Fusion Energy Applications Research Articles

Simulations for experimental study of warm dense matter and inertial fusion energy applications on NDCX-II

The Neutralized Drift Compression Experiment II (NDCX II) is an induction accelerator planned for initial commissioning in 2012. The final design calls for a ∼3 MeV, ∼30 A Li+ ion beam, delivered in a bunch with characteristic pulse duration of 1 ns, and transverse dimension of order 1 mm. The purpose of NDCX II is to carry out experimental studies of material in the warm dense matter regime, and ion beam/hydrodynamic coupling experiments relevant to heavy ion based inertial fusion energy. In preparation for this new machine, we have carried out hydrodynamic simulations of ion-beam-heated, metallic solid targets, connecting quantities related to observables, such as brightness temperature and expansion velocity at the critical frequency, with the simulated fluid density, temperature, and velocity. We examine how these quantities depend on two commonly used equations of state.

Creep and fatigue issues for structural materials in demonstration fusion energy systems

The structural materials proposed for use in future fusion energy systems must perform reliably in an environment consisting of intense neutron irradiation, high temperatures, and cyclic stress. Therefore, thermal creep and creep-fatigue (in addition to irradiation creep) are anticipated to be important issues for the engineering design of structural materials for fusion reactors. The key materials systems under consideration for structures of fusion reactors include 8–9%Cr ferritic/martensitic steels, oxide dispersion strengthened ferritic steels, vanadium alloys and SiC fiber-reinforced SiC matrix ceramic composites. The current elevated temperature creep-fatigue design rules based on the American Society of Mechanical Engineers (ASME) code are discussed, along with a brief review of creep-fatigue interaction mechanisms. Refinements to current international design codes to include radiation-induced phenomena such as reduction in uniform elongation have been performed in association with the engineering design of the ITER fusion energy device currently under construction in France. Several other creep-fatigue issues of potential importance for fusion energy applications are discussed.

Mission and Overview of a Fusion Development Facility

The objective of the Fusion Development Facility (FDF) under consideration is to fill the gaps between ITER and current experiments and a fusion demonstration power plant (DEMO). FDF should carry forward advanced tokamak (AT) physics and enable development of fusion's energy applications. Near-term AT physics and nonsuperconducting magnet technology will be used to achieve steady state with burn, producing 100-250 MW fusion power with modest energy gain (Q<5) in a modest sized device (between DIII-D and Joint European Torus). FDF will further develop all elements of AT physics for an advanced performance DEMO. With neutron flux at the outboard midplane of 1-2 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> , continuous operation for periods of up to two weeks, and a goal of a duty factor of 0.3 per year, FDF can produce fluences of 3-6 MW·year/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> in ten years of operation, for fusion nuclear component research and development. The development of blankets suitable for tritium, electricity, and hydrogen production will be done first in port modules. Then, the most promising candidates will be deployed as full blankets in FDF. Two to three full blankets and about a dozen port-blanket types could be tested. The goal of FDF is to demonstrate closure of the fusion fuel cycle, producing its own tritium. FDF, ITER, International Fusion Materials Irradiation Facility, and other AT devices will provide the basis for a fusion DEMO power plant of the ARIES-AT type.

Fusion Power Measurement Using a Combined Neutron Spectrometer-Camera System at JET

Fusion power production is the ultimate goal of fusion research, and its determination is crucial in any fusion energy application. In this paper the principles of collimated neutron flux measurements for fusion plasma power determination are described. In this method, a high-resolution neutron spectrometer provides an absolutely calibrated neutron flux, and a neutron profile monitor (“camera”) gives information on the neutron emission profile of the plasma. The total neutron flux seen by the spectrometer is discussed in terms of direct and scattered flux, and a model is set up to evaluate the magnitude of these different components. Particular care is taken to estimate the uncertainties involved, both in the model and the measurements. The method is put to practical use at JET, where a magnetic proton recoil spectrometer and a neutron profile monitor are available. Results from JET’s trace tritium experimental campaign in 2003 are presented and show that the systematic uncertainties in fusion power measurements are reduced in comparison to what has been presented for foil activation systems. A systematic error of 6% is reported here. For ITER these results imply that the fusion power can be redundantly measured and with better accuracies than for traditional methods.

