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We report on the detection of the time-dependent B-field amplitude and topology in a laser-driven solenoid. The B-field inferred from both proton deflectometry and Faraday rotation ramps up linearly in time reaching 210 ± 35 T at the end of a 0.75-ns laser drive with 1 TW at 351 nm. A lumped-element circuit model agrees well with the linear rise and suggests that the blow-off plasma screens the field between the plates leading to an increased plate capacitance that converts the laser-generated hot-electron current into a voltage source that drives current through the solenoid. ALE3D modeling shows that target disassembly and current diffusion may limit the B-field increase for longer laser drive. Scaling of these experimental results to a National Ignition Facility (NIF) hohlraum target size (∼0.2cm^{3}) indicates that it is possible to achieve several tens of Tesla.

Dynamic stabilization of the ablative Rayleigh–Taylor instability of a heavy ion fusion target induced by a beam wobbling system is studied. Using a sharp-boundary model and Courant–Synder theory, it is shown, with an appropriately chosen modulation waveform, that the instability can be stabilized in certain parameter regimes. It is found that the stabilization effect has a strong dependence on the modulation frequency and the waveform. Modulation with frequency comparable to the instability growth rate is most effective in terms of stabilizing the instability. A modulation with two frequency components can result in a reduction of the growth rate larger than the sum of that due to the two components when applied separately.

Recent heavy ion fusion target studies show that it is possible to achieve ignition with direct drive and energy gain larger than 100 at 1 MJ. To realize these advanced, high-gain schemes based on direct drive, it is necessary to develop a reliable beam smoothing technique to mitigate instabilities and facilitate uniform deposition on the target. The dynamics of the beam centroid can be explored as a possible beam smoothing technique to achieve a uniform illumination over a suitably chosen region of the target. The basic idea of this technique is to induce an oscillatory motion of the centroid for each transverse slice of the beam in such a way that the centroids of different slices strike different locations on the target. The centroid dynamics is controlled by a set of biased electrical plates called “wobblers.” Using a model based on moments of the Vlasov-Maxwell equations, we show that the wobbler deflection force acts only on the centroid motion, and that the envelope dynamics are independent of the wobbler fields. If the conducting wall is far away from the beam, then the envelope dynamics and centroid dynamics are completely decoupled. This is a preferred situation for the beam wobbling technique, because the wobbler system can be designed to generate the desired centroid motion on the target without considering its effects on the envelope and emittance. A conceptual design of the wobbler system for a heavy ion fusion driver is briefly summarized.

Plasmas are a source of unbound electrons for charge neutralizing intense heavy ion beams to focus them to a small spot size and compress their axial length. The plasma source should operate at low neutral pressures and without strong externally-applied fields. To produce long plasma columns, sources based upon ferroelectric ceramics with large dielectric coefficients have been developed. The source utilizes the ferroelectric ceramic BaTiO{sub 3} to form metal plasma. The drift tube inner surface of the Neutralized Drift Compression Experiment (NDCX) is covered with ceramic material. High voltage ({approx} 8 kV) is applied between the drift tube and the front surface of the ceramics. A BaTiO{sub 3} source comprised of five 20-cm-long sources has been tested and characterized, producing relatively uniform plasma in the 5 x 10{sup 10} cm{sup -3} density range. The source was integrated into the NDCX device for charge neutralization and beam compression experiments, and yielded current compression ratios {approx} 120. Present research is developing multi-meter-long and higher density sources to support beam compression experiments for high energy density physics applications.

Plasmas are a source of unbound electrons for charge neutralizing intense heavy ion beams to allow them to focus to a small spot size and compress their axial pulse length. The plasma source should be able to operate at low neutral pressures and without strong externally applied electric or magnetic fields. To produce 1m-long plasma columns, sources based upon ferroelectric ceramics with large dielectric coefficients are being developed. The sources utilize the ferroelectric ceramic BaTiO3 to form metal plasma. The drift tube inner surface of the Neutralized Drift Compression Experiment (NDCX) will be covered with ceramic material, and high voltage (! 7kV) will be applied between the drift tube and the front surface of the ceramics. A prototype ferroelectric source, 20cm in length, has produced plasma densities of 5" 10 11 cm # 3 . It was integrated into the Neutralized Transport Experiment (NTX), and successfully charge neutralized the K + ion beam. A 1m-long source comprised of five 20-cm-long sources has been tested. Simply connecting the five sources in parallel to a single pulse forming network power supply yielded non-uniform performance due to the time-dependent nature of the load that each of the five plasma sources experiences. Other circuit combinations have been considered, including powering each source by its own supply. The 1-m-long source has now been successfully characterized, producing relatively uniform plasma over the 1m length of the source in the mid-10 10 cm # 3 density range. This source will be integrated into the NDCX device for charge neutralization and beam

