several generations to observe the proportion of two form colonies. The spores of opaque colony were irradiated at 80, 800 and 2000 Gy of fast neutron from CFBR-II pulse pile, and the survivors of opaque colony were irradiated again at the same doses. The results showed that: (1) Bacillus pumilus E601 observed two types of colony form: translucent and opaque colony; (2) the translucent colony could produce both translucent and opaque colonies in equal, while the opaque colony couldn't produce translucent colony generally; (3) neutron irradiation could affect the colony form distribution. The ratio of survival translucent colony was increased with the increase of the first neutron irradiation doses, and the second neutron irradiation also increased the ratio of translucent colony. It was concluded that the instability of translucent colony was the main reason to produce two colony forms of Bacillus pumilus E601. The strain of translucent colony had a stronger ability to resist neutron irradiation than the opaque colony. (authors)

  4. Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

    International Nuclear Information System (INIS)

    Oya, Yasuhisa; Kobayashi, Makoto; Okuno, Kenji; Shimada, Masashi; Calderoni, Pattrick; Oda, Takuji; Hara, Masanori; Hatano, Yuji; Watanabe, Hideo

    2014-01-01

    The behavior of D retentions for Fe 2+ irradiated tungsten with the damage of 0.025-3 dpa was compared with that for neutron irradiated tungsten with 0.025 dpa. The D 2 TDS spectra for Fe 2+ irradiated tungsten consisted of two desorption stages at 450 K and 550 K although that for neutron irradiated tungsten was composed of three stages and addition desorption stage was found around 750 K. The desorption rate of major desorption stage at 550 K increased as the number of dpa by Fe 2+ irradiation increased. In addition, the first desorption stage at 450 K was only found for the damaged samples, indicating that the second stage would be based on intrinsic defects or vacancy produced by Fe 2+ irradiation and the first stage should be the accumulation of D in mono vacancy leading to the lower activation energy, where the dislocation loop and vacancy was produced. The third one was only found for the neutron irradiation, showing the D trapping by void or vacancy cluster and the diffusion effect is also contributed due to high FWHM of TDS spectrum. It can be said that the D 2 TDS spectra for Fe 2+ -irradiated tungsten could not represent that for neutron-irradiated one, showing that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten has a difference from that for ion-irradiated one. (author)

  5. Microchemical evolution of neutron-irradiated stainless steel

    International Nuclear Information System (INIS)

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

    1980-04-01

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

  6. 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

  7. 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

  8. 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

  9. Primary defect production by high energy displacement cascades in molybdenum

    Energy Technology Data Exchange (ETDEWEB)

    Selby, Aaron P. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Xu, Donghua, E-mail: xudh@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Juslin, Niklas; Capps, Nathan A. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Oak Ridge National Laboratory, P.O. Box 2008, MS6003, Oak Ridge, TN 37831 (United States)

    2013-06-15

    We report molecular dynamics simulations of primary damage in molybdenum produced by high energy displacement cascades on the femto- to pico-second and Angstrom to nanometer scales. Clustering directly occurred for both interstitials and vacancies in the 1–50 keV cascade energy range explored. Point defect survival efficiency and partitioning probabilities into different sized clusters were quantified. The results will provide an important reference for kinetic models to describe the microstructural evolution in Mo under ion or neutron irradiations over much longer time and length scales.

  10. Conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at SINQ

    International Nuclear Information System (INIS)

    Zanini, L.; Baluc, N.; Simone, A. De; Eichler, R.; Joray, S.; Manfrin, E.; Pouchon, M.; Rabaioli, S.; Schumann, D.; Welte, J.; Zhernosekov, K.

    2011-12-01

    This comprehensive, illustrated report by the Paul Scherrer Institute PSI in Switzerland documents the proposals concerning the conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at the PSI's Swiss Spallation Neutron Source SINQ facility. The need for fast neutron irradiation is discussed and the possibility of using SINQ as a fast neutron irradiation facility is considered. The production of isotopes, tracers and medical isotopes is discussed, as are fission and fusion reactor technologies. The characteristics of the neutron spectrum in SINQ are discussed. The neutronic and radioprotection calculations for an irradiation station at SINQ are looked at in detail and extensive examples of work done and results obtained are presented and discussed. Radioprotection issues are also looked at. Further contributions in the report cover the hot/cold irradiation station in the SINQ target. An appendix provides detailed drawings of the facility's pneumatic delivery system

  11. 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(......, as predicted by molecular dynamics simulations. (C) 2008 Elsevier B.V. All rights reserved....... at a very low-dose of similar to 10(-4) dpa. The average size of the cavities did not change significantly with dose, in contrast to neutron-irradiated bcc Fe where cavity sizes increased with increasing dose. It is suggested that the in-cascade vacancy clustering may be significant in neutron-irradiated Mo...

  12. Effects of neutron irradiation on red blood cell labeling with technetium-99m

    International Nuclear Information System (INIS)

    Eng, R.R.; Conklin, J.J.; Grissom, M.P.

    1982-01-01

    The effects of in vivo and in vitro neutron irradiation on red blood cell radiolabeling with technetium-99m (Tc-99m) were studied. Blood from three dogs was irradiated with neutrons (725 rads, free in air dose) followed by radiolabeling with Tc-99m. The three dogs were subsequently whole body, neutron irradiated (250 rads, midline dose); and blood samples were drawn for radiolabeling at 24, 48, 72 and 96 hours post-irradiation. Blood from three control dogs was also drawn and radiolabeled on each day for comparison. The results show that there were no significant differences between the radiolabeling capacities of in vivo or in vitro neutron irradiated and control RBCs

  13. Neutron irradiation effects on magnetic properties of Fe-based ferromagnetic metallic glasses

    International Nuclear Information System (INIS)

    Miglierini, M.; Nasu, Saburo; Skorvanek, I.; Sitek, J.

    1992-01-01

    Transmission 57 Fe Moessbauer spectroscopy, J-H quasistatic hysteresis loop and AC susceptibility measurements are used to study effects of neutron irradiation on magnetic properties of Fe-based-ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transition from the ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. The presence of Ni in the samples reduces the effects of radiation damage as revealed also from position lifetime data. Possible sources of a radiation damage are discussed using the results of γ-ray spectroscopy. (author)

  14. Neutron irradiation effects on magnetic properties of Fe-based ferromagnetic metallic glasses

    Energy Technology Data Exchange (ETDEWEB)

    Miglierini, M.; Nasu, Saburo (Osaka Univ., Toyonaka (Japan). Faculty of Science); Skorvanek, I.; Sitek, J.

    1992-04-01

    Transmission {sup 57}Fe Moessbauer spectroscopy, J-H quasistatic hysteresis loop and AC susceptibility measurements are used to study effects of neutron irradiation on magnetic properties of Fe-based-ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transition from the ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. The presence of Ni in the samples reduces the effects of radiation damage as revealed also from position lifetime data. Possible sources of a radiation damage are discussed using the results of {gamma}-ray spectroscopy. (author).

  15. Protons in neutron-irradiated and thermochemically reduced MgO crystals doped with lithium impurities

    International Nuclear Information System (INIS)

    Gonzalez, R.; Pareja, R.; Chen, Y.

