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Sample records for alloy-tzm

  1. Pressure bonding molybdenum alloy (TZM) to reaction-bonded silicon nitride

    International Nuclear Information System (INIS)

    Huffsmith, S.A.; Landingham, R.L.

    1978-01-01

    Topping cycles could boost the energy efficiencies of a variety of systems by using what is now waste heat. One such topping cycle uses a ceramic helical expander and would require that a reaction-bonded silicon nitride (RBSN) rotor be bonded to a shaft of TZM (Mo-0.5 wt % Ti-0.08 wt % Zr). Coupon studies show that TZM can be bonded to RBSN at 1300 0 C and 69 MPa if there is an interlayer of MoSi 2 . A layer of finely ground (10 μm) MoSi 2 facilitates bond formation and provides a thicker bond interface. The hardness and grain structure of the TZM and RBSN were not affected by the temperature and pressure required to bond the coupons

  2. 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

  3. 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.

  4. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. General synthesis

    International Nuclear Information System (INIS)

    Hery, M.; Lecocq, A.

    1983-03-01

    After a brief recall of the MSBR project, French studies on molten salt reactors are summed up. Theoretical and experimental studies for a graphite moderated 1000 MWe reactor using molten Li, Be, Th and U fluorides cooled by salt-lead direct contact are given. These studies concern the core, molten salt chemistry, graphite, metals (molybdenum, alloy TZM), corrosion, reactor components [fr

  5. 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)

  6. Plasma facing surface composition during NSTX Li experiments

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H., E-mail: cskinner@pppl.gov [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States); Sullenberger, R. [Department of Mechanical and Aerospace Engineering, Princeton University, NJ 08540 (United States); Koel, B.E. [Department of Chemical and Biological Engineering, Princeton University, NJ 08540 (United States); Jaworski, M.A.; Kugel, H.W. [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States)

    2013-07-15

    Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma–lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium have been examined using X-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1–2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.

  7. Effect of low fatigue on the ductile-brittle transition of molybdenum

    International Nuclear Information System (INIS)

    Furuya, K.; Nagata, N.; Watanabe, R.; Yoshida, H.

    1982-01-01

    An explicit ductile-brittle transition of molybdenum occurring in both tensile and low cycle fatigue tests was investigated. Tests were performed on several sorts of molybdenum and its alloy TZM, and effects of heat treatment, fabrication method and alloying on the transition behavior and fracture mode are described in detail. All the materials exhibited a brittle failure with degraded fatigue behavior at room temperature, while they became ductile as temperature increased up to 573 K. The tendency of fatigue results was qualitatively in accordance with that of reduction of area in tensile tests. Differences among the materials were minor on the ductile-brittle transition temperature (DBTT), but major on the fatigue life for the embrittled materials. (orig.)

  8. Activation of TZM and stainless steel divertor materials in the NET fusion machine

    International Nuclear Information System (INIS)

    Cepraga, D.G.; Menapace, E.; Cambi, G.; Ciattaglia, S.; Petrizzi, L.; Cavallone, G.; Costa, M.; Broccoli, U.

    1994-01-01

    This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code. ((orig.))

  9. Activation of TZM and stainless steel divertor materials in the NET fusion machine

    Energy Technology Data Exchange (ETDEWEB)

    Cepraga, D G [ENEA, INN-FIS, 8 Viale Ercolani, 40138, Bologna (Italy); Menapace, E [ENEA, INN-FIS, 8 Viale Ercolani, 40138, Bologna (Italy); Cambi, G [Bologna University, Physics Department, 33 Via Irnerio, 40126, Bologna (Italy); Ciattaglia, S [ENEA, NUC-FUS, 27 Via E. Fermi, 00044, Frascati (Italy); Petrizzi, L [ENEA, NUC-FUS, 27 Via E. Fermi, 00044, Frascati (Italy); Cavallone, G [NIER S.r.l., 16 Via S. Stefano, 40125, Bologna (Italy); Costa, M [NIER S.r.l., 16 Via S. Stefano, 40125, Bologna (Italy); Broccoli, U [ENEA, NUC-RIN, 4 Via Martiri del Sole, 40100, Bologna (Italy)

    1994-09-01

    This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code. ((orig.))