Critical issues and ion beam requirements are explored for fast ignition using ion beams to provide fuel compression using indirect drive and to provide separate short-pulse ignition heating using direct drive. Several ion species with different hohlraum geometries are considered for both accelerator-produced and laser-produced ion ignition beams. Ion-driven fast ignition targets are projected to have modestly higher gains than with conventional heavy ion fusion and may offer some other advantages for target fabrication and for use of advanced fuels. However, much more analysis and additional experiments are needed before conclusions can be drawn regarding the feasibility for meeting the ion beam transverse and longitudinal emittances, focal spots, pulse lengths, and target standoff distances required for ion-driven fast ignition.

Highly ionized plasmas are being employed as a medium for charge neutralizing heavy ion beams in order to focus to a small spot size. Calculations suggest that plasma at a density of 1–100 times the ion beam density and at a length ∼0.1–1 m would be suitable for achieving a high level of charge neutralization. A radio frequency (RF) source was constructed at the Princeton Plasma Physics Laboratory (PPPL) in support of the joint Neutralized Transport Experiment (NTX) at the Lawrence Berkeley National Laboratory (LBNL) to study ion beam neutralization. Pulsing the source enabled operation at pressures ∼10 −6 Torr with plasma densities of 10 11 cm −3 . Near 100% ionization was achieved. The plasma was 10 cm in length, but future experiments require a source 1 m long. The RF source does not easily scale to the length. Consequently, large-volume plasma sources based upon ferroelectric ceramics are being considered. These sources have the advantage of being able to increase the length of the plasma and operate at low neutral pressures. The source will utilize the ferroelectric ceramic BaTiO 3 to form metal plasma. A 1 m long section of the drift tube inner surface of NTX will be covered with ceramic. A high voltage (∼1–5 kV) is applied between the drift tube and the front surface of the ceramic by placing a wire grid on the front surface. Plasma densities of 10 12 cm −3 and neutral pressures ∼10 −6 Torr are expected. A test stand to produce 20 cm long plasma is being constructed and will be tested before a 1 m long source is developed.

There has been rapid progress in inertial fusion in the past few years. This progress spans the construction of ignition facilities, a wide range of target concepts and the pursuit of integrated programmes to develop fusion energy using lasers, ion beams and z-pinches.Two ignition facilities are under construction, the national ignition facility (NIF) in the United States and the laser megajoule (LMJ) in France, and both projects are progressing towards an initial experimental capability. The laser integration line prototype beamline for LMJ and the first four beams of NIF will be available for experiments in 2003. The full 192 beam capability of NIF will be available in 2009 and ignition experiments are expected to begin shortly after that time.There is steady progress in target science and target fabrication in preparation for indirect-drive ignition experiments on NIF. Advanced target designs may lead to 5–10 times more yield than initial target designs. There has also been excellent progress on the science of ion beam and z-pinch-driven indirect-drive targets.Excellent progress on direct-drive targets has been obtained on the Omega laser at the University of Rochester. This includes improved performance of targets with a pulse shape predicted to result in reduced hydrodynamic instability. Rochester has also obtained encouraging results from initial cryogenic implosions.There is widespread interest in the science of fast ignition because of its potential for achieving higher target gain with lower driver energy and relaxed target fabrication requirements. Researchers from Osaka have achieved outstanding implosion and heating results from the Gekko XII Petawatt facility and implosions suitable for fast ignition have been tested on the Omega laser.A broad-based programme to develop lasers and ion beams for inertial fusion energy (IFE) is under way with excellent progress in drivers, chambers, target fabrication and target injection. KrF and diode pumped solid-state lasers are being developed in conjunction with dry-wall chambers and direct-drive targets. Induction accelerators for heavy ions are being developed in conjunction with thick-liquid protected wall chambers and indirect-drive targets.

Highly ionized plasmas are being considered as a medium for charge neutralizing heavy ion beams in order to focus beyond the space-charge limit. Calculations suggest that plasma at a density of 1–100 times the ion beam density and at a length ∼0.1–2 m would be suitable for achieving a high level of charge neutralization. An Electron Cyclotron Resonance (ECR) source has been built at the Princeton Plasma Physics Laboratory (PPPL) to support a joint Neutralized Transport Experiment (NTX) at the Lawrence Berkeley National Laboratory (LBNL) to study ion beam neutralization with plasma. The ECR source operates at 13.6 MHz and with solenoid magnetic fields of 1–10 gauss. The goal is to operate the source at pressures ∼10−6 Torr at full ionization. The initial operation of the source has been at pressures of 10−4–10−1 Torr. Electron densities in the range of 108 to 1011 cm−3 have been achieved. Low-pressure operation is important to reduce ion beam ionization. A cusp magnetic field has been installed to improve radial confinement and reduce the field strength on the beam axis. In addition, axial confinement is believed to be important to achieve lower-pressure operation. To further improve breakdown at low pressure, a weak electron source will be placed near the end of the ECR source. This article also describes the wave damping mechanisms. At moderate pressures (> 1 mTorr), the wave damping is collisional, and at low pressures (< 1 mTorr) there is a distinct electron cyclotron resonance.