    1992-01-01

    H - (hydride) ions have been observed in lithium-doped MgO crystals which have been neutron irradiated or thermochemically reduced (TCR). Infrared-absorption measurements have been used to identify the local modes of the H - ions in these crystals. The concentration of the H - ions in the neutron-irradiated crystals is found to be far less than that found in the TCR crystals. The thermal stability of H - and oxygen vacancies in both oxidizing and reducing atmospheres are investigated. The emergence of sharp structures due to OH - ions is attributed to the displacements of substitutional Li + ions, leaving behind unperturbed OH - ions, via a mechanism of rapid radiation-induced diffusion during irradiation in a reactor. Results of neutron-irradiated MgO:Li, which had previously been oxidized at high temperature, are also presented

  16. Positron annihilation spectroscopy of vacancy aggregates in neutron-irradiated MgO crystals

    International Nuclear Information System (INIS)

    Pareja, R.; De La Cruz, R.M.; Gonzalez, R.; Chen, Y.; Department of Energy, Washington, DC

    1992-01-01

    Positron annihilation measurements in neutron-irradiated MgO crystals show that the positron lifetime is shorter than in as-grown crystals, suggesting that most of the defects produced by neutron irradiations are positively charged. The concentration of the neutral anion vacancy (possibly also the neutral anion divacancy) is estimated to be no more than ∼ 10 16 cm -3 for samples irradiated to a dose of 10 17 to 10 19 n cm -2 . Annealing experiments on the neutron-irradiated crystals show a significant increase in the positron lifetime after anneals at 900 K. The increase is attributed to positron trapping by anion-vacancy aggregates. A lifetime of (284±15)ps is tentatively assigned to positrons trapped in these aggregates. (Author)

  17. Micro-Raman and photoluminescence studies of neutron-irradiated gallium nitride epilayers

    International Nuclear Information System (INIS)

    Wang, R.X.; Xu, S.J.; Fung, S.; Beling, C.D.; Wang, K.; Li, S.; Wei, Z.F.; Zhou, T.J.; Zhang, J.D.; Huang Ying; Gong, M.

    2005-01-01

    GaN epilayers grown on sapphire substrate were irradiated with various dosages of neutrons and were characterized using Micro-Raman and photoluminescence. It was found that the A 1 (LO) peak in the Raman spectra clearly shifted with neutron irradiation dosage. Careful curve fitting of the Raman data was carried out to obtain the carrier concentration which was found to vary with the neutron irradiation dosage. The variation of the full width at half maximum height of the photoluminescence was consistent with the Raman results. The neutron irradiation-induced structural defects (likely to be Ge Ga ) give rise to carrier trap centers which are responsible for the observed reduction in carrier concentration of the irradiated GaN

  18. The effect of cooling rate from the γ-phase on the strain-rate sensitivity of a uranium 2 sup(w)/o molybdenum alloy

    International Nuclear Information System (INIS)

    Boyd, G.A.C.; Harding, J.

    1983-01-01

    Tensile tests have been performed at strain rates from 10 -4 to about 2000/s and temperatures from -150 deg C to +250 deg C on a uranium 2 w/o molybdenum alloy which had been aged for 2 hours at 500 deg C after a fast gas cool from the γ-phase at a controlled rate of 40 deg C/minute. The results are compared with those for standard as-extruded material which had received the same aging treatment. Stress-strain curves are presented and the effect of strain rate and temperature on the flow stress, the ultimate tensile stress and the elongation to fracture is determined. A thorough structural characterisation of the specimen materials, using X-ray analysis and scanning and transmission electron microscopy, allows the different mechanical responses to be related to the corresponding microstructural state of the material. Flow stress data at different temperatures and strain rates are analysed in terms of the theory of thermally-activated flow and estimates made of the various activation parameters. (author)

  19. A comparison of torque expression between stainless steel, titanium molybdenum alloy, and copper nickel titanium wires in metallic self-ligating brackets.

    Science.gov (United States)

    Archambault, Amy; Major, Thomas W; Carey, Jason P; Heo, Giseon; Badawi, Hisham; Major, Paul W

    2010-09-01

    The force moment providing rotation of the tooth around the x-axis (buccal-lingual) is referred to as torque expression in orthodontic literature. Many factors affect torque expression, including the wire material characteristics. This investigation aims to provide an experimental study into and comparison of the torque expression between wire types. With a worm-gear-driven torquing apparatus, wire was torqued while a bracket mounted on a six-axis load cell was engaged. Three 0.019 x 0.0195 inch wire (stainless steel, titanium molybdenum alloy [TMA], copper nickel titanium [CuNiTi]), and three 0.022 inch slot bracket combinations (Damon 3MX, In-Ovation-R, SPEED) were compared. At low twist angles (wires were not statistically significant. At twist angles over 24 degrees, stainless steel wire yielded 1.5 to 2 times the torque expression of TMA and 2.5 to 3 times that of nickel titanium (NiTi). At high angles of torsion (over 40 degrees) with a stiff wire material, loss of linear torque expression sometimes occurred. Stainless steel has the largest torque expression, followed by TMA and then NiTi.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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.

  1. Irradiation-induced microstructural changes in alloy X-750

    International Nuclear Information System (INIS)

    Kenik, E.A.

    1997-01-01

    Alloy X-750 is a nickel base alloy that is often used in nuclear power systems for it's excellent corrosion resistance and mechanical properties. The present study examines the microstructure and composition profiles in a heat of Alloy X-750 before and after neutron irradiation

  2. Inhomogeneous strain induced by fast neutron irradiation in NaKSO4 crystals

    International Nuclear Information System (INIS)

    Kandil, S.H.; El Gamal, M.A.; El Khatib, A.; El Wahidy, E.F.

    1987-06-01

    The effect of fast neutron irradiation on the thermal properties of NaKSO 4 crystals was studied around the phase transition temperature T c =453 K. The thermal expansion coefficient as well as the phase transition temperature were found to be dependent upon the irradiation dose. The specific heat, C p , showed multiple peaks in the phase transition temperature region. An explanation of this behaviour was based on the induced inhomogeneous strain in the crystal casued by the neutron irradiation process. (author). 10 refs, 3 figs

  3. Surface damage in the small intestine of the mouse after X - or neutron irradiation

    International Nuclear Information System (INIS)

    Hamlet, R.; Carr, K.E.; Nias, A.H.; Watt, C.

    1981-01-01

    Damage after X-irradiation includes lateral villous collapse, progressing after 3 - 5 days to villi which sometimes show signs of vertical collapse. After neutron irradiation vertical villous collapse is established earlier, with less swelling of villous tips. It seems, therefore, that at radiobiologically equivalent doses, neutron and X-irradiation produce different levels of surface damage, with neutron irradiation being the more destructive. Early villous tip damage may perhaps be due to disruption of susceptible cells already at the extrusion zone, or to stromal damage

  4. Superconducting and normal state properties of carbon doped and neutron irradiated MgB2

    International Nuclear Information System (INIS)

    Wilke, R.H.T.; Samuely, P.; Szabo, P.; Holanova, Z.; Bud'ko, S.L.; Canfield, P.C.; Finnemore, D.K.

    2007-01-01

    Current research in MgB 2 focuses on the effects various types of perturbations have on the superconducting properties of this novel two-gap superconductor. In this article we summarize the effects of carbon doping and neutron irradiation in bulk MgB 2 . Low levels of carbon doping and light neutron irradiation result in significant enhancements in H c2 . At high fluences, where superconductivity is nearly fully suppressed, superconductivity can be restored through post exposure annealing. However, this results in a change in the interdependencies of the normal state and superconducting properties (ρ 0 , T c , H c2 ), with little or no enhancement in H c2

  5. 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

    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....../II were elevated in liver tissue. The present data do not tell whether kidney and liver are the primary target organs or what possible toxicological effect the different metal ions might have, but they show that metal ions are liberated from CoCrMo alloys that are not subjected to mechanical wear...... and that they accumulate in liver and kidney tissue. That the liberated metal ions affect the tissues is supported by an up-regulation of the detoxifying/pacifying metalloprotein I/II in the liver. Udgivelsesdato: 2007-Dec...