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-27

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

  11. Fabrication of thin-wall, freestanding inertial confinement fusion targets by chemical vapor deposition

    International Nuclear Information System (INIS)

    Carroll, D.W.; McCreary, W.J.

    1982-01-01

    To meet the requirements for plasma physics experiments in the inertial confinement fusion (ICF) program, chemical vapor deposition (CVD) in fluid beds was used to fabricate freestanding tungsten spheres and cylinders with wall thicknesses less than 5.0 μm. Molybdenum and molybdenum alloy (TZM) mandrels of the desired geometry were suspended in a carrier bed of dense microspheres contained in an induction-heated fluid-bed reactor. The mandrels were free to float randomly through the bed, and using the reaction WF 6 +3H 2 →/sub /KW +6HF, very fine-grained tungsten was deposited onto the surface at a rate and in a grain size determined by temperature, gas flow rate, system pressure, and duration of the reaction. After coating, a portion of each mandrel was exposed by hole drilling or grinding. The mandrel was then removed by acid leaching, leaving a freestanding tungsten shape. Experimental procedures, mandrel preparation, and results obtained are discussed

  12. Developments toward the use of tungsten as armour material in plasma facing components promoted by Euratom-CEA Association

    International Nuclear Information System (INIS)

    Mitteau, R.; Missiaen, J.M.; Brustolin, P.

    2006-01-01

    Tungsten is increasingly considered as a prime candidate armour material facing the plasma in fusion experiments (ASDEX, JET, ITER). This material is, however, a challenge for the engineers due to its brittleness at room temperature. Its bonding to structural or cooled substrates is a critical issue. The Euratom-CEA Association promotes the development of evolutionary techniques aiming to produce high performance assemblies between tungsten and various substrates. These are 1) functionally graded tungsten to copper, 2) direct electron beam welding of tungsten to Mo-alloy TZM and 3) the characterisation of tungsten coatings deposited on carbon fibre composite by high energy deposition processes. 1) A functionally graded material eliminates the singular point which weakens the heterogeneous assembly, reducing the stresses and allowing a better behaviour. The sintering of submicronic W-Cu powders is investigated. The green shape is processed from W-CuO powder, which is reduced by a hydrogen flow. The compaction and sintering of layers of various compositions (10 to 30 % Cu) produces an assembly (density of ∼ 94%) with a good cohesion. However, the gradient is not effectively controlled, because of the migration of melt copper during the sintering. Future work aims to improve the process by using spark or microwave assisted sintering. 2) Electron beam welding of Mo-alloy TZM is investigated, to produce high temperature components required by radiation cooled PFCs. They require only mechanical properties and no vacuum sealing. The driving line is to use simple tungsten shapes to reduce the milling cost. In spite of low weldable properties of the refractory alloys, a good bonding up to a depth of 5 mm is obtained. Hardness measurements show that the melt area and the heat affected zone are harder than TZM, the weakest materials at 230 Hv. Quench tests in water from up to 2000 o C are done without apparent crack formation. 3) Finally, characterisation techniques are

  13. Thermal and mechanical analysis of the Faraday shield for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Vesey, R.A.

    1988-02-01

    The antenna for the ion cyclotron resonance heating (ICRH) system of the Compact Ignition Tokamak (CIT) is protected from the plasma environment by a Faraday shield, an array of gas-cooled metallic tubes. The plasma side of the tubes is armored with graphite tiles, which can be either brazed or mechanically attached to the tube. The Faraday shield has been analyzed using finite element codes to model thermal and mechanical responses to typical CIT heating and disruption loads. Four representative materials (Inconel 718, tantalum-10 tungsten, copper alloy C17510, and molybdenum alloy TZM) and several combinations of tube and armor thicknesses were used in the thermal analysis, which revealed that maximum allowable temperatures were not exceeded for any of the four materials considered. The two-dimensional thermal stress analysis indicated Von Mises stresses greater than twice the yield stress for a tube constructed of Inconel 718 (the original design material) for the brazed-graphite design. Analysis of stresses caused by plasma disruption (/rvec J/ /times/ /rvec B/) loads eliminated the copper and molybdenum alloys as candidate tube materials. Of the four materials considered, tantalum-10 tungsten performed the best for a brazed graphite design, showing acceptable thermal stresses (69% of yield) and disruption stresses (42% of yield). A preliminary thermal analysis of the mechanically attached graphite scheme predicts minimal thermal stresses in the tube. The survivability of the graphite tubes in this scheme is yet to be analyzed. 8 refs., 19 figs., 2 tabs

  14. Materials for advanced high temperature reactors

    International Nuclear Information System (INIS)

    Graham, L.W.

