Publications

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Utramet, Inc. fabricated one-piece heat exchanger tubes of chemical vapor deposited (CVD) tungsten (W), each with an internal porous mesh fused along either 51 or 38 mm of the axial length of a tube 15 mm in outer diameter. The open porous mesh has a structure of joined ligaments that combines relatively low resistance to flow and a large area for heat transfer. In tests at the Electron Beam Test Stand (EBTS) at Sandia National Laboratories, the maximum absorbed heat load was 22.4 MW/m2 with helium at 4 MPa, flowing at 27 g/s and with inlet and outlet temperatures of 40 and 91 °C and a pressure drop of ∼0.07 MPa. The preparation and testing of the samples was funded through a Phase I grant by the US Department of Energy's Small Business Innovation Research Program. The paper reports the surface temperature distribution indicated by an infrared camera, test conditions, a post-test examination in a scanning electron microscope and other details. © 2007 Elsevier B.V. All rights reserved.

Two water-cooled mockups with CuCrZr heat sinks and plasma sprayed beryllium (PS Be) armor, 5 and 10 mm thick respectively, were fabricated at Los Alamos National Laboratory and thermally cycled at Sandia at 1 and 2 MW/m2. The castellated surface of the CuCrZr mechanically locked the armor. The resulting PS Be morphology controlled cracking during thermal cycling. Post test examinations showed transverse cracks perpendicular to the surface of the armor that would relieve thermal stresses but not degrade heat transfer. The mockups and two others previously produced for the European Fusion Development Agreement had somewhat porous armor, with a thermal conductivity estimated to be about 1/4 that of fully dense beryllium, due to the low (600-650 °C) substrate temperature during deposition specifically requested by EFDA to avoid subsequent heat treating of CuCrZr. Some melting of the armor was expected and observed in the tests. © 2007 Elsevier B.V. All rights reserved.

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The US will supply outboard Module 18 for the International Thermonuclear Experimental Reactor. This module, radially thinner than other modules with a "nose" that curves radially outward to mate with the divertor, has the potential for high electromagnetic (EM) loads from vertical displacement events and high heat loads. The 316LN-IG shield block and first wall (FW) panels must be slotted to mitigate the EM loads and progress in developing the design is summarized. The FW has beryllium (Be) armor joined to a water-cooled CuCrZr heat sink with embedded 316LN-IG cooling channels. The US Team is considering possible fabrication methods as the design develops. Brief results of high heat flux experiments at Sandia on mockups with plasma-sprayed Be armor prepared at Los Alamos National Laboratory are noted. © 2005 Elsevier B.V. All rights reserved.

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An experiment at Sandia National Laboratories confirmed that a ternary salt (Flinabe, a ternary mixture of LiF, BeF{sub 2} and NaF) had a sufficiently low melting temperature ({approx}305 C) to be useful for first wall and blanket applications using flowing molten salts that were investigated in the Advanced Power Extraction (APEX) Program.[1] In the experiment, the salt pool was contained in a stainless steel crucible under vacuum. One thermocouple was placed in the salt and two others were embedded in the crucible. The results and observations from the experiment are reported in the companion paper.[2] The paper presented here will cover a 3-D finite element thermal analysis of the salt pool and crucible. The analysis was done to evaluate the thermal gradients in the salt pool and crucible and to compare the temperatures of the three thermocouples. One salt mixture appeared to melt and to solidify as a eutectic with a visible plateau in the cooling curve (i. e, time versus temperature for the thermocouple in the salt pool). This behavior was reproduced with the thermal model. Cases were run with several values of the thermal conductivity and latent heat of fusion to see the parametric effects of these changes on the respective cooling curves. The crucible was heated by an electrical heater in an inverted well at the base of the crucible. It lost heat primarily by radiation from the outer surfaces of the crucible and the top surface of the salt. The primary independent factors in the model were the emissivity of the crucible (and of the salt) and the fraction of the heater power coupled into the crucible. The model was 'calibrated' using (thermocouple) data and heating power from runs in which the crucible contained no salt.

Sandia has developed and tested mockups armored with W rods over the last decade and pioneered the initial development of W rod armor for International Thermonuclear Experimental Reactor (ITER) in the 1990's. We have also developed 2D and 3D thermal and stress models of W rod-armored plasma facing components (PFCs) and test mockups and are applying the models to both short pulses, i.e. edge localized modes (ELMs), and thermal performance in steady state for applications in C-MOD, DiMES testing and ITER. This paper briefly describes the 2D and 3D models and their applications with emphasis on modeling for an ongoing test program that simulates repeated heat loads from ITER ELMs.

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Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840 MW of fusion power of which 767 MW is in the form of energetic particles (alpha power) and 3073 MW is in the form of neutrons. The alpha plus auxiliary power total 909 MW of which 430 MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.