  6. 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

  7. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  8. Scanning tunneling spectroscopy on neutron irradiated MgB2 thin films

    International Nuclear Information System (INIS)

    Di Capua, Roberto; Salluzzo, Marco; Vaglio, Ruggero; Ferdeghini, Carlo; Ferrando, Valeria; Putti, Marina; Xi Xiaoxing; Aebersold, Hans U.

    2007-01-01

    Neutron irradiation was performed on MgB 2 thin films grown by hybrid physical chemical vapor deposition. Samples irradiated with different neutron fluences, having different critical temperatures, were studied by scanning tunneling spectroscopy in order to investigate the effect of the introduced disorder on the superconducting and spectroscopic properties. A monotonic increase of the π gap with increasing disorder was found

  9. Scanning tunneling spectroscopy on neutron irradiated MgB{sub 2} thin films

    Energy Technology Data Exchange (ETDEWEB)

    Di Capua, Roberto [University of Napoli and CNR-INFM/Coherentia, Via Cinthia, Naples I-80126 (Italy)], E-mail: rdicapua@na.infn.it; Salluzzo, Marco; Vaglio, Ruggero [University of Napoli and CNR-INFM/Coherentia, Via Cinthia, Naples I-80126 (Italy); Ferdeghini, Carlo [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Ferrando, Valeria [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Pennsylvania State University, University Park, PA 16802 (United States); Putti, Marina [CNR-INFM/LAMIA, Via Dodecaneso 33, Genova I-16146 (Italy); Xi Xiaoxing [Pennsylvania State University, University Park, PA 16802 (United States); Aebersold, Hans U. [Paul Scherrer Institut, Villigen CH-5232 (Switzerland)

    2007-09-01

    Neutron irradiation was performed on MgB{sub 2} thin films grown by hybrid physical chemical vapor deposition. Samples irradiated with different neutron fluences, having different critical temperatures, were studied by scanning tunneling spectroscopy in order to investigate the effect of the introduced disorder on the superconducting and spectroscopic properties. A monotonic increase of the {pi} gap with increasing disorder was found.

  10. Physical model of evolution of oxygen subsystem of PLZT-ceramics at neutron irradiation and annealing

    CERN Document Server

    Kulikov, D V; Trushin, Y V; Veber, K V; Khumer, K; Bitner, R; Shternberg, A R

    2001-01-01

    The physical model of evolution of the oxygen subsystem defects of the ferroelectric PLZT-ceramics by the neutron irradiation and isochrone annealing is proposed. The model accounts for the effect the lanthanum content on the material properties. The changes in the oxygen vacancies concentration, calculated by the proposed model, agree well with the polarization experimental behavior by the irradiated material annealing

  11. Chemical consequences of the neutron irradiation of ionic antimony oxides and Fe Sb2O4

    Energy Technology Data Exchange (ETDEWEB)

    Facetti, J F [Asuncion Nacional Univ. (Paraguay). Inst. de Ciencias

    1970-01-01

    The chemical consequences fo the neutron irradiation of ionic antimony oxides and Fe Sb2O4 are studied. The nature of the Sb-O2 bond effects the yield of SbV the higher the yield the more covalent the bond. In addition, the Fe Sb2O4 obeys the Maddock's rule.

  12. Neutron irradiation effects on in situ Nb3Sn superconducting wires

    International Nuclear Information System (INIS)

    Hirano, Y.; Fukumoto, M.; Kodaka, H.; Nishijima, S.; Okada, T.; Yoshida, H.

    1985-01-01

    Three types of ''in situ'' Nb 3 Sn conductors have been studied to determine the detrimental effects of neutron irradiation on critical temperature, critical current and AC losses. It was found that at fluences of 10 18 n/cm 2 , the critical temperature is degraded by approximately 10%. Degradation of AC loss is discussed in comparison with that of critical current density

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

    International Nuclear Information System (INIS)

    Kwon, Tae-Hyun; Park, Jiho; Kim, Jun Yeon; Kim, HyungTae; Park, Kyoungsoo; Kim, Sang-Ho

    2015-01-01

    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 19 n/cm 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

  14. Analysis of microstress in neutron irradiated polyester fibre by X-ray ...

    Indian Academy of Sciences (India)

    Unknown

    Abstract. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  16. Influence of interstitial impurity atoms on point defect relaxation in neutron irradiated iron

    International Nuclear Information System (INIS)

    Weller, M.; Diehl, J.

    1975-01-01

    The aim of the investigation was to study the combined influence of neutron irradiation and interstitial impurities on the low temperature internal friction peaks as well as on those appearing at higher temperatures after annealing, in a more systematic manner, using irradiations at [de

  17. Report of the Working Group on low-temperature neutron irradiation

    International Nuclear Information System (INIS)

    1982-07-01

    This report summarizes deliberations at a Working Group meeting sponsored by the Department of Energy, Division of Materials Sciences for the purpose of: (1) assessing the need for maintaining a low temperature neutron irradiation program in the United States; and (2) recommending a course of action based on this assessment

  18. Investigation of hydrogen isotopes interaction processes with lithium under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Zaurbekova, Zhanna, E-mail: zaurbekova@nnc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Skakov, Mazhyn; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Tazhibayeva, Irina; Baklanov, Viktor; Barsukov, Nikolay [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Chikhray, Yevgen [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan)

    2016-11-01

    Highlights: • The experiments on study of helium and tritium generation and release processes under neutron irradiation from lithium saturated with deuterium are described in paper. ​ • The values of relative tritium and helium yield from lithium sample at different levels of neutron irradiation is calculated. • It was concluded that the main affecting process on tritium release from lithium is its interaction with lithium atoms with formation of lithium tritide. - Abstract: The paper describes the experiments on study of helium and tritium generation and release processes from lithium saturated with deuterium under neutron irradiation (in temperature range from 473 to 773 K). The diagrams of two reactor experiments show the time dependences of helium, DT, T{sub 2}, and tritium water partial pressures changes in experimental chamber with investigated lithium sample. According to experimental results, the values of relative tritium and helium yield from lithium sample at different levels of neutron irradiation were calculated. The time dependences of relative tritium and helium yield from lithium sample were plotted. It was concluded that the main affecting process on tritium release from lithium is its interaction with lithium atoms with formation of lithium tritide.

  19. DT fusion neutron irradiation of ORNL magnesium oxide crystals and BNL--LASL superconductor wires

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1978-01-01

    The DT fusion neutron irradiation of two ORNL magnesium oxide crystals and eleven BNL-LASL superconductor wires is described. The sample position and neutron dose record are given. The maximum neutron fluence on any sample was 2.16 x 10 16 neutrons/cm 2

  20. Time factor of BSH from intravenous infusion to neutron irradiation for BNCT in patients with glioblastoma

    International Nuclear Information System (INIS)

    Kageji, T.; Nagahiro, S.; Kitamura, K.; Nakagawa, Y.; Hatanaka, H.; Haritz, D.; Grochulla, F.; Haselsberger, K.; Gabel, D.

    2000-01-01

    The present report evaluates the time factor of BSH from infusion to irradiation in patients with glioblastoma as a cooperative study in Europe and Japan. For BNCT with BSH after intravenous infusion, this work confirms that the planned neutron irradiation after intravenous BSH infusion appears to be optimal around 12-19 hours after the infusion. (author)

  1. Theoretical description of the influence of neutron irradiation on viscoplastic properties of mild steel

    International Nuclear Information System (INIS)

    Pecherski, R.

    1978-01-01

    The physical bases of plastic deformation of mild steel are described. The influence of neutron irradiation on the change of mechanisms of plastic deformation is discussed in detail. Constitutive equations of viscoplasticity for irradiated mild steel are given. The problem of thickwalled viscoplastic spherical tank irradiated by neutrons is studied. (Z.R.)