    1977-01-01

    Materials are studied in advanced applications of high temperature reactors: helium gas turbine and process heat. Long term creep behavior and corrosion tests are conducted in simulated HTR helium up to 1000 deg C with impurities additions in the furnace atmosphere. Corrosion studies on AISI 321 steels at 800-1000 deg C have shown that the O 2 partial pressure is as low as 10 -24+-3 atm, Ni and Fe cannot be oxidised above about 500 and 600 deg C, Cr cease to oxidise at 800 to 900 deg C and Ti at 900 to 1000 deg C depending on alloy composition γ' strengthened superalloys must depend on a protective corrosion mechanism assisted by the presence of Ti and possibly Cr. Carburisation has been identified metallographically in several high temperature materials: Hastelloy X and M21Z. Alloy TZM appears to be inert in HTR Helium at 900 and 1000 deg C. In alloy 800 and Inconel 625 surface cracks initiation is suppressed but crack propagation is accelerated but this was not apparent in AISI steels, Hastelloy X or fine grain Inconel at 750 deg C

  15. 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

  16. 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

  17. Modular He-cooled divertor for power plant application

    International Nuclear Information System (INIS)

    Diegele, Eberhard; Kruessmann, R.; Malang, S.; Norajitra, P.; Rizzi, G.

    2003-01-01

    Gas cooled divertor concepts are regarded as a suitable option for fusion power plants because of an increased thermal efficiency for power conversion systems and the use of a coolant compatible with all blanket systems. A modular helium cooled divertor concept is proposed with an improved heat transfer. The concept employs small tiles made of tungsten and brazed to a finger-like structure made of Mo-alloy (TZM). Design goal was a heat flux of at least 15 MW/m 2 and a minimum temperature of the structure of 600 deg.C. The divertor has to survive a number of cycles (100-1000) between operating temperature and room temperature even for the steady state operation assumed. Thermo-hydraulic design requirements for the concepts include to keep the pumping power below 10% of the thermal power to the divertor plates, and simultaneously achieving a heat transfer coefficient in excess of 60 kW/m 2 K. Inelastic stress analysis indicates that design allowable stress limits on primary and secondary (thermal) stresses as required by the ITER structural design criteria are met even under conservative assumptions. Finally, critical issues for future development are addressed

  18. 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

  19. Performance demonstration of a high-power space-reactor heat-pipe design

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Martinez, E.H.; Keddy, E.S.; Runyan, J.; Kemme, J.E.

    1983-01-01

    Performance of a 15.9-mm diam, 2-m long, artery heat pipe has been demonstrated at power levels to 22.6 kW and temperatures to 1500 0 K. The heat pipe employed lithium as a working fluid with distribution wicks and arteries fabricated from 400 mesh Mo-41 wt % Re screen. Molybdenum alloy (TZM) was used for the container. Peak axial power density attained in the testing was 19 kW/cm 2 at 1465 0 K. The corresponding radial flux density in the evaporator region of the heat pipe was 150 W/cm 2 . The extrapolated limit for the heat pipe at its 1500 0 K design point is 30 kW, corresponding to an axial flux density of 25 kW/cm 2 . Sonic and capillary limits for the design were investigated in the 1100 to 1500 0 K temperature range. Excellent agreement of measured and predicted temperature and power levels was observed

  20. The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

    2017-08-01

    Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

  1. High-energy high-rate pulsed-power processing of materials by powder consolidation and by railgun deposition. Technical report (Final), 10 April 1985-10 February 1987

    Energy Technology Data Exchange (ETDEWEB)

    Persad, C.; Marcus, H.L.; Weldon, W.F.