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter dated September 10, 2002 from Dr. Ray Orbach, Director of the DOE's Office of Science. In that letter, Dr. Orbach asked FESAC to develop a plan with the end goal of the start of operation of a demonstration power plant in approximately 35 years. This report, submitted March 5, 2003, presents such a plan, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets. The report also responds to a request from DOE to FESAC to describe what new or upgraded fusion facilities will “best serve our purposes” over a time frame of the next twenty years.

The Innovative Confinement Concepts Workshop, ICC2002, provided a forum for presentations and exchange of ideas on the science and status of innovative concepts in the U.S. Fusion Energy Program. The workshop, held at the University of Maryland on January 22-24, 2002, included oral presentations addressing the important science and status of the concepts, posters focussed on details of the work, a skunkworks for novel ideas, and breakout sessions preparing for the July 2002 fusion energy Snowmass meeting. This report summarizes the oral presentations. A web site (https:// wormhole.ucllnl.org/ICC2002/) has been established with the abstracts and many of the presentations, both oral and poster, from the workshop.

Research has been initiated in the U.S. Virtual Laboratory for Technology on IFE chamber and target technologies. The critical issues, and the approaches taken to address these issues, are discussed.

This report of the Integrated Program Planning Activity (IPPA) has been prepared in response to a recommendation by the Secretary of Energy Advisory Board that, “Given the complex nature of the fusion effort, an integrated program planning process is an absolute necessity.” We therefore undertook this activity to integrate the various elements of the program, to improve communication and performance accountability across the program, and to show the interconnectedness and interdependency of the diverse parts of the national Fusion Energy Sciences Program. This report is based on the September 1999 Fusion Energy Sciences Advisory Committee's (FESAC) report “Priorities and Balance within the Fusion Energy Sciences Program.” In its December 5, 2000, letter to the Director of the Office of Science, the FESAC reaffirmed the validity of the September 1999 report and stated that the IPPA presents a framework and process to guide the achievement of the 5-year goals listed in the 1999 report. The report also outlines a process for establishing a database for the fusion research program that will indicate how each research element fits into the overall program. This database will also include near-term milestones associated with each research element and will facilitate assessments of the balance within the program at different levels.

A newly proposed type of multicharged ion source offers the possibility of an economically advantageous high-charge-state fusion driver. Multiphoton absorption in an intense uniform laser focus can give multiple charge states of high purity, simplifying or eliminating the need for charge-state separation downstream. Very large currents (hundreds of amperes) can be extracted from this type of source. Several arrangements are possible. For example, the laser plasma could be tailored for storage in a magnetic bucket, with beam extracted from the bucket. A different approach, described in this report, is direct beam extraction from the expanding laser plasma. We discuss extraction and focusing for the particular case of a 4.1 MV beam of Xe 16+ ions. The maximum duration of the beam pulse is limited by the total charge in the plasma, while the practical pulse length is determined by the range of plasma radii over which good beam optics can be achieved. The extraction electrode contains a solenoid for beam focusing. Our design studies were carried out first with an envelope code and then with a self-consistent particle code. Results from our initial model showed that hundreds of amperes could be extracted, but that most of this current missed the solenoid entrance or was intercepted by the wall and that only a few amperes were able to pass through. We conclude with an improved design which increases the surviving beam to more than 70 A.