  2. Impurity effects in neutron-irradiated simple oxides: Implications for fusion devices

    International Nuclear Information System (INIS)

    Gonzalez, R.; Chen, Y.; Caceres, D.; Vergara, I.

    2006-01-01

    Radiation damage induced by neutron irradiation was studied in undoped MgO crystals and in MgO doped with either iron, hydrogen or lithium impurities. The oxygen-vacancy concentration produced by irradiation increases with neutron fluence. The net production rates resulting from irradiations with 14.8 MeV neutrons are about twice those produced by fission neutrons. In nominally pure crystals, the oxygen-vacancy concentration incurred by the fission-neutron irradiation is higher in crystals with a larger number of inherent impurities (such as iron) due to trapping of interstitials by impurities. Suppression of these defects is observed in MgO:H crystals and attributed to migration of oxygen vacancies to microcavities filled with H 2 gas. In MgO:Li crystals irradiated with neutron fluences below 10 18 n/cm 2 , most of the oxygen vacancies are camouflaged as hydride ions. Nanoindentation experiments show that hardness increases with neutron fluence and is independent of the presence of lithium in the crystal. Comparison between a neutron-irradiated and a thermochemically reduced crystal containing similar concentrations of oxygen vacancies shows that 70% of the neutron-irradiation hardening is produced by interstitials, 30% by oxygen vacancies and a negligible amount by higher-order point defects

  3. DT fusion neutron irradiation of BPNL niobium nickel and 316 stainless steel at 1750C

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation at 175 0 C of 17 niobium wires, one niobium foil, 14 316 stainless steel wires, one 316 stainless steel foil, nine nickel wires, and two nickel foils from BPNL is described. The sample position, beam-on time, neutron dose record, and neutron fluence are given

  4. Neutron irradiation of sapphire for compressive strengthening. II. Physical properties changes

    Energy Technology Data Exchange (ETDEWEB)

    Regan, Thomas M. E-mail: thomas_regan@uml.edu; Harris, Daniel C. E-mail: harrisdc@navair.navy.mil; Blodgett, David W.; Baldwin, Kevin C.; Miragliotta, Joseph A.; Thomas, Michael E.; Linevsky, Milton J.; Giles, John W.; Kennedy, Thomas A.; Fatemi, Mohammad; Black, David R.; Lagerloef, K. Peter D

    2002-01-01

    Irradiation of sapphire with fast neutrons (0.8-10 MeV) at a fluence of 10{sup 22}/m{sup 2} increased the c-axis compressive strength and the c-plane biaxial flexure strength at 600 deg. C by a factor of {approx}2.5. Both effects are attributed to inhibition of r-plane twin propagation by damage clusters resulting from neutron impact. The a-plane biaxial flexure strength and four-point flexure strength in the c- and m-directions decreased by 10-23% at 600 deg. C after neutron irradiation. Neutron irradiation had little or no effect on thermal conductivity, infrared absorption, elastic constants, hardness, and fracture toughness. A featureless electron paramagnetic resonance signal at g=2.02 was correlated with the strength increase: This signal grew in amplitude with increasing neutron irradiation, which also increased the compressive strength. Annealing conditions that reversed the strengthening also annihilated the g=2.02 signal. A signal associated with a paramagnetic center containing two Al nuclei was not correlated with strength. Ultraviolet and visible color centers also were not correlated with strength in that they could be removed by annealing at temperatures that were too low to reverse the compressive strengthening effect of neutron irradiation.

  5. Fiscal year 1976 DT fusion neutron irradiations and dosimetry at the LLL rotating target neutron source

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation of 319 samples during 19 irradiation periods (beam-on time of more than 1026 hours) is described. Experiments from 24 individuals representing 11 institutions are summarized. The numbers of the UCID dosimetry reports detailing each of the irradiations are given

  6. Interdiffusion, Intrinsic Diffusion, Atomic Mobility, and Vacancy Wind Effect in γ(bcc) Uranium-Molybdenum Alloy

    Science.gov (United States)

    Huang, Ke; Keiser, Dennis D.; Sohn, Yongho

    2013-02-01

    U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning's formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

  7. Influence of Chromium-Cobalt-Molybdenum Alloy (ASTM F75) on Bone Ingrowth in an Experimental Animal Model.

    Science.gov (United States)

    Zuchuat, Jésica; Berli, Marcelo; Maldonado, Ysaí; Decco, Oscar

    2017-12-26

    Cr-Co-Mo (ASTM F75) alloy has been used in the medical environment, but its use as a rigid barrier membrane for supporting bone augmentation therapies has not been extensively investigated. In the present study, Cr-Co-Mo membranes of different heights were placed in New Zealand white, male rabbit tibiae to assess the quality and volume of new bone formation, without the use of additional factors. Animals were euthanized at 20, 30, 40, and 60 days. Bone formation was observed in all of the cases, although the tibiae implanted with the standard membranes reached an augmentation of bone volume that agreed with the density values over the timecourse. In all cases, plasmatic exudate was found under the membrane and in contact with the new bone. Histological analysis indicated the presence of a large number of chondroblasts adjacent to the inner membrane surface in the first stages, and osteoblasts and osteocytes were observed under them. The bone formation was appositional. The Cr-Co-Mo alloy provides a scaffold with an adequate microenvironment for vertical bone volume augmentation, and the physical dimensions and disposition of the membrane itself influence the new bone formation.

  8. Influence of Chromium-Cobalt-Molybdenum Alloy (ASTM F75 on Bone Ingrowth in an Experimental Animal Model

    Directory of Open Access Journals (Sweden)

    Jésica Zuchuat

    2017-12-01

    Full Text Available Cr-Co-Mo (ASTM F75 alloy has been used in the medical environment, but its use as a rigid barrier membrane for supporting bone augmentation therapies has not been extensively investigated. In the present study, Cr-Co-Mo membranes of different heights were placed in New Zealand white, male rabbit tibiae to assess the quality and volume of new bone formation, without the use of additional factors. Animals were euthanized at 20, 30, 40, and 60 days. Bone formation was observed in all of the cases, although the tibiae implanted with the standard membranes reached an augmentation of bone volume that agreed with the density values over the timecourse. In all cases, plasmatic exudate was found under the membrane and in contact with the new bone. Histological analysis indicated the presence of a large number of chondroblasts adjacent to the inner membrane surface in the first stages, and osteoblasts and osteocytes were observed under them. The bone formation was appositional. The Cr-Co-Mo alloy provides a scaffold with an adequate microenvironment for vertical bone volume augmentation, and the physical dimensions and disposition of the membrane itself influence the new bone formation.

  9. Stability of thermally induced copper precipitates under neutron irradiation

    International Nuclear Information System (INIS)

    Phythian, W.J.; Dumbill, S.; Brown, P.; Sinclair, R.

    1993-01-01

    Model Fe 1.3%Cu and Fe 1.3%Cu 1.1%Ni alloys have been thermally aged at 550 C for 2 hours (peak) and 10 hours prior to irradiation at 288 C to a dose of 5.10 22 n/m 2 . Results of a microstructural investigation using dedicated field emission gun scanning transmission electron microscopy (FEGSTEM) and small angle neutron scattering (SANS) to assess precipitate stability in the binary alloy, are presented. These data are then used to predict a hardness change as a result of copper precipitation for comparison with the measured values obtained using standard 5 kg Vickers hardness tests on the SANS samples. Implications of these data to the re-embrittlement of the RPV by subsequent copper precipitation is discussed. (authors). 16 refs., 5 figs., 5 tabs

  10. Stability of thermally induced copper precipitates under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Phythian, W J; Dumbill, S; Brown, P; Sinclair, R [AEA Technology, Harwell (United Kingdom)

    1994-12-31

    Model Fe 1.3%Cu and Fe 1.3%Cu 1.1%Ni alloys have been thermally aged at 550 C for 2 hours (peak) and 10 hours prior to irradiation at 288 C to a dose of 5.10{sup 22} n/m{sup 2}. Results of a microstructural investigation using dedicated field emission gun scanning transmission electron microscopy (FEGSTEM) and small angle neutron scattering (SANS) to assess precipitate stability in the binary alloy, are presented. These data are then used to predict a hardness change as a result of copper precipitation for comparison with the measured values obtained using standard 5 kg Vickers hardness tests on the SANS samples. Implications of these data to the re-embrittlement of the RPV by subsequent copper precipitation is discussed. (authors). 16 refs., 5 figs., 5 tabs.

