Publications

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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.

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. © 2004 Elsevier B.V. All rights reserved.

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. © 2004 Elsevier B.V. All rights reserved.

This paper describes the liquid metal integrated test system (LIMITS) at Sandia National Laboratories1. 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. © 2004 Elsevier B.V. All rights reserved.

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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.

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.

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The development of advanced thermal barrier coatings of yttria stabilized zirconia (YSZ) that exhibit lower thermal conductivity through better control of electron beam-physical vapor deposition (EB-PVD) processing is of prime interest to both the aerospace and power industries. Recently, processing technology was developed to create graded TBCs by coupling ion beam assisted deposition (IBAD) with substrate pivoting in the alumina-YSZ system. The Electron Beam-1200 kW (EB-1200) PVD system was used to deposit a variety of TBC coatings with micron layered microstructures and reduced thermal conductivity of 1.5 W/mK. The use of IBAD produced fully stoichiometric coatings at a reduced substrate temperature of 600 °C and a reduced oxygen background pressure of 0.1 Pa. In addition to the process technology, the results of Direct Simulation Monte Carlo plume modeling and spectroscopic characterization of the PVD plumes are presented. © 2003 Elsevier B.V. All rights reserved.

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.