Summary form only given. Heavy ion beams are suitable for heating matter in the studies of high energy density physics and inertial fusion. To obtain the required rapid target heating and high beam intensity, the beam has to be longitudinally compressed and radially focused. The former can be achieved by neutralized drift compression which has been demonstrated by experiments [P. K. Roy et al., Phys. Rev. Lett. 95, 234801 (2005)]. After imparting a head-to-tail velocity gradient to the beam, the beam compresses as it drifts in a dense background plasma where the effects of space charge are neutralized. For low energy beams, the beam velocity is ramped by induction modules with tailored voltage waveforms. This method is, however, technically difficult and highly expensive in the high energy regime. A physics design is introduced to generate a velocity tilt for high energy beams at a low cost. By sweeping beam ions transversely with a time-dependent dipole upstream of a wedge, different slices of the beam pass through matter of variable thickness. The energy loss creates a velocity tilt for the drifting beam which subsequently compresses within a dense background plasma. Using stopping power models and particle codes, once the target position is given, the shape of the wedge can be designed accordingly to give maximum compression at the target. Simulations have been conducted using typical beam parameters of FAIR at GSI and HIAF at IMP Lanzhou. The results show that energy straggling has negligible effect on the bunching factor, whereas beam ion scattering and fragmentation within the wedge depend on the type of beam ion, beam energy and wedge material. A decisive factor affecting the current amplification at the target is the ratio between the head-to-tail sweep length and the spot size on the wedge. Based on these results, an experiment that demonstrates longitudinal compression using a wedge is proposed. The wedge is also incorporated into the design of a full system that delivers both transverse focusing and temporal compression. Numerical studies are employed to determine the time-dependent parameters of the dipole and quadrupole elements. Simulation results are presented and the constraints and trade-offs in the design are discussed.

Development of inertial fusion energy (IFE) will require continued R&D in target physics, driver technology, target production and delivery systems, and chamber technologies. It will also require the integration of these technologies in tests and engineering demonstrations of increasing capability and complexity. Development needs in each of these areas are discussed. It is shown how IFE development will leverage off the DOE Defense Programs funded inertial confinement fusion (ICF) work.

The economy of scale for multiunit inertial fusion energy (IFE) power plants is explored based on the molten salt HYLIFE-II fusion chamber concept, for the purpose of producing lower cost electricity and hydrogen fuel. The cost of electricity (CoE) is minimized with a new IFE systems code IFEFUEL5 for a matrix of plant cases with one to eight fusion chambers of 250 to 2000-MW (electric) net output each, sharing a common heavy-ion driver and target factory. Improvements to previous HYLIFE-II models include a recirculating induction linac driver optimized as a function of driver energy and rep-rate (average driver power), inclusion of beam switchyard costs, a fusion chamber cost scaling dependence on both thermal power and fusion yield, and a more accurate bypass pump power scaling with chamber rep-rate. A CoE less than 3 cents/kW(electric)-h is found for plant outputs greater than 2 GW(electric), allowing hydrogen fuel production by wafer electrolysis to provide lower fuel cost per mile for higher efficiency hydrogen engines compared with gasoline engines. These multiunit, multi-GW(electric) IFE plants allow staged utility plant deployment, lower optimum chamber rep-rates, less sensitivity to driver and target fabrication costs, and a CoE possibly lower than future fission, fossil, and solar competitors.more » 37 refs., 12 figs., 4 tabs.« less

Dynamic stabilization of the ablative Rayleigh-Taylor instability of a heavy ion fusion target induced by a beam wobbling system is studied. Using a sharp-boundary model and Courant-Synder theory, it is shown, with an appropriately chosen modulation waveform, that the instability can be sta- bilized in certain parameter regimes. It is found that the stabilization e ect has a strong dependence on the modulation frequency and the waveform. Modulation with frequency comparable to the instability growth rate is the most e ective in terms of stabilizing the instability. A modulation with two frequency components can result in a reduction of the growth rate larger than the sum of that due to the two components when applied separately.

This study evaluates the potential performance (efficiency and cost) of inertial fusion reactors assumed capable of vaporizing blankets of various working materials to a temperature (10,000-20,000 K) suitable for economical MHD conversion in a Compact Fusion Advanced Rankine II (CFARII) power cycle. Using a conservative model, 1-D neutronics calculations of the fraction of fusion yield captured as a function of the blanket thickness of Flibe, lithium and lead—lithium blankets are used to determine the optimum blanket thicknesses for each material to minimize CoE for various assumed fusion yields, “generic” driver costs, and target gains. Lithium-hydride blankets are also evaluated using an extended neutronics model. Generally optimistic (“advanced”) combinations of lower driver cost/joule and higher target gain are assumed to allow high enough fusion yields to vaporize and ionize target blankets thick enough to stop most 14 MeV neutrons, and to breed tritium. A novel magnetized, prestressed reactor chamber concept is modeled together with previously developed models for the CFARII Balance-of-Plant (BoP), consisting of a supersonic plasma jet, MHD generator, and “raindrop” condensor. High fusion yields (20 to 80 GJ) are found necessary to heat and ionize the Flibe, lithium, and lead-lithium blankets for MHD conversion, with initial solid thicknesses sufficient to capture most of the fusion yield. Much smaller fusion yields (1 to 20 GJ) are required for lithium-hydride blankets. For Flibe, lithium, and lead—lithium blankets, improvements in target gain and/or driver cost/joule, characterized by a “Bang per Buck” figure-of-merit of ≥ 20 joules yield per driver $, would be required for competitive CoE, while a figure-of-merit of ≥ 1 joule yield per driver $ would suffice for lithium-hydride blankets. Advances in targets/driver costs would benefit any IFE reactor, but the very low CFARII BoP costs (contributing only 3 mills/kWh to CoE) allows this type of reactor, given sufficient advances that non-driver costs dominate, to ultimately produce electricity at a much lower cost than any current nuclear plant.