  11. 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

  12. 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.)

  13. Study on preparation and properties of molybdenum alloys reinforced by nano-sized ZrO2 particles

    International Nuclear Information System (INIS)

    Cui, Chaopeng; Gao, Yimin; Zhou, Yucheng; Wei, Shizhong; Zhang, Guoshang; Zhu, Xiangwei; Guo, Songliang

    2016-01-01

    The nano-sized ZrO 2 -reinforced Mo alloy was prepared by a hydrothermal method and a subsequent powder metallurgy process. During the hydrothermal process, the nano-sized ZrO 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 2 particles, and the fine-grain sintered structure is obtained correspondingly due to hereditation. In addition to a few of nano-sized ZrO 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.)

  14. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  15. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Science.gov (United States)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  16. 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

    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......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...

  17. High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert

    2018-04-30

    Project Objective: The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiated microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations

  18. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  19. Comparison of properties and microstructures of Trefimetaux CuNiBe and Hycon 3HP TM before and after neutron irradiation. (ITER R and D Task no. T213)

    International Nuclear Information System (INIS)

    Edwards, D.J.; Eldrup, M.; Toft, P.; Singh, B.N.

    2000-07-01

    The precipitation strengthened CuNiBe alloys are among the three candidate copper alloys that are 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 Cu-A1 2 O 3 alloys. Brush-Wellman Inc. has developed an improved version of their Hycon CuNiBe alloy that has higher conductivity while maintaining a reasonable level of strength. In the present work we have investigated the physical and mechanical properties of the Hycon 3HP TM alloy both before and after neutron irradiation and have compared its microstructure and properties with the European CuNiBe candidate alloy manufactured by Trefimetaux. Tensile specimens of both alloys were irradiated in the DR-3 reactor at Risoe to displacement dose levels of up to 0.3 dpa at 100, 250 and 350 d eg C . Both alloys were tensile tested in vacuum in the unirradiated and irradiated conditions at 100, 250 and 350 d eg C and the microstructures of the alloys were investigated using transmission electron microscopy. Electrical resistivity measurements were made on tensile specimens be-fore and after irradiation; all measurements were made at 23 d eg C . Results of these investigations are presented and discussed in terms of the sensitivity of these alloys to test temperature, which becomes increasingly problematic when the irradiation and test temperature reaches 250 d eg C and above. (au)

  20. Effect of preliminary neutron irradiation on helium blistering of 0Kh16N15M3B steel

    International Nuclear Information System (INIS)

    Chernov, I.I.; Kalin, B.A.; Skorov, D.M.; Shishkin, G.N.; Ivanov, M.V.

    1982-01-01

    The method of electron microscopy has been applied to investigate the effect of preliminary neutron irradiation on the OKh16N15M3B steel blistering under irradiation by 20 keV helium ions with (1-10)x10 21 ion/m 2 doses at the temperature below 373 K. It is shown that neutron irradiation shifts critical doses of blister formation and intense scaling towards higher doses. But after the incubation period the erosion of steel preliminary neutron irradiated grows with the increase of helium ion dose above 7x10 21 ion/m 2 . Short-term heating of neutron irradiated samples during 15 min at 1173 K does not practically affect the beginning of intense scaling of the surface

  1. Deuterium trapping at vacancy clusters in electron/neutron-irradiated tungsten studied by positron annihilation spectroscopy

    Science.gov (United States)

    Toyama, T.; Ami, K.; Inoue, K.; Nagai, Y.; Sato, K.; Xu, Q.; Hatano, Y.

    2018-02-01

    Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ∼373 K and to a dose of ∼1 × 10-3 dpa) or neutron-irradiated (at 573 K to a dose of ∼0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ∼0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancy-type defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.

  2. 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.

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

    International Nuclear Information System (INIS)

    Chun, Ki Jung; Seo, Won Sook; Son, Hwa Young

    2006-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Irradiation hardening and localized deformation of neutron-irradiated α-iron single crystals

    International Nuclear Information System (INIS)

    Mughrabi, H.; Stroehle, D.; Wilkens, M.

    1981-01-01

    The early yielding behaviour of neutron-irradiated α iron single crystals orientated for single slip was investigated as a function of neutron dose. In the range of neutron doses between approx. equal to 10 18 and approx. equal to 10 19 n/cm 2 , the irradiation hardening increment was found to be almost constant. Qualitative modifications of this behaviour were observed in the case of predeformed specimens. The localized deformation of the neutron-irradiated specimens by dislocation channelling was investigated by slip-line observations, transmission electron microscopy and X-ray topography. A model of localized deformation is proposed in order to explain the development of the observed asymmetric dislocation double layers which bound the channels and transmit characteristic misorientations. (orig.)

  6. 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.

  7. Model of defect reactions and the influence of clustering in pulse-neutron-irradiated Si

    International Nuclear Information System (INIS)

    Myers, S. M.; Cooper, P. J.; Wampler, W. R.

    2008-01-01

    Transient reactions among irradiation defects, dopants, impurities, and carriers in pulse-neutron-irradiated Si were modeled taking into account the clustering of the primal defects in recoil cascades. Continuum equations describing the diffusion, field drift, and reactions of relevant species were numerically solved for a submicrometer spherical volume, within which the starting radial distributions of defects could be varied in accord with the degree of clustering. The radial profiles corresponding to neutron irradiation were chosen through pair-correlation-function analysis of vacancy and interstitial distributions obtained from the binary-collision code MARLOWE, using a spectrum of primary recoil energies computed for a fast-burst fission reactor. Model predictions of transient behavior were compared with a variety of experimental results from irradiated bulk Si, solar cells, and bipolar-junction transistors. The influence of defect clustering during neutron bombardment was further distinguished through contrast with electron irradiation, where the primal point defects are more uniformly dispersed

  8. Spallation and 14-MeV neutron irradiation of stabilized NbTi superconductors

    International Nuclear Information System (INIS)

    Hahn, P.; Brown, B.S.; Weber, H.W.; Guinan, M.W.

    1983-08-01

    The results on 5 K irradiation available so far may be summarized as follows. (1) Increases of j/sub c/ following neutron irradiation occur only in conductors which are far from the optimal metallurgical treatments. (2) The changes of j/sub c/ following neutron irradiation and a thermal cycle to room temperature are small and in most cases comparable to the results obtained after 77 K irradiation. (3) The data available so far indicate that the degradation of j/sub c/ at 8 T is larger by about 5 to 10% than the corresponding changes at 5 T at a neutron fluence of 1.3 x 10 22 m -2 (E > 0.1 MeV). (4) The increase of Cu-resistivity is significant even after a thermal cycle to room temperature and requires design changes for a stable magnet operation

  9. Microhardness measurement in AISI 321 stainless steel with niobium additions before and after fast neutron irradiation

    International Nuclear Information System (INIS)

    Galli, V.L.; Lucki, G.

