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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. 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)

  5. 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

  6. High energy high rate pulsed power processing of materials by powder consolidation and by railgun deposition

    Science.gov (United States)

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

    1987-03-01

    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, have been defined. Molybdenum Alloy TZM, A Nickel based metallic glass, Copper graphite composites, and P/M Aluminum Alloy X7091 have been investigated. The powder consolidation process produced high densification rates. Density values of 80% to 99% could be obtained with sub second 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. 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.

  7. 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

  8. 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

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

    Science.gov (United States)

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

    2000-02-01

    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.

  10. 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

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. Parametric study of FER first wall and divertor plate performance

    International Nuclear Information System (INIS)

    Haines, J.R.; Kitamura, Kazunori; Kobayashi, Takeshi; Iida, Hiromasa

    1986-07-01

    Thermal, mechanical, and lifetime performance of various first wall and divertor plate materials were examined over a broad range of conditions, representative of those considered for next-generation tokamaks such as FER. Candidate plasma side materials include beryllium, graphite, silicon carbide, molybdenum, tantalum, and tungsten. Copper, copper alloy C17510, austenitic stainless steel (316SS), ferritic stainless steel (HT-9), vanadium alloy V-15Cr-5Ti, and molybdenum alloy TZM were considered as candidate heat sink/structural materials. Performance was examined at heat fluxes ranging from 0.05 MW/m 2 for the first wall up to 5.0 MW/m 2 for the divertor plate. Ion flux, plasma edge temperature, burn time per pulse, and number of operating cycles were the other major parameters varied in this study. The analysis model used for these studies includes: (1) a thermal model; (2) a thermal stress model; (3) a disruption erosion model; (4) a sputtering erosion model; and (5) a fatique lifetime model. Results show that recommended first wall and divertor plate designs perform adequately over most of the range of conditions considered for FER design options. Thermal shock of the plasma facing material during intense disruption heating and radiation damage and temperature limitations for graphite are identified as major concerns reguiring experimental investigation. (author)