At the 14th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, Germany, several senior experts on magnetic fusion development shared with me their view that finally, the world fusion effort has accumulated enough understanding of the requirements and conditions for ignited long-pulse burn, that a tokamak experiment to achieve that objective could be built and expected to work with high confidence. Although this is not yet a unanimous view of all the IAEA participants, I do agree with this majority view. As a reflection of this growing confidence, the governments of the United States, the European Community, Russia and Japan have recently commissioned the Engineering Design Activity (EDA) phase of the International Thermonuclear Experimental Reactor (ITER). The technical vision of ITER presented at that conference by the new ITER director, Paul-Henri Rebut, was widely accepted as a most practical and workable approach to achieving the above objective, assuming a successful demonstration of the proposed gas-target divertor during the EDA R&D phase. Having pioneered an early version of the gas-target divertor [l], I also agree with this assessment of the technical viability of Rebut’s proposed design. At the end of his ITER presentation, however, Rebut touched on an issue of much lower confidence among the fusion community, namely, that ITER would demonstrate “a reasonable cost per unit of fusion power”. Estimating myself that the magnet set (TF + PF) of his proposed design would store about 80 GJ, would have a cold mass exceeding 13000 tons, including magnet support structure, and would alone cost more than 5 B $ at 4OO$/kg (for 13 T max, Nb,Sn), I take a charitable interpretation of Rebut’s remark to mean that ITER demonstrate a reasonable unit cost, not by itself alone, but viewed as a significant step toward a practical power reactor. This then shifts the confidence issue on economics to a judgment on how much improvement in unit cost can be reasonable expected in tokamak reactors beyond ITER. Improvement beyond ITER was an important motivation behind the recent recommendation by the US Secretary’s Energy Advisory Board (SEAB) to proceed with the tokamak Physics Experiment (TPX) at the Princeton Plasma Physics Laboratory. This prompts the question as to how much improvement in unit cost would be needed from the present embodiment of ITER to an economically-competitive power reactor, assuming a 1 GW, power reactor would be at least no larger in size or mass than ITER. I would estimate the present expectation of ITER’s unit cost, based on a “median” performance with operation at prudent margins below accepted limits, to be = 10 B $/l GW,,, or, equivalently, 25$/W,, as if hot blankets and a 40% efficient conversion system were included. For a “competitive” commercial power reactor, I would take 2.5$/W, (based on total capital as in the ITER number). Thus, by this figuring, I estimate than an improvement factor of about 10 is needed (I wouldn’t quibble between 7 and 14).

The centroid and envelope dynamics of a high-intensity charged-particle beam are investigated as a beam smoothing technique to achieve uniform illumination over a suitably chosen region of the target for applications to ion-beam-driven high energy density physics and heavy ion fusion. The motion of the beam centroid projected onto the target follows a smooth pattern to achieve the desired illumination, for improved stability properties during the beam-target interaction. The centroid dynamics is controlled by an oscillating "wobbler," a set of electrically biased plates driven by rf voltage.

The charge to Panel 3 was to look at the idea of concept improvements in the context of US DOE management of the magnetic fusion program. The panel suggested that if DOE were commited to the idea of concept improvement, it needed to overcome the existing impression it was not receptive to new ideas. In part the long time scale for development of fusion energy, coupled with the rate of change of scientific programs and research based on emerging knowledge, means that in general the program will be much different ten to twenty years in the future. To be able to meet this changing direction, the US program must maintain an openness to look at promising alternative ideas, spend money on developing the ideas, and consider funding some to intermediate development levels. Stellerator research was offered as one alternative to consider in light of present international work. The panel urged supporting the development of new concepts and ideas, as well as continued support for plasma physics basic research.

This paper examines the extensions of the physics and engineering guidelines for the International Thermonuclear Experimental Reactor (ITER) device needed for acceptable operating points for a steady-state tokamak power reactor. Noninductive current drive is provided in steady state by high-energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma, and bootstrap current. Three different levels of extension of the ITER physics/engineering guidelines, with differing assumptions on the possible plasma beta, elongation, and aspect ratio, are considered for power reactor applications. Plasma gain Q = fusion power/input power in excess of 20 and average neutron wall fluxes from 2.3 to 3.6 MW/m{sup 2} are predicted in devices with major radii varying from 7.0 to 6.0 m and aspect ratios from 2.9 to 4.3.