    1980-01-01

    Data about influence of neutron irradiation on the microhardness of stainless steel of type AISI 321 with 0.05 and 0.1wt.% Nb additions are presented. The microhardness measurements were made in the range of 300 to 650 0 C, before and after fast neutron irradiation with fluences about 10 17 n/cm 2 . Our results indicate that radiation damage peaks occur around 480 0 C for the stainless steel of type AISI 321 without Nb addition, around 500 0 C for the composition with 0.05 wt.% Nb addition and around 570 0 C for the composition with 0.1 wt.% Nb addition. Microhardness data are in agreement with those obtained by means of electrical resistivity measurements, performed at the same conditions. (Author) [pt

  10. 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.

  11. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Abou Taleb, W.M. [Alexandria Univ. (Egypt); Madi, N.K.; Kassem, M.E.; El-Khatib, A.M. [Alexandria Univ. (Egypt). Dept. of Physics

    1996-05-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10{sup 9} n/cm{sup 2}. The optical energy gap E{sub op} exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C{sub p} showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author).

  12. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    International Nuclear Information System (INIS)

    Abou Taleb, W.M.; Madi, N.K.; Kassem, M.E.; El-Khatib, A.M.

    1996-01-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10 9 n/cm 2 . The optical energy gap E op exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C p showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author)

  13. Point defects and magnetic properties of neutron irradiated MgO single crystal

    Directory of Open Access Journals (Sweden)

    Mengxiong Cao

    2017-05-01

    Full Text Available (100-oriented MgO single crystals were irradiated to introduce point defects with different neutron doses ranging from 1.0×1016 to 1.0×1020 cm-2. The point defect configurations were studied with X-ray diffuse scattering and UV-Vis absorption spectra. The isointensity profiles of X-ray diffuse scattering caused by the cubic and double-force point defects in MgO were theoretically calculated based on the Huang scattering theory. The magnetic properties at different temperature were measured with superconducting quantum interference device (SQUID. The reciprocal space mappings (RSMs of irradiated MgO revealed notable diffuse scattering. The UV-Vis spectra indicated the presence of O Frenkel defects in irradiated MgO. Neutron-irradiated MgO was diamagnetic at room temperature and became ferromagnetic at low temperature due to O Frenkel defects induced by neutron-irradiation.

  14. Irradiation response of rapidly solidified Path A type prime candidate alloys

    International Nuclear Information System (INIS)

    Imeson, E.; Tong, C.; Lee, M.; Vander Sande, J.B.; Harling, O.K.

    1981-01-01

    The objective of this study is to present a first assessment of the microstructural response to neutron irradiation shown by Path A alloys prepared by rapid solidification processing. To more fully demonstrate the potential of the method, alloys with increased titanium and carbon content have been used in addition to the Path A prime candidate alloy

  15. Preliminary microstructural characterization by transmission electron microscopy of 14 MeV neutron irradiated type 316 stainless steel

    International Nuclear Information System (INIS)

    Echer, C.J.

    1977-01-01

    Substantial changes in the mechanical properties of 316 stainless steel were observed after neutron irradiation (phi/sub t/ = 2.3 x 10 21 n/m 2 and E = 14 MeV) at 25 0 C. Comparison of microstructures of the unirradiated and neutron irradiated materials were evaluated using transmission electron microscopy. Evidence of small defect clusters in the irradiated material was found. These findings are consistent with other investigators also evaluating low dose irradiations

  16. A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

    A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

  17. Characteristics of neutron irradiation facility and dose estimation method for neutron capture therapy at Kyoto University research reactor institute

    International Nuclear Information System (INIS)

    Kobayashi, T.; Sakurai, Y.; Kanda, K.

    2001-01-01

    The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)

  18. Increase of the electrical resistance of thin aluminium film due to 14 MeV neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, S K; Kumar, U; Singh, S P; Bhattacharya, S; Nigam, A K [Banaras Hindu Univ. (India). Dept. of Physics

    1978-01-01

    The effect of 14 MeV neutron bombardment on the electrical resistance of 500 A thick vacuum-coated Al film is investigated. In the beginning, a slow, then sharp and finally again slow increase is observed in the electrical resistance of the film. Transmission electron micrographs of the film after the same dose of neutron irradiation show a large number of defects produced in the film due to neutron irradiation, which seems to be the cause of this increase.

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

    International Nuclear Information System (INIS)

    Dalton, J.H.

    1978-01-01

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

  20. Resistivity recovery of neutron-irradiated and cold-worked thorium

    International Nuclear Information System (INIS)

    Tang, J.T.

    1977-02-01

    Results from a study of the resistivity recovery of neutron-irradiated and cold-worked thorium on isochronal annealing, activation energies, and isothermal annealing and kinetics are discussed. The nature and extent of radiation effects on the resistivity of thorium at 80 0 K, interpretation of stage II recovery above 80 0 K, and activation energy and interpretation of stage III recovery are also discussed. There are 79 references

  1. Comparison of damage microstructures in neutron-irradiated vanadium and iron

    International Nuclear Information System (INIS)

    Horton, L.L.; Farrell, K.

    1983-01-01

    The cavity morphology and dislocation loop geometry in bcc vanadium are compared with the previously reported observations for neutron-irradiated iron. The specimens were vanadium (V) with 100 wppM of interstitial impurities and vanadium with boron carbide additions (V-B 4 C) which were irradiated to approx. 1 dpa in the same Oak Ridge Research Reactor capsules as the iron specimens

  2. Contribution to the study of recoil species produced by potassium ferrocyanide neutron irradiation

    International Nuclear Information System (INIS)

    Meriadec Vernier de Byans, B.

    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) [fr

  3. Anisotropy variation of crystallographic orientation in pyrocarbon coatings of fuel particles by annealing and neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Koizlik, K.

    1973-04-15

    This document is a translation of those parts of the German report Jul-868-RW concerned with changes in anisotropy as determined using an optical technique on pyrocarbon coatings on fuel particles resulting from annealing and neutron irradiations. Two lists of contents are included, one is for the present document and the other is the full contents of the original report and is included for the generl interest of users.

  4. Investigation of the combined effect of neutron irradiation and electron beam exposure on pure tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Van Renterghem, W., E-mail: wvrenter@sckcen.be; Uytdenhouwen, I., E-mail: iuytdenh@sckcen.be

    2016-08-15

    Pure tungsten samples were neutron irradiated in the BR2 reactor of SCK·CEN to fluences of 1.47 × 10{sup 20} n/cm{sup 2} and 4.74 × 10{sup 20} n/cm{sup 2} at 300 °C under Helium atmosphere and exposed to the electron beam of the Judith 1 installation The effect of these treatments on the defect structure was studied with transmission electron microscopy. In the irradiated samples the defect structure in the bulk is compared to the structure at the surface. The neutron irradiation created a large amount of a/2‹111› type dislocation loops forming dislocation rafts. The loop density increased from 8.5 × 10{sup 21}/m³ to 9 × 10{sup 22}/m³ with increasing dose, while the loop size decreased from 5.2 nm to 3.5 nm. The electron beam exposure induced significant annealing of the defects and almost all of the dislocation loops were removed. The number of line dislocations in that area increased as a result of the thermal stresses from the thermal shock. - Highlights: • Neutron irradiated and electron beam exposed tungsten samples were studied with transmission electron microscopy. • Neutron irradiation creates dislocation loops and rafts, while voids are created at higher irradiation dose. • No precipitates of transmutation products were found under these low dose irradiation conditions. • Electron beam exposure annihilates the dislocation loops and rafts.

  5. Effect of fast neutron irradiation on the thermal and electrical properties of acrylic resin

    International Nuclear Information System (INIS)

    Madi, N.K.; El-Khatib, A.M.; Kassem, M.E.; Ammar, E.A.