    1987-03-31

    This exploratory research program was initiated to investigate the potential of using pulse power sources for powder consolidation, deposition and other high-energy high-rate processing. The characteristics of the high-energy-high-rate (1MJ/s) powder consolidation using megampere current pulses from a homopolar generator, were defined. Molybdenum Alloy TZM, a nickel-based metallic glass, copper/graphite composites, and P/M aluminum alloy X7091 were investigated. The powder-consolidation process produced high densification rates. Density values of 80% to 99% could be obtained with subsecond high-temperature exposure. Specific energy input and applied pressure were controlling process parameters. Time temperature transformation (TTT) concepts underpin a fundamental understanding of pulsed power processing. Inherent control of energy input, and time-to-peak processing temperature developed to be held to short times. Deposition experiments were conducted using an exploding-foil device (EFD) providing an armature feed to railgun mounted in a vacuum chamber. The material to be deposited - in plasma, gas, liquid, or solid state - was accelerated electromagnetically in the railgun and deposited on a substrate. Deposits of a wide variety of single- and multi-specie materials were produced on several types of substrates. In a series of ancillary experiments, pulsed-skin-effect heating and self quenching of metallic conductors was discovered to be a new means of surface modification by high-energy high-rate-processing.

  2. A new system for sodium flux growth of bulk GaN. Part I: System development

    Science.gov (United States)

    Von Dollen, Paul; Pimputkar, Siddha; Alreesh, Mohammed Abo; Albrithen, Hamad; Suihkonen, Sami; Nakamura, Shuji; Speck, James S.

    2016-12-01

    Though several methods exist to produce bulk crystals of gallium nitride (GaN), none have been commercialized on a large scale. The sodium flux method, which involves precipitation of GaN from a sodium-gallium melt supersaturated with nitrogen, offers potentially lower cost production due to relatively mild process conditions while maintaining high crystal quality. We successfully developed a novel apparatus for conducting crystal growth of bulk GaN using the sodium flux method which has advantages with respect to prior reports. A key task was to prevent sodium loss or migration from the growth environment while permitting N2 to access the growing crystal. We accomplished this by implementing a reflux condensing stem along with a reusable capsule containing a hermetic seal. The reflux condensing stem also enabled direct monitoring of the melt temperature, which has not been previously reported for the sodium flux method. Furthermore, we identified and utilized molybdenum and the molybdenum alloy TZM as a material capable of directly containing the corrosive sodium-gallium melt. This allowed implementation of a crucible-free system, which may improve process control and potentially lower crystal impurity levels. Nucleation and growth of parasitic GaN ("PolyGaN") on non-seed surfaces occurred in early designs. However, the addition of carbon in later designs suppressed PolyGaN formation and allowed growth of single crystal GaN. Growth rates for the (0001) Ga face (+c-plane) were up to 14 μm/h while X-ray omega rocking (ω-XRC) curve full width half-max values were 731″ for crystals grown using a later system design. Oxygen levels were high, >1019 atoms/cm3, possibly due to reactor cleaning and handling procedures.

  3. Interactions of reactor helium and simulating gas mixtures with high-temperature metals with particular regard to simultaneous deformation

    International Nuclear Information System (INIS)

    Berchtold, L.

    1983-01-01

    For the observation of multicomponent alloys (Inconel 617 and 713LC, chroman (Ni20Cr), vacromium (Ni20Cr+Si), TZM) in multicomponent HTR atmospheres (HHT search gas), interaction between gases and metals was studied, both in theoretical descriptions and experimentally. From the experimental viewpoint, gradual simplification employs, on the one hand, tests effected in undiluted atmospheres with exclusively oxidizing or carburizing properties; on the other hand, more simple alloys and pure metals are applied specifically in the helium atmosphere. For an evaluation of the materials, it is maintained that in a strongly oxidizing (H 2 O-rich) atmosphere, e.g. in HHT search gas, materials with sufficient chrome content (e.g. 20% Cr in Ni alloys such as IN 617) offer favourable conditions for an almost complete interruption of carburizing reactions. In that case, the maintenance of the shielding effect of coating during rapid deformation and a tendency to planar delamination during deformation, which becomes stronger as the layer thickness increases, appear to be critical. Concentrations of oxide-forming agents stronger than chromium offer disadvantages rather than advantages. Owing to its tendency to flake off as the covering oxide SiO 2 or as part of a cover layer, silicon may more than destroy the light advantage of a slowed down process of carbon diffusion. The cast alloy IN 713LC shows a deep-reaching carburation in HHT search gas, both with and without deformation. No deep-reaching corrosive damage is noticeable on the molybdenum alloy TZM. (orig./MM) [de