A new divertor configuration is suggested as a possible solution to the problems of high heat flux and erosion at the divertors in large high-power tokamaks. The proposed configuration is a toroidally symmetrical slot in the divertor that allows part of the edge plasma and most of its power to enter a cavity in a thin annular sheet. The large surface area of the sheet is exposed to interaction with gas in the cavity. This results in radiation and a reflux of fast neutral atoms, both of which transport power to the cavity walls. The heat flux is reduced because the power is spread over a much larger area. Erosion due to sputtering is also reduced because the decreased power flux reduces the sheath potential and, therefore, the average ion impact energy. Sputtering by fast neutrals should not be a serious problem because neutrals are not accelerated by a sheath as are ions. Helium ash and impurity atoms that are ionized within the cavity tend to be trapped there by the electric field that must exist...

In heavy-ion inertial-confinement fusion systems, intense beams of ions must be transported from the exit of the final-focus magnet system through the fusion chamber to hit spots on the target with radii of about 2 mm. For the heavy-ion-fusion power-plant scenarios presently favored in the U.S., a substantial fraction of the ion-beam space charge must be neutralized during this final transport. The most effective neutralization technique found in numerical simulations is to pass each beam through a low-density plasma after the final focusing. To provide quantitative comparisons of these theoretical predictions with experiment, the Virtual National Laboratory for Heavy Ion Fusion has completed the construction and has begun experimentation with the neutralized-transport experiment. The experiment consists of three main sections, each with its own physics issues. The injector is designed to generate a very high-brightness, space-charge-dominated potassium beam, while still allowing variable perveance by a beam aperturing technique. The magnetic-focusing section, consisting of four pulsed quadrupoles, permits the study of magnet tuning, as well as the effects of phase-space dilution due to higher-order nonlinear fields. In the final section, the converging ion beam exiting the magnetic section is transported through a drift region with plasma sources for beam neutralization, and the final spot size is measured under various conditions of neutralization. In this paper, we discuss the design and characterization of the three sections in detail and present initial results from the experiment.

In December 1998, LBNL Director Charles Shank and LLNL Director Bruce Tarter signed a Memorandum of Agreement to create the Heavy-Ion Fusion Virtual National Laboratory (HIF-VNL) with the purpose of improving the efficiency and productivity of heavy ion research through coordination of the two laboratories' efforts under one technical director. In 1999, PPPL Director Robert Goldston signed the VNL MOA for PPPL's heavy-ion fusion group to join the VNL. LBNL and LLNL each contribute about 45% of the $10.6 M/yr trilab VNL effort, and PPPL contributes currently about 10% of the VNL effort. The three labs carry out collaborative experiments, theory and simulations of a variety of intense beam scientific issues described below. The tri-lab HIF VNL program is part of the DOE Office of Fusion Energy Sciences (OFES) fusion program. A short description of the four major tasks areas of HIF-VNL research is given in the next section. The task areas are: High Current Experiment, Final Focus/Chamber Transport, Source/Injector/Low Energy Beam Transport (LEBT), and Theory/Simulation. As a result of the internal review, more detailed reviews of the designs, costs and schedules for some of the tasks have been completed, which will provide more precision in the scheduled completion dates of tasks. The process for the ongoing engineering reviews and governance for the future management of tasks is described in section 3. A description of the major milestones and scientific deliverables for flat guidance budgets are given in section 4. Section 5 describes needs for enabling technology development for future experiments that require incremental funding.

Inertial Fusion Energy (IFE) is the primary fusion approach alternative to the Tokamak in the U.S., with a significant IFE driver development program on heavy-ion accelerators supported by DOE Energy Research, and with a large inertial confinement fusion target physics program supported by DOE Defense Programs, aimed at the goal of fusion ignition and energy gain in the laboratory with the National Ignition Facility (NIF) around 2005.

The use of a 1.8 MJ glass laser in both direct and indirect drive fusion ignition is discussed in reference to the future experiments. Methods of laser pumping, target physics and design are reviewed.

