Office of Fusion Energy Sciences

An international study conducted by technical experts from Europe, Japan, Russia, and the United States has evaluated the technical issues and the required testing facilities for the development of fusion blanket/first-wall systems and has found that some of the key requirements for the engineering feasibility of blanket concepts cannot be established prior to extensive testing in the fusion environment. However, because of availability and low cost, testing in nonfusion facilities (e.g., fission reactors and laboratory experiments) serves a critical role in blanket research and development (R&D) and reduces the risks and costs of the more complex and expensive fusion experiments. A comprehensive analysis shows that the fusion testing requirements for meeting the goal of demonstrating a blanket system availability in DEMO > 50% are as follows: a 1 to 2 MW/m2 neutron wall load, a steady-state plasma operation, a > 10-m2 test area, and a fluence of > 6 MW·yr/m2. This testing fluence includes 1 to 3 MW·yr/m2 for concept performance verification and >4 to 6 MW·yr/m2 for component engineering development and reliability growth/demonstration. Reliability and availability analyses reveal critical concerns that need to be addressed in fusion power development. For a DEMO reactor availability goal of 50%, the blanket availability needs to be ∼80%. For a mean time to recover from a failure of ∼3 months, the mean time between failure (MTBF) for the entire blanket must be >1 yr. For a blanket that has 80 modules, the corresponding MTBF per module is 80 yr. These are very ambitious goals that require an aggressive fusion technology development program. A number of scenarios for fusion facilities were evaluated using a cost/benefit/risk analysis approach. Blanket tests in the International Thermonuclear Experimental Reactor (ITER) alone with a fluence of 1 MW·yr/m2 can address most of the needs for concept verification, but it cannot adequately address the blanket component reliability growth/demonstration testing requirements. An effective path to fusion DEMO is suggested. It involves two parallel facilities: (a) ITER to provide data on plasma performance, plasma support technology, and system integration and (b) a high-volume plasma-based neutron source (HVPNS) dedicated to testing, developing, and qualifying fusion nuclear components and material combinations for DEMO. For HVPNS to be attractive and cost effective, its capital cost must be significantly lower than ITER, and it should have low fusion power (∼150 MW). Exploratory studies indicate the presence of a design window with a highly driven plasma. A testing and development strategy that includes HVPNS would decisively reduce the high risk of initial DEMO operation with a poor blanket system availability and would make it possible – if operated parallel to the ITER basic performance phase – to meet the goal of DEMO operation by the year 2025. Such a scenario with HVPNS parallel to ITER provides substantial savings in the overall R&D cost toward DEMO compared with an ITER-alone strategy. The near-term cost burden is negligible in the context of an international fusion program with HVPNS and ITER sited in two different countries.

A status report of the current U.S. program in magneto-inertial fusion (MIF) conducted by the Office of Fusion Energy Sciences (OFES) of the U.S. Department of Energy is given. Magneto-inertial fusion is an emerging concept for inertial fusion and a pathway to the study of dense plasmas in ultrahigh magnetic fields (magnetic fields in excess of 500 T). The presence of magnetic field in an inertial fusion target suppresses cross-field thermal transport and potentially could enable more attractive inertial fusion energy systems. The program is part of the OFES program in high energy density laboratory plasmas (HED-LP).

The understanding of tokamak transport depends on the exploration of a wide range of theoretical models and of a variety of toroidal experiments. This report considers the contributions that nontokamak, but toroidal, experiments can make to our understanding of tokamak transport as well as theoretical alternatives to the standard drift wave model of tokamak transport.

The corrosion rate of type 316 stainless steel in molten LiF-BeF/sub 2/ (66-34 mole percent) has been measured in a thermal convection loop operating with a maximum temperature of 650/sup 0/C and a temperature difference of 160/sup 0/C. The corrosion rate was correlated with the concentration of impurities in the salt and with the fluoride ion oxidation potential as determined by an on-line voltameter. A corrosion rate of -10 ..mu..m/year was observed initially in the as-received salt. This rate decreased as reactions with initial salt impurities went to completion. Direct addition of beryllium metal to the salt further reduced the corrosion rate.