This paper describes the liquid metal integrated test system (LIMITS) at Sandia National Laboratories. This system was designed to study the flow of molten metals and salts in a vacuum as a preliminary study for flowing liquid surfaces inside of magnetic fusion reactors. The system consists of a heated furnace with attached centrifugal pump, a vacuum chamber, and a transfer chamber for storage and addition of fresh material. Diagnostics include an electromagnetic flow meter, a high temperature pressure transducer, and an electronic level meter. Many ports in the vacuum chamber allow testing the thermal behavior of the flowing liquids heated with an electron beam or study of the effect of a magnetic field on motion of the liquid. Some preliminary tests have been performed to determine the effect of a static magnetic field on stream flow from a nozzle.

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In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion ({approximately}25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of {approximately}22MW/m{sup 2} were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results.

The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed.

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This report describes the instrumentation locations of the Tore Supra Phase III Outboard Limiter, including the locations and signal names of the flowmeters and thermocouples. Shot 11044 was evaluated in some detail. The heat loads in the fourteen cooling tubes that form the limiter head were calculated from the data and the results compared with the heat loads predicted using a 3-D model heat transfer calculation that calculates the distribution of power on the limiter based upon the power scrape-off length, the mag magnetic configuration and the shape of the limiter.

The Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups typically hold a joint meeting each year to provide a forum for discussion of technical issues of current interest as well as an opportunity for program reviews by the Department of Energy (DOE). At the meeting in September 1990, reported here, research programs in support of the International Thermonuclear Experimental Reactor (ITER) were highlighted. The first part of the meeting was devoted to research and development (R&D) for ITER on plasma facing components plus introductory presentations on some current projects and design studies. The balance of the meeting was devoted to program reviews, which included presentations by most of the participants in the Small Business Innovative Research (SBIR) Programs with activities related to plasma wall interactions. The Task Groups on Plasma/Wall Interaction and on High Heat Flux Materials and Components were chartered as continuing working groups by the Division of Development and Technology in DOE`s Magnetic Fusion Program. This report is an addition to the series of ``blue cover`` reports on the Joint Meetings of the Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups. Among several preceding meetings were those in October 1989 and January 1988.

A brief discussion of the following topics is given in this report: Liquid Metal Divertors; Lithium Droplet Beam Divertor; Preferential Pumping of Helium; Reduced Erosion with Cu-Li, W-Li, etc.; Reduction of Erosion by Thermionic Emission; Reduced Erosion in Boronized Graphites; Proposal for Materials Experiments in TRIAM; Carbon-SiC for Plasma Facing Components; Helium Pumping with Palladium; Large Area Pump Limiter; Techniques for Enhanced Heat Removal; New Outlook on Gaseous Divertors; Gaseous Divertor Simulations; Impurity Seeding to Control ITER Particle and Heat Loads; Gaseous Divertor Experiments; Electrical Biasing to Control SOL Particle Fluxes; Biased Limiter in TEXTOR and Biased Divertor in PBX-M; Particle and Heat Flux Control Using Ponderomotive Forces; Helium Exhaust Using ICRF; Ergodic Magnetic Limiter Experiments in JFT-2M; and Helium Exhaust Using Fishbones.

Erosion of POCO graphite by helium in PISCES-A was measured by carbon spectroscopy for a temperature range from 900{degree}-- 2000{degree}C, ion energies of 30--300 eV, ion fluxes of 1 {minus} 6 {times} 10{sup 22} m{sup {minus}2} s{sup {minus}1} and electro temperatures of 4--22 eV. Yields at low energies were higher than predicted in current models. The role of redeposition is discussed. 15 refs., 4 figs.

Ion beam studies on radiation enhanced sublimation (RES) have shown that above 800{degree} C energetic ions incident on graphite produce erosion in the form of carbon atoms with thermal energies and that the erosion rate rises roughly exponentially with temperature. Until recently, the question remained whether RES would scale linearly with flux over three to four orders of magnitude to the plasma edge fluxes in CIT and ITER, where the predicted erosion rates would severely limit the designs for plasma-facing components. Also, RES and carbon self-sputtering may also be involved in the carbon blooms'' observed in TFTR and JET. The data reported here from PISCES, a plasma source at UCLA, are the first RES data at fluxes approaching the plasma edge conditions in a large tokamak and they show little reduction from a direct linear dependence upon flux. Erosion rates measured by weight loss are reported for POCO graphite exposed to helium plasmas for a temperature range from 900--2000{degree} C, ion energies of 30--300 eV, ion fluxes of 1--6 {times} 10{sup 18} cm{sup {minus}2} s{sup {minus}1}, densities of 2--10 {times} 10{sup 12} cm{sup {minus}3} and electron temperatures of 4-10 eV. For these conditions, the amount of redeposition and carbon self-sputtering was minimal. Over 1700{degree} C, there is evidence of electron emission from the sample. 26 refs., 4 figs., 1 tabs.