A few % wobbling-beam illumination nonuniformity is realized in heavy ion inertial confinement fusion (HIF) by a spiraling beam axis motion. So far the wobbling heavy ion beam (HIB) illumination was proposed to realize a uniform implosion in HIF. However, the initial imprint of the wobbling HIBs introduces a large unacceptable energy deposition nonuniformity. In the wobbling HIBs illumination, the illumination nonuniformity oscillates in time and space. The oscillating-HIB energy deposition may contribute to the reduction of the HIBs’ illumination nonuniformity. The wobbling HIBs can be generated in HIB accelerators and the oscillating frequency may be several 100MHz ∼1GHz. Three-dimensional HIBs illumination computations presented here show that the few % wobbling HIBs illumination nonuniformity oscillates successfully with the same wobbling HIBs frequency.

Recently [E. Henestroza et al., Phys. Plasmas 18, 032702 (2011)], a new inertial-fusion target configuration, the X-target, using one-sided axial illumination has been explored. This class of target uses annular and solid-profile heavy ion beams to compress and ignite deuterium-tritium (DT) fuel that fills the interior of metal cases that have side-view cross sections in the shape of an “X.” X-targets using all-DT-filled metal cases imploded by three annular ion beams resulted in fuel densities of ∼50 g/cm3 at peak compression, and fusion gains of ∼50, comparable to heavy ion driven hohlraum targets [D. A. Callahan-Miller and M. Tabak, Phys. Plasmas 7, 2083 (2000)]. This paper discusses updated X-target configurations that incorporate inside the case a propellant (plastic) and a pusher (aluminum) surrounding the DT fuel. The updated configurations are capable of assembling higher fuel areal densities ∼2 g/cm2 using two annular beams to implode the target to peak DT densities ∼100 g/cm3, followed by a fast-ignition solid ion beam which heats the high-density fuel to thermonuclear temperatures in ∼200 ps to start the burn propagation, obtaining gains of ∼300. These targets have been modeled using the radiation-hydrodynamics code HYDRA [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] in two- and three- dimensions to study the properties of the implosion as well as the ignition and burn propagation phases. At typical Eulerian mesh resolutions of a few microns, the aluminum-DT interface shows negligible Rayleigh–Taylor (RT) and Richtmyer–Meshkov instability growth; also, the shear flow of the DT fuel as it slides along the metal X-target walls, which drives the RT and Kelvin Helmholtz instabilities, does not have a major effect on the burning rate. An analytic estimate of the RT instability process at the Al-DT interface shows that the aluminum spikes generated during the pusher deceleration phase would not reach the ignition zone in time to affect the burning process. Also, preliminary HYDRA calculations, using a higher resolution mesh to study the shear flow of the DT fuel along the X-target walls, indicate that metal-mixed fuel produced near the walls would not be transferred to the DT ignition zone (at maximum ρR) located at the vertex of the X-target.

It is presented an extended summary for a paper describing: a) the motivation of the inertial fusion in general and particularly for the production of the hydrogen fuel powering low‐emission vehicles, b) the general requirements for fusion electric plants, c) a comparative economic analysis concerning the design of drivers and target chambers. (AIP)

The HYDRA radiation-hydrodynamics code [M. M. Marinak et al., Phys. Plasmas 8, 2275 (2001)] is used to explore one-sided axial target illumination with annular and solid-profile uranium ion beams at 60 GeV to compress and ignite deuterium-tritium fuel filling the volume of metal cases with cross sections in the shape of an “X” (X-target). Quasi-three-dimensional, spherical fuel compression of the fuel toward the X-vertex on axis is obtained by controlling the geometry of the case, the timing, power, and radii of three annuli of ion beams for compression, and the hydroeffects of those beams heating the case as well as the fuel. Scaling projections suggest that this target may be capable of assembling large fuel masses resulting in high fusion yields at modest drive energies. Initial two-dimensional calculations have achieved fuel compression ratios of up to 150X solid density, with an areal density ρR of about 1 g/cm2. At these currently modest fuel densities, fast ignition pulses of 3 MJ, 60 GeV, 50 ps, a...

This presentation discusses the motivation for inertial fusion energy, a brief synopsis of five recently-completed inertial fusion power plant designs, some general conclusions drawn from these studies, and an exmaple of an IEE hydrogen synfuel plant to suggest that future fusion studies consider broadening fusion use to low-emission fuels production as well as electricity.

A newly proposed type of multicharged ion source has several potential advantages over existing types and a number of useful applications. The basic principle is that multiphoton absorption in an intense uniform laser focus can give multiple charge states of high purity (Ref. Reference 1). Thus, charge state separation downstream is simplified or made unnecessary. Another advantage is that large currents (hundreds of amperes) can be extracted. This type of source could be used for heavy-ion fusion drivers (see Ref. Reference 1) or storage rings. There are also industrial application such as materials processing. We describe conceptual design studies for several specific cases. For example, we discuss extraction and focusing of a 4.1 MV, 144 A beam of Xe16+ ions from an expanding plasma created by an intense laser. The maximum duration of the beam pulse is determined by the total charge in the plasma, while the practical pulse length is determined by the range of plasma radii over which good beam optics ca...