For a given chord through a plasma, changes in the line integrated index of refraction as a result of a transverse density gradient can be observed by measuring the angle of deflection of a laser beam. In contrast to laser interferometers, this method of density profile measurement places modest requirements on laser quality and alignment procedures, allowing measurements to be conducted with short coherence length commercial laser diodes and segmented photodiode detectors. A prototype implementation of this scheme has been constructed and tested on the compact toroid injection experiment (CTIX). At densities comparable to magnetic fusion plasmas, laser deflections in the nanoradian range were measured. By assuming a particular density profile, a sensitivity of ∼1012cm−3∕nrad was obtained. This produced estimates of CTIX peak density in reasonable agreement with conventional interferometry data. The final goal of this diagnostic is a simple, reliable, array deployable density profile diagnostic.

It is widely agreed that the development of materials for fusion systems requires a high flux, 14 MeV neutron source. The European Union, Japan, Russia and the US have initiated the conceptual design of such a facility. This activity, under the International Energy Agency (IEA) Fusion Materials Agreement, will develop the design for an accelerator-based D-Li system. The first organizational meeting was held in June 1994. This paper describes the system to be studied and the approach to be followed to complete the conceptual design by early 1997.

Accelerator-based radiation sources are ubiquitous tools for imaging and treatment in the fields of medicine and security that save millions of lives, impact billions of dollars of commercial goods annually, and fulfill a critical role in US national security. Today’s commercially available accelerator technology has fallen significantly behind the state-of-the-art. Advancing and transferring state-of-the-art compact accelerator technology into broader use holds the promise of achieving greater control, power, and automation, which can significantly enhance society’s ability to sense and control the world around us. The Workshop identified a wealth of opportunities to advance these technologies from today’s commercial baselines, largely based on half-century old developments, by employing emerging concepts in charged particle acceleration and radiation generation and detection, along with modern ways of thinking about and utilizing developments in systems engineering, advanced materials, supply chain management, manufacturing, advanced computation, energy storage, and artificial intelligence.

The Defect Induced Mix Experiment (DIME-II) will measure the implosion and mix characteristics of CH capsules filled with 5 atmospheres of DT by incorporating mid-Z dopant layers of Ge and Ga. This polar direct drive (PDD) experiment also will demonstrate the filling of a CH capsule at target chamber center using a fill tube. Diagnostics for these experiments include areal x-ray backlighting to obtain early time images of the implosion trajectory and a multiple-monochromatic imager (MMI) to collect spectrally-resolved images of the capsule dopant line emission near bangtime. The inclusion of two (or more) thin dopant layers at separate depths within the capsule shell facilitates spatial correlation of mix between the layers and the hot gas core on a single shot. The dopant layers are typically 2 μm thick and contain dopant concentrations of 1.5%. Three dimensional Hydra simulations have been performed to assess the effects of PDD asymmetry on capsule performance.

An analysis of a scaled-up version of IGNITOR (to a major radius of 2.16 m) is discussed. The design, referred to as DIGNITOR, is a direct extrapolation of IGNITOR. The consequences from the increased size are discussed (mainly due to decreased temperature excursions). A summary of comprehensive calculations of the stresses are presented. The case of a divertor plasma configuration is analyzed. The implications of a CIT-like vacuum vessel are also discussed.

The role of superconducting magnets in Magnetic Fusion Energy Research and Development is assessed from a consideration of program plans and schedules, the present status of the programs and the research and development suggestions arising from recent studies and workshops. A principal conclusion is that the large superconducting magnet systems needed for commercial magnetic fusion reactors can be constructed. However such magnets working under severe conditions, with increasingly stringent reliability, safety and cost restrictions can never be built unless experience is first gained in a number of important installations designed to prove physics and technology steps on the way to commercial power demonstration. The immediate problem is to design a technology program in the absence of definite device needs and specifications, giving a priority weighting to the multiplicity of good, high quality development program suggestions when all proposals cannot be supported.

