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During the ITER EDA (Engineering Design Activity), the US Home Team developed improved methods for fabricating tungsten armored plasma facing components. Thermo-mechanical modeling indicated the de...

Several types of porous media heat exchangers are being evaluated for use infusion applications. Broadly, these devices can be classified as capillary-pumped (heat pipes) or mechanically-pumped hea...

Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m{sup 2} and surface temperatures near 300{degrees}C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15{degrees}C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m{sup 2} and surface temperatures up to 690{degrees}C before debonding at 10 MW/m{sup 2}. A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m{sup 2}, while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue.

(1996). Second International Energy Agency Workshop on Beryllium Technology for Fusion, Jackson Lake Lodge, Wyoming, September 6–8, 1995. Fusion Technology: Vol. 29, No. 3, pp. 409-412.

Tokamaks currently in operation deposit on the order of 1-30 MW/m2 onto plasma facing surfaces during normal operation and hundreds of MW/m2 for shorter periods of time (0.1–3 ms) during disruptions. Disruption deposited energies on future high-power tokamaks may be well in excess of 20 MJ/m2 Design of plasma facing components (PFCs) for such severe environments requires considerable advancements in materials development, armor tile bonding to actively cooled substrates, heat transfer, and many other areas of engineering concern. Considerable improvements in PFC performance, reliability and lifetime can also be accomplished through improved understanding and control of the edge plasma boundary layer. This paper covers both engineering and edge plasma physics issues that must be addressed in the development of reliable PFCs for ITER. Several specific examples are addressed since a complete treatment of all critical development issues would be lengthy. Topics covered include impurity generation and ...

The divertor of a near-term fusion device has to withstand high heat fluxes, heat shocks, and erosion caused by the plasma. Furthermore, it has to be maintainable through remote techniques. Above a...

The monoblock geometry is proposed for the ITER Physics Phase divertor for brazing of carbon armor tiles to copper or molybdenum cooling tubes. Elastic/plastic finite element analyses predicted high residual stresses except with OFHC copper. Samples of pyrolytic graphite tiles brazed to OFHC copper, Glidcop™ Al-15 copper alloy, and molybdenum tubing show cracking in all of the samples, except with the OFHC copper. A 3-tile divertor target consisting of 12 mm thick pyrolytic graphite brazed with a copper-silver alloy to a 12 mm diameter OFHC copper tubing was tested at 15 MW/m2 with a rastered 30 keV electron beam for 1000 thermal cycles. A gradual rise in surface temperature from 1000 C to 1200 C over the 1000 cycles was observed, along with hot stripes (1500 C) at the tile edges. However, no delamination cracks could be detected.

Critical Heat Flux (CHF), also called burnout, is one of the major design limits for water-cooled divertors in tokamaks. Another important design issue is the correct thermal modeling of the divert...

A collaborative program is underway to field a comprehensive pump limiter experimental program on TEXTOR: Advanced Limiter Test-I (ALT-I). Either of two interchangeable limiter modules may be attached to an insertion/rotation mechanism to direct particles to the about1 m/sup 3/ pumping chamber. Pumping is provided primarily by a solid getter assembly at about20,000 1/sec. Variation of geometric dimensions and gas puffing in the modules will allow the study of plasma and neutral interaction in differing recycle regimes. Multiple diagnostic systems are incorporated into the pump limiter design for use in conjunction with TEXTOR plasma diagnostics. Initial experiments are scheduled for December 1983.

Advanced Limiter Test II (ALT-II) is an axisymmetric toroidal belt pump limiter scheduled for installation on the TEXTOR tokamak in 1986. ALT-II consists of eight blade segments mounted on adjustab...

Sandia National Laboratories, Division of Fusion Technology, has performed a series of tests in support of the Compact Ignition Tokamak first-wall tile design. A screening study was done to rank the thermal shock resistance of 25 candidate graphite materials. Standardized thermal shock samples were subjected to identical intense heat pulses generated by the electron beam test system. Most of the fine-grained graphites cracked, but none of the carbon-carbon (C-C) composite samples fractured. The best performing fine-grained graphites were, in order, Union Carbide's TS-1909, TS-1792, ATJ-S, and CGW, as well as Fiber Material Inc.'s (FMI's) high-density graphite (HDG).