    1993-01-01

    Infrared technique was used to elucidate the effect of neutron irradiation on the structure of the polymethyl methacrylate (PMMA). It was found that PMMA resists large doses of irradiation at room temperature. The thermo-mechanical and electrical conductivity experiments have been applied to confirm the chemical results. The results show that the physical properties were slightly improved. The observations are correlated with the crystallinity produced by the accumulation of stable radiation defects. (author). 12 refs., 5 figs

  6. Spontaneous recombination volumes of Frenkel defects in neutron-irradiated non-fcc metals

    International Nuclear Information System (INIS)

    Nakagawa, M.; Mansel, W.; Boening, K.; Rosner, P.; Vogl, G.

    1979-01-01

    Production and production-rate curves for the non-fcc metals Fe, Mo, Ta, W, Zr, and Sn are obtained by electrical-resistivity measurements taken at 4.6 K during reactor neutron irradiations. The saturation concentration of Frenkel defects, c/sub s/, and the recombination volume v/sub o/ are evaluated. A parabolic relation between the spontaneous recombination volume v 0 and the compressibility kappa for a series of bcc metals is found

  7. Efigie: a computer program for calculating end-isotope accumulation by neutron irradiation and radioactive decay

    International Nuclear Information System (INIS)

    Ropero, M.

    1978-01-01

    Efigie is a program written in Fortran V which can calculate the concentration of radionuclides produced by neutron irradiation of a target made of either a single isotope or several isotopes. The program includes optimization criteria that can be applied when the goal is the production of a single nuclide. The effect of a cooling time before chemical processing of the target is also accounted for.(author) [es

  8. The Edinburgh experience of treating sarcomas of soft tissues and bone with neutron irradiation

    International Nuclear Information System (INIS)

    Duncan, W.; Arnott, S.J.; Jack, W.J.L.

    1986-01-01

    The experience of treating 30 patients with sarcomas of soft tissue and bone with d(15)+Be neutron irradiation is reported. The local control of measurable soft-tissue sarcomas was 38.5% (minimum follow-up 2 years), which is similar to that expected after photon therapy. The radiation morbidity was unacceptably high (50%). Bone tumours did not respond well; in only one out of nine was lasting local tumour control achieved. (author)

  9. A comparison of space closure rates between preactivated nickel-titanium and titanium-molybdenum alloy T-loops: a randomized controlled clinical trial.

    Science.gov (United States)

    Keng, Feng-Yi; Quick, Andrew N; Swain, Michael V; Herbison, Peter

    2012-02-01

    The purpose of this study was to conduct a prospective randomized controlled clinical trial to evaluate the rate of space closure and tooth angulation during maxillary canine retraction using preactivated T-loops made from titanium-molybdenum alloy (TMA) and nickel-titanium (NiTi). Twelve patients (six males and six females) aged between 13 and 20 years who had upper premolar extractions were included, and each acted as their own control, with a NiTi T-loop allocated to one quadrant and TMA to the other using a split mouth block randomization design. The loops were activated 3 mm at each visit to deliver a load of approximately 150 g to the upper canine teeth. Maxillary dental casts, taken at the first and each subsequent monthly visit, were used to evaluate changes in extraction space and canine angulation. All used T-loops were compared with unused loops in order to assess distortion. Mixed model statistical analysis was used to adjust for confounding variables. The mean rate of canine retraction using preactivated NiTi and TMA T-loops was 0.91 mm/month (±0.46) and 0.87 mm/month (±0.34), respectively. The canine tipping rates were 0.71 degrees/month (±2.34) for NiTi and 1.15 degrees/month (±2.86) for TMA. Both the rate of space closure and the tipping were not significantly different between the two wire types. The average percentage distortion of the TMA T-loop was 10 times greater than that of the NiTi loops when all other variables were matched. There was no difference in the rate of space closure or tooth angulation between preactivated TMA or NiTi T-loops when used to retract upper canines. The NiTi loops possessed a greater ability to retain and return to their original shapes following cyclical activation.

  10. Apoptosis of nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation

    International Nuclear Information System (INIS)

    Liang Ke; He Shaoqin; Feng Yan; Tang Jinhua; Feng Qinfu; Shen Yu; Yin Weibo; Xu Guozhen; Liu Xinfan; Wang Luhua; Gao Li

    1999-01-01

    Objective: To study the apoptotic response of the nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation. Methods: CNE-2 cells were cultured as usual. Using the techniques of DNA agarose gel electrophoresis and DNA special fluorescent staining, the status of apoptosis in CNE-2 cells after neutron irradiation was detected. Results: It was shown that the apoptosis can be induced in CNE-2 cell after neutron radiation. Six hrs, after different doses of neutron (0/0.667/1.333/2.000/2.667/3.333 Gy) and X-ray 0/2/4/6/8/10 Gy) irradiation the apoptotic rates were 2.4%, 6.3%, 7.1%, 9.5%, 13.5%, 14.6% and 2.4%, 3.8%, 5.7%, 7.8%, 10.4%, 11.7%, respectively; at 48 hrs they were 18.3%, 21.5%, 22.8%, 29.3%, 34.2% and 13.7%, 17.6%, 21.3%, 25.6%, 28.9%, respectively. At 10 hrs after neutron irradiation the DNA ladder of apoptosis could be detected between 0.667-3.333 Gy doses in CNE-2 cells by DNA agarose gel electrophoresis. Conclusion: Neutron radiation can induce apoptosis in tumor cells. Compared with the X-ray, neutron induces apoptosis in larger extent than X-ray in the same condition; meanwhile, apoptosis after irradiation is dose and time dependent

  11. Determination of tritium generation and release parameters at lithium CPS under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ponkratov, Yuriy, E-mail: ponkratov@nnc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Baklanov, Viktor; Skakov, Mazhyn; Kulsartov, Timur; Tazhibayeva, Irina; Gordienko, Yuriy; Zaurbekova, Zhanna; Tulubayev, Yevgeniy [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Chikhray, Yevgeniy [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan); Lyublinski, Igor [JSC “Star”, Moscow (Russian Federation); NRNU “MEPhI”, Moscow (Russian Federation); Vertkov, Alexey [JSC “Star”, Moscow (Russian Federation)

    2016-11-01

    Highlights: • The main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor is described in paper. • In the experiments a very small tritium release was fixed likely due to its high solubility in liquid lithium. • If the lithium CPS will be used as a plasma facing material in temperature range up to 773 K under neutron irradiation only helium will release from lithium CPS into a vacuum chamber. - Abstract: This paper describes the main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor. The experiments were carried out using the method of mass-spectrometric registration of released gases and using a specially constructed ampoule device. Irradiation was carried out at different reactor thermal powers (1, 2 and 6 MW) and sample temperatures from 473 to 773 K. In the experiments a very small tritium release was detected likely due to its high solubility in liquid lithium. It can be caused by formation of lithium tritide during tritium diffusion to the lithium surface.

  12. The effect of low-dose neutron irradiation on extracellular matrix

    International Nuclear Information System (INIS)

    Chen Tiehe; Lu Yongjie; Chai Mingsheng; Peng Wulin; Yang Yifang; Pan Yan; Chen Jinguo

    2003-01-01

    Projective: To study the effect of neutron irradiation on extracellular matrix. Methods: 120 male wistar rats were divided into four groups at random, and then exposed to neutron of 252 Cf-source at the doses of 0, 0.29, 0.62 and 1.20 Gy, respectively. After the exposure of 3 days, 1 month and 2 months, the rats were sacrificed and lung tissue specimens stored at -30 degree C. Hyaluronan, laminin, type III procollagen and type IV collagen in the lung tissue were detected by the method of radioimmunoassay. Results: The differences of the levels of hyaluronan in lung tissue among the groups were unsignificant. The levels of laminin in 0.29, 0.62 and 1.20 Gy groups after the 3-day exposure were remarkably different to those of the control group, and unable to recover completely even 2 months after the exposure. The levels of type IV collagen in higher three irradiated groups were all higher, but not significantly. The levels of type III procollagen in the early stage after exposure were higher, and later they lowered. Conclusion: The levels of some components of extracellular matrix in the lung tissue of rat can be changed by low-dose of neutron irradiation, but their variational modes and degrees depend on the dose of neutron irradiation and the length of period after exposure

  13. Large lattice relaxation deep levels in neutron-irradiated GaN

    International Nuclear Information System (INIS)

    Li, S.; Zhang, J.D.; Beling, C.D.; Wang, K.; Wang, R.X.; Gong, M.; Sarkar, C.K.