A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

The Department of Energy (DOE) Office of Science 2003 publication Facilities for the Future of Science: A Twenty-Year Outlook was the first long-range facilities plan prioritized across disciplinary lines ever issued by a government science funding agency anywhere in the world. The Office of Science publication listed 28 proposed facilities, ranking them along two dimensions: scientific priority and technological readiness. This Interim Report provides a summary update on the status of the facilities listed in the Twenty-Year Outlook.

Magnetic confinement research has demonstrated the ability to produce controlled thermonuclear fusion conditions in the laboratory. As a result, fusion research is turning from the question of scientific feasibility to the question of engineering feasibility. The author summarizes the rapid progress that has been made in fusion science and technology, as well as what remains to be done. She emphasizes the progress that has been made in building the unprecedented level of international collaboration that is so important for ensuring rapid progress in the future. She discusses the status of fusion science and development, superconducting magnets, plasma heating and fueling technology, tritium handling, materials development, economics and safety, and fusion reactor design.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench, and current quench. Inclusion of a 1-to 4-eV halo surrounding the main plasma was found to be necessary to match simulation and experimental results for plasma current decay, trajectory, toroidal and poloidal vessel currents, and magnetic probe and flux loop values for the entire VDE.

This study explores the practical limits of helium cooling in a simple geometry unconstrained by a particular conceptual design. Specifically, the configuration was chosen to be an externally heated straight tube considering both uniform heating and heating of half the external parimeter. Both thermal hydraulic and structural limits to the heat flux have been investigated. Curves are presented to show the heat flux and tube length which simultaneously attain both a well temperature and pressure drop/pumping power limit for a range of diameters from 0.05 to 8.0 inches and pressures from 50 to 5000 psia. Tube wall stress limits on heat flux are also shown for the same range of pressure and diameter. These results should serve as an aid in planning more complex concepts as well as evaluating helium cooling in this specific configuration.

Summary form only given. Over the past six years, the Air Force Research Laboratory in Albuquerque, NM has been working in close collaboration with Los Alamos National Laboratory on their field-reversed configuration (FRC) experiment, FRX-L. Through these joint efforts a second experiment has been designed and is now being assembled and tested at the AFRL. This new experiment, which is referred to as the FRC Heating Experiment (FRCHX), has the goal of not only forming a plasma in a field-reversed configuration but of also translating it into an aluminum flux conserving shell (solid liner), where it will be subsequently heated through rapid compression of the liner. The FRC formation portion of FRCHX has been designed to closely match the electrical properties of FRX-L so that FRCs of similar parameters can be formed. Likewise, the translation portion of FRCHX, which has been designed and fabricated concurrently with the new translation section of FRX-L, also closely matches that of FRX-L. The design approach being taken to compressively heat the FRC in the final portion of FRCHX relies on the experimental setup used during two earlier "deformable-contact" vacuum liner experiments that were performed with the Shiva Star Capacitor Bank. In these experiments the liner electrodes had 8-cm-diameter holes on their axes, and both tests were found to be successful in that the ends of the 10-cm diameter, 30-cm long aluminum liner stretched and maintained contact with the electrodes while the body of the liner glided radially inward to implode uniformly. This presentation focuses on the system design and integration of the first two portions of the FRCHX experiment, the FRC formation and translation sections. The performance characteristics of each, as determined by recent test results, are discussed, along with the various magnetic and plasma diagnostics that are being fielded in both sections. Remaining tasks to be accomplished before a complete FRC formation, translation, and compression experiment can be performed are also outlined at the end.