The Office of Fusion Energy discusses the completion of the TFTX Preconceptual Design work. It is proposed that a STARFIRE II tokamak power reactor design is in want of initiation, and they are now trying to determine what the boundaries should be for the STARFIRE II study that should be started next year. Parallel to that, the Office of Fusion Energy wants to go back and continue work on the MARS Tandem Mirror power reactor design activity. This year, they are looking at an Engineering Test Reactor for a tandem mirror, and will continue the INTOR workshop. KMS Fusion, Inc. discusses compelling features of inertial fusion which can lead to reactors that have promise for low cost, low radioactivity, low afterheat, and high maintainability and reliability. The Lawrence Livermore National Laboratory discusses the need for a project which is smaller, higher in power density and something that can be modularly built. PSE and G discusses a number of studies which have been conducted to examine the charateristics of research programs which led to the eventual commercialization of a new technology, and that the utilities may become direct participants in fusion research and development. M.I.T. comments on the role that fusionmore » can play, in that one fusion plant could produce fissile fuel for a large number of fission plants, resulting in a relatively low cost for fissile fuel.« less

A comparison of the direct capital costs of recent light-water-reactor fission plants and recent magnetic fusion designs indicates that cost reductions by innovation in both the fusion reactor and the fusion balance of the plant (nonreactor part) will be required for fusion direct costs to become competitive. Both future fusion and future fission designs would benefit from modularity, standardization, and inherent safety (passive afterheat removal without damage in loss-of-coolant and loss-of-flow accidents) to improve their utility and public acceptance and to reduce nonreactor balance-of-plant costs, indirect costs, and time-related costs. If this can be achieved, inherently safe fusion and fission designs should have comparable reactor enclosure power densities, of the order of 1 MWth/m3. With a reduction by a factor of about 2 in the fusion magnet and heating-system cost per watt, such fusion and fission designs should also have comparable direct and total capital costs per watt, provided the fusion plants can be modularly built in somewhat larger units (300 to 600 MWe) than some of the modular fission units (100 MWe). With the eventual fusion capital cost per watt thus attaining parity with fission some decades later than the development and deployment of second-generation fission plants, the economic incentive for switching growth to new fusion electric plants would derive from lower fuel-cycle costs (fuel startup, operating, processing, and waste-disposal costs). At the same time, there would also be strong economic incentives to build hybrid fusion plants to supply fissile fuel for the established fission plants

The Senior Committee on Environmental, Safety, and Economic Aspects of Magnetic Fusion Energy (ESECOM) summarizes its recent assessment of magnetic fusion energy's (MFE's) prospects for providing energy with economic, environmental, and safety characteristics that would be attractive compared with other energy sources (mainly fission) available in the time frame of the year 2015 and beyond. Accordingly, ESECOM has given particular attention to the interaction of environmental, safety, and economic characteristics of a variety of magnetic fusion reactors, and compared those fusion cases with a variety of fission cases. Eight fusion cases, two fusion-fission hybrid cases, and four fission cases are examined, using consistent economic and safety models, to permit exploration of the environmental, safety, and economic potential of fusion concepts using a wide range of possible materials choices, power densities, power conversion schemes, and fuel cycles.

The U.S. Dept. of Energy discusses concept improvement in the tokamak and concept improvement in the mirror. Controlled Thermonuclear Research comments on what constitutes an attractive fusion reactor, and provides a table of achieved parameters of RFP, FRC and the spheromak experiments. GA Technologies Inc. remarks on the direction which industry must take in the fusion program. The Lawrence Livermore National Laboratory concentrates on commercial reactor studies. Spectra Technology focuses on problems dealing with fusion proponents making a convincing and clear economic argument for fusion based on a mils per kilowat basis, and the large costs of flagship experiments. The Oak Ridge National Laboratory remarks on the need for an economic energy source for fusion. A table of cost of electricity contours is shown.

A multiple mirror experiment confirms the predictions of the theory that the axial confinement time exceeds that of a single mirror of the same length, and that the confinement scales as L2, where L is the system length. The experiment indicates that the improved confinement occurs in an intermediate mean free path regime in which the mean free path for scattering out of a loss cone is of the order of a cell length. The absolute value of the axial confinement is smaller than the optimum confinement predicted from the theory by a factor between two and three, which is accounted for by the deviation of experimental parameters from optimum conditions. The scaling of the confinement time with mirror ratio is also investigated. A reactor calculation using the multiple mirror confinement time gives QE = 2 for a 400‐m system with 3000‐MW(e) output.