R.E. Nygren

On the exploration of innovative concepts for fusion chamber technology

Abstract This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load >10 MW/m 2 and surface heat flux >2 MW/m 2 , (2) high power conversion efficiency (>40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid “bare” first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn–Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin (∼2 cm) or thick (∼40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma β and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma–liquid interactions including both plasma–liquid surface and liquid wall–bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W–5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at ∼1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

ALPS–advanced limiter-divertor plasma-facing systems

The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter:divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW:m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com:locate:fusengdes

Plasma–surface interaction issues of an all-metal ITER

We assess key plasma surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor surface, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport, and formation of mixed surface layers, tritium codeposition, core plasma contamination, ELM response, and He on W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma edge regime with convective transport. The HEIGHTS package analyzes divertor plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer wall sputter erosion rates are acceptable for Be (0.3 nm/s) or bare (stainless steel/Fe) wall (0.05 nm/s) for the low duty factor ITER, and are very low (0.002 nm/s) for W. Most wall-sputtered Be is redeposited on the wall itself or baffle region, with about 10% transported to the divertor target. T/Be codeposition in redeposited wall material could be significant (~2 gT per 400 s ITER pulse). Core plasma contamination potential from wall sputtering appears acceptable for Be (~2%), and negligible for W (or Fe) due to near-surface ionization of sputtered W (Fe) atoms and subsequent strong redeposition. A tungsten divertor likewise appears acceptable from the self-sputtering and plasma contamination standpoints, and would have negligible T/W codeposition. Be can grow on/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ~800°C, deleterious Be/W alloy formation may be avoided. ELMs are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using carbon, but critical R&D issues remain, e.g., in areas of transient surface erosion (of all materials), W surface integrity with energetic He etc. bombardment, and in predictive plasma/surface interaction modeling generally. Steps are suggested to ameliorate problems and reduce uncertainties, e.g., via a 300 or 400°C baking capability for T/Be reduction, and using a deposited tungsten first wall test section.

Liquid lithium divertor characteristics and plasma?material interactions in NSTX high-performance plasmas

Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The Liquid Lithium Divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. The LLD consisted of a copper-backed structure with a porous molybdenum front face. Nominal Li filling levels by the end of the run campaign exceeded the porosity void fraction by 150%. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m−2, no macroscopic ejection events were observed. In addition, no substrate line emission was observed after achieving lithium-melt temperatures indicating the lithium wicks and provides a protective coating on the molybdenum porous layer. Impurity emission from the divertor suggests that the plasma is interacting with oxygen-contaminated lithium whether diverted on the LLD or not. A database of LLD discharges is analysed to consider whether there is a net effect on the discharges over the range of total deposited lithium in the machine. Examination of H-97L indicates that performance was constant throughout the run, consistent with the hypothesis that it is the quality of the surface layers of the lithium that impact performance. The accumulation of impurities suggests a fully flowing liquid lithium system to obtain a steady-state PFC on timescales relevant to NSTX.

Runaway electron damage to the Tore Supra Phase III outboard pump limiter

Operation of the Phase III outboard pump limiter (OPL) in Tore Supra in 1994 was terminated prematurely when runaway electrons during the current decay following a disruption pierced leading edge tube on the electron side and caused a water leak. The location, about 20 mm outside the last closed flux surface during normal operation, and the infrared (IR) images of the limiter indicate that the runaways moved in large outward steps, i.e. tens of millimeters, in one toroidal revolution. For plasma (runaway) currents in the range of 155 to 250 kA, the drift orbits open to the outside. Basic trajectory computations suggest that such motion is possible under the conditions present for this experiment. Activation measurements made on sections of the tube to indicate the area of local damage are presented here. An understanding of this event may provide important guidance regarding the potential damage from runaways in future tokamaks.

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Abstract Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W–5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. Systems study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kW h. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

High heat flux tests on heat sinks armored with tungsten rods

Abstract This paper presents the results of high heat flux tests at Sandia National Laboratories of mock-ups armored with W rods embedded in water-cooled copper-alloy heat sinks. The major result is the excellent performance of these mock-ups in thermal response tests at up to 30 MW/m2 and in thermal cycling tests of 500 cycles (10 s on, 10 s off) at ∼25 MW/m2. Melting of rod tips and the subsequent ‘self-repair’ of a degraded thermal bond were seen. Issues for further development in both testing and manufacturing of the mock-ups were identified. Evaluation of the surface heat flux was found to be somewhat problematic and this issue is discussed.

Ultra Low Pressure-Drop Helium-Cooled Porous-Tungsten PFC

Abstract A new class of helium-cooled high heat-flux plasma facing heat exchanger (HX) concept is presented. These unique “Foam-In-Tube” HX concepts are composed of a thin tungsten shell integrally bonded to an open-cell tungsten foam core. High heat flux tests show maximum heat loads of 22.4 MW/m2 using 4 MPa helium at a flow rate of 27 g/s. Based on these impressive performance results, a unique and scalable heat exchanger channel with ultra-low pressure drop through the porous foam is presented. The primary advantage of the new concept is that pressure drop through the porous media and structure temperatures are nearly independent of HX tube length. The concept is modular in design and can be combined to meet divertor size requirements. From a manufacturing and reliability point of view, the advantage of the proposed concept is that it minimizes the need for joining to other functional materials.

Effects of supra-thermal particle impacts on Tore Supra plasma facing components

Abstract Actively cooled plasma facing components (PFCs) for Tore Supra (TS) have been designed basically for heat exhaust of ‘normal’ (convected and radiated) plasma power. However, in some cases, fast particles have been observed, which locally increased the power flux density, leading to damage of these PFCs and other inner vessel components. Three different examples for irreversible component damage, such as component melting and water leaks, are described involving runaway and supra-thermal particle strikes. In view of the capability for TS to handle larger input powers and to control the particles over long pulse durations, inner vessel components have been completely redesigned. The improved design concepts retained for the CIEL upgrade and preliminary results in the new configuration are presented.

Assessing braze quality in the actively cooled Tore Supra phase III outboard pump limiter

Abstract The quality of brazing of pyrolytic graphite armor brazed to copper tubes in Tore Supras phase III outboard pump limiter was assessed through pre-service qualification testing of individual copper-tile assemblies. The evaluation used non-destructive hot-water transient heating tests performed in the high temperature, high pressure flow loop at Sandias Plasma Materials Test Facility. The surface temperatures of tiles were monitored with an IR camera as water at 120 °C at about 2.07 MPa (300 lbf in −2 ) passed through a tube assembly initially at 30 °C. For tiles with braze voids or cracks, the surface temperatures lagged behind those of adjacent well-bonded tiles. Temperature lags were correlated with flaw sizes observed during repairs based upon detailed two-dimensional heat transfer analyses. “Bad” tiles, i.e. temperature lags of 10–20 °C depending upon a tiles size, were easy to detect and, when removed, revealed braze voids of roughly 50% of the joint area. 11 of the 14 tubes were rebrazed after bad tiles were detected and removed. Three tubes were rebrazed twice.