A program of ECH on the new Oak Ridge tokamak, ISX-B, beginning in 1978 is proposed. Experiments will attempt bulk heating, preionization, and current profile control. The first experiments will be a test of the principle with a single 200-kW, 28-GHz gyrotron. It is then proposed to add one additional tube in FY 1979 for high ..beta.. studies, and--assuming successful heating experiments--an additional eight tubes during FY 1979-80 for an ECH flux conserving tokamak experiment. Target date for completion of the first experiments on ISX-B with approx. 200 kW is December 1978 at an estimated cost of $0.9 M. The second tube can be added in FY 1979 at an estimated cost of between $0.4 and $0.7 M, and the experiments with two tubes can proceed starting in the summer of 1979. Additional power up to 2 MW can be added at an estimated total cost of $1.75 to $5.00/W. These cost ranges reflect current technical uncertainties which will be resolved in the early phase(s) of the program. This report presents a design description of the proposed experiments, gives results of some theoretical analyses to predict performance, and includes estimates of costs and schedules.

Sources of Li/sup /minus// and H/sup /minus// ions are needed for diagnostic neutral beam and for current drive in fusion plasmas. Previous efforts to generate Li/sup /minus// beams have focused on electron capture in a gas or production on a low work function surface in a plasma. Volume production of Li/sup /minus// by dissociative attachment of optically pumped lithium molecules has also been studied. This thesis presents the first experimental results for volume production of a Li/sup /minus// ion beam from a plasma discharge. A theoretical model for volume production of Li/sup /minus// ions and separate model for Li/sub 2/ production in the lithium discharge are developed to explain the experimental results. The model is in good agreement with the experiment and shows favorable parameter scalings for further improvement of the Li/sup /minus// ion source. A /sup 6/Li/degree/ diagnostic neutral beam based on this ion source is proposed for measurement of magnetic pitch angle in the International Thermonuclear Experimental Reactor (ITER). Previous efforts in developing H/sup /minus// ion sources have concentrated on volume production in a plasma discharge. Experiments to improve the H/sup /minus// current density from a magnetically filtered multicusp ion source by seeding the discharge with cesium or barium have been conducted. A substantial (> factor of five) increase in H/sup /minus// output is achieved for both cesium and barium addition. Further experiments with barium have shown that the increase is due to H/sup /minus// production on the anode walls. The experiments with cesium are consistent with this formation mechanism. These results show that this new type of 'converterless' surface production H/sup /minus// source provides greatly improved performance when compared to a volume H/sup /minus// source. 92 refs., 47 figs.

The Magnetics and Superconductivity Section has the responsibility for developing superconducting magnet systems for tokamak fusion machines. This is being accomplished by carrying out those research and development needs which will provide the physics understanding and engineering data necessary to design, fabricate, and test large toroidal field (TF) and poloidal field (PF) coils. This information, in addition, supports the Large Coil Program (LCP). A number of design projects have been performed, some in support of other programs and some of a continuing nature. These efforts support the goals and requirements for both the TF and PF magnet systems. Examples are the magnet designs for the EPR, Demo, EBTR, EBT-II, and preliminary scoping for the INS project. The principal effort was expended on the iteration of the EPR Reference Design. Three features of the original reference design--the honeycomb coil structure, the oval coil shape, and the forced-flow cooling of the conductor by supercritical helium--remain as key features of the TF coils. Considerable progress has been made in the theoretical understanding of forced-flow-cooled conductors, and optimized designs with maximum stability margin can be designed to meet specific applications. Experiments which will test the theory are in progress.

This document highlights the technical requirements of the key subsystems of the Reference Design for The Next Step (TNS) and presents a preliminary assessment of the adequacy of the technical capabilities available to satisfy these requirements. Additional information on the Reference Design and the FY 1978 TNS Program activities can be found in the associated technical memoranda, ORNL/TM-6720, ORNL/TM-6721, and ORNL/TM-6723 - ORNL/TM-6733.