    2005-01-01

    Deep level transient spectroscopy (DLTS) and deep level optical spectroscopy (DLOS) measurements have been carried out in neutron-irradiated n-type hydride-vapor-phase-epitaxy-grown GaN. A defect center characterized by a DLTS line, labeled as N1, is observed at E C -E T =0.17 eV. Another line, labeled as N2, at E C -E T =0.23 eV, seems to be induced at the same rate as N1 under irradiation and may be identified with E1. Other defects native to wurtzite GaN such as the C and E2 lines appear to enhance under neutron irradiation. The DLOS results show that the defects N1 and N2 have large Frank-Condon shifts of 0.64 and 0.67 eV, respectively, and hence large lattice relaxations. The as-grown and neutron-irradiated samples all exhibit the persistent photoconductivity effect commonly seen in GaN that may be attributed to DX centers. The concentration of the DX centers increases significantly with neutron dosage and is helpful in sustaining sample conductivity at low temperatures, thus making possible DLTS measurements on N1 an N2 in the radiation-induced deep-donor defect compensated material which otherwise are prevented by carrier freeze-out

  14. Installation and measurement capacity of 3 x 592 GBq 241Am-Be neutron irradiation cell

    International Nuclear Information System (INIS)

    Bulut, Serdar; Celenk, I.

    2013-01-01

    In this study, the installation and measurement capacity of the neutron irradiation system are investigated. First of all an irradiation geometry enabling optimum irradiation was designed for three 241 Am-Be sources each of it having 592 GBq activity. Neutron irradiation system was installed after design and optimization of the system including the design of appropriate moderator and shielding were completed. Radiation safety standards of the Neutron Research Laboratory fulfilling the requirements of national regulation were achieved with unique configuration of the shielding materials. In this study the results of qualitative and quantitative detection limits obtained for Na, Al, Cl, K, Ti, V, Mn, Fe, Co, Cu, Zn, Ga, Ge, As, Se, Br, Rb, Sr, Y, Mo, Ru, Ag, Cd, In, Sb, I, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu, Gd, Dy, Ho, Yb, Lu, Hf, W, Pt, Au, Th and U elements by using the neutron irradiation cell comprising 3 x 592 GBq 241 Am-Be isotopic neutron source are presented and discussed. (orig.)

  15. 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

  16. Impact of helium implantation and ion-induced damage on reflectivity of molybdenum mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Carrasco, A., E-mail: alvarogc@kth.se [Department of Fusion Plasma Physics, Royal Institute of Technology (KTH), Teknikringen 31, 100 44 Stockholm (Sweden); Petersson, P.; Hallén, A. [Department of Fusion Plasma Physics, Royal Institute of Technology (KTH), Teknikringen 31, 100 44 Stockholm (Sweden); Grzonka, J. [Faculty of Materials Science and Engineering, Warsaw University of Technology, 02-507 Warsaw (Poland); Institute of Electronic Materials Technology, 133 Wolczynska Str., 01-919 Warsaw (Poland); Gilbert, M.R. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Fortuna-Zalesna, E. [Faculty of Materials Science and Engineering, Warsaw University of Technology, 02-507 Warsaw (Poland); Rubel, M. [Department of Fusion Plasma Physics, Royal Institute of Technology (KTH), Teknikringen 31, 100 44 Stockholm (Sweden)

    2016-09-01

    Molybdenum mirrors were irradiated with Mo and He ions to simulate the effect of neutron irradiation on diagnostic first mirrors in next-generation fusion devices. Up to 30 dpa were produced under molybdenum irradiation leading to a slight decrease of reflectivity in the near infrared range. After 3 × 10{sup 17} cm{sup −2} of helium irradiation, reflectivity decreased by up to 20%. Combined irradiation by helium and molybdenum led to similar effects on reflectivity as irradiation with helium alone. Ion beam analysis showed that only 7% of the implanted helium was retained in the first 40 nm layer of the mirror. The structure of the near-surface layer after irradiation was studied with scanning transmission electron microscopy and the extent and size distribution of helium bubbles was documented. The consequences of ion-induced damage on the performance of diagnostic components are discussed.

  17. 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...... 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...... 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...

  18. Neutron resistant irradiation alloy and usage thereof

    International Nuclear Information System (INIS)

    Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

    1997-01-01

    A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

  19. Method to obtain molybdenum 99 from a matrix containing neutron-irradiated, fissionable materials and fission products

    International Nuclear Information System (INIS)

    Ali, S.A.H.; Knapp, J.; Reinhardt, J.

    1977-01-01

    A method to obtain a highly pure Mo 99 compound is suggested which is easy to perform even with remote handling. The matrix irradiated by neutrons is rendered soluble with aqueous alkali hydroxide solution, the Mo 99 being dissolved. The Mo rhodanide complex is collected on an organic ion exchanger. The individual process steps are described. A suitable iodine reducing agent (e.g. Na or Ka sulphite, hydroxyl ammonium sulphate) is already added to the first step and to the washing of the ion exchanger. Where necessary Tc 99 is eluated from the fixed Mo 99 mother nuclide for use in nuclear medicine. (UWI) [de

  20. Molybdenum solubility in aluminium nitrate solutions

    Energy Technology Data Exchange (ETDEWEB)

    Heres, X.; Sans, D.; Bertrand, M.; Eysseric, C. [CEA, Centre de Marcoule, Nuclear Energy Division, DRCP, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Brackx, E.; Domenger, R.; Excoffier, E. [CEA, Centre de Marcoule, Nuclear Energy Division, DTEC, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France); Valery, J.F. [AREVA-NC, DOR/RDP, Paris - La Defense (France)

    2016-07-01

    For over 60 years, research reactors (RR or RTR for research testing reactors) have been used as neutron sources for research, radioisotope production ({sup 99}Mo/{sup 99m}Tc), nuclear medicine, materials characterization, etc... Currently, over 240 of these reactors are in operation in 56 countries. They are simpler than power reactors and operate at lower temperature (cooled to below 100 C. degrees). The fuel assemblies are typically plates or cylinders of uranium alloy and aluminium (U-Al) coated with pure aluminium. These fuels can be processed in AREVA La Hague plant after batch dissolution in concentrated nitric acid and mixing with UOX fuel streams. The aim of this study is to accurately measure the solubility of molybdenum in nitric acid solution containing high concentrations of aluminium. The higher the molybdenum solubility is, the more flexible reprocessing operations are, especially when the spent fuels contain high amounts of molybdenum. To be most representative of the dissolution process, uranium-molybdenum alloy and molybdenum metal powder were dissolved in solutions of aluminium nitrate at the nominal dissolution temperature. The experiments showed complete dissolution of metallic elements after 30 minutes long stirring, even if molybdenum metal was added in excess. After an induction period, a slow precipitation of molybdic acid occurs for about 15 hours. The data obtained show the molybdenum solubility decreases with increasing aluminium concentration. The solubility law follows an exponential relation around 40 g/L of aluminium with a high determination coefficient. Molybdenum solubility is not impacted by the presence of gadolinium, or by an increasing concentration of uranium. (authors)