Operating Temperature Window for SiC Ceramics and Composites for Fusion Energy Applications

Limitations in operating conditions, primarily the steady-state operating temperature, of silicon carbide-based ceramics and composites for applications to structural and functional components in fusion blanket systems were critically examined based on the latest experimental results. Irradiation-induced high temperature swelling and irradiation creep were identified to be the likely factors limiting the upper temperature bound for structural applications, whereas irradiation-induced thermal conductivity degradation was identified to be the primary factor to limit the lower temperature bound when substantial heat flux is anticipated. For the application to flow channel inserts in liquid metal blankets, insulating properties will likely limit the upper temperature bound, whereas the lower temperature bound may be limited by swelling-induced secondary stress. Additionally, key scientific issues which need to be addressed for the better definition of design limitations were identified.

Studies on possible alternative schemes based on two-laser driver for inertial fusion energy applications

AbstractUnconventional, promising approaches to inertial fusion energy (IFE) proposed and studied in the framework of IFE keep-in-touch activity at Frascati EURATOM-ENEA Association will be reconsidered for the present more advanced technological context associated to a planned large installations scenario. Then the possibility to generate laser-produced fast heavy-ion sources (e.g., Bi ions, tens GeV) will be explored taking as reference results obtained in the FIGEX experiment scaled toward thermonuclear regimes. A possible dimensioning of such sources will be also given as well as the associated laser requirements.

Operating Temperature Window for SiC Ceramics and Composites for Fusion Energy Applications

Operating Temperature Window for SiC Ceramics and Composites for Fusion Energy Applications

Status and strategy of fusion materials development in China

The liquid metal and solid ceramic pebble breeder blankets have become the most promising blankets for ITER-TBMs or DEMO reactors in China and the world due to their potential advantages. In recent years the corresponding research work on fusion reactor materials mainly focuses on structural materials, plasma facing materials and the functional materials for the blanket such as breeder, coating and flow channel insert etc. for the successful application of fusion energy in the near future. The R&D on those materials in the two kinds of blankets is being carried out widely in China, including fabrication and manufacturing techniques, physical/mechanical properties assessment before and after irradiation, joining techniques for structural materials, compatibility evaluation, and the development and verification of the criteria for fusion material designs. The progress on main R&D activities of fusion reactor materials in China is introduced and prospected in the paper.

The Mercury Project: A High Average Power, Gas-Cooled Laser for Inertial Fusion Energy Development

Hundred-joule, kilowatt-class lasers based on diode-pumped solid-state technologies, are being developed worldwide for laser-plasma interactions and as prototypes for fusion energy drivers. The goal of the Mercury Laser Project is to develop key technologies within an architectural framework that demonstrates basic building blocks for scaling to larger multi-kilojoule systems for inertial fusion energy (IFE) applications. Mercury has requirements that include: scalability to IFE beamlines, 10 Hz repetition rate, high efficiency, and 109 shot reliability. The Mercury laser has operated continuously for several hours at 55 J and 10 Hz with fourteen 4 × 6 cm2 ytterbium doped strontium fluoroapatite amplifier slabs pumped by eight 100 kW diode arrays. A portion of the output 1047 nm was converted to 523 nm at 160 W average power with 73 % conversion efficiency using yttrium calcium oxy-borate (YCOB).

Initial Conditions and Near-Field Dynamics in Turbulent Liquid Sheets

High-speed liquid “curtains” have been proposed to protect solid structures in fusion energy applications. Minimizing free-surface waves and spray formation in such flows is important for effective protection in this application. In this work, free-surface waves and turbulent breakup were studied experimentally in jets of water issuing from a rectangular nozzle into ambient air at a Reynolds number of 1.2 × 105. Laser-Doppler anemometry was used to characterize the streamwise and transverse velocity components in the nozzle for two different flow calming section designs. Planar laser-induced fluorescence was used to measure the free-stream position in the near-field of the sheet. The results suggest that transverse velocity fluctuations in the nozzle are the primary factor in determining the amplitudes of free-surface waves. Removing a small amount of low-speed fluid immediately downstream of the nozzle exit (“boundary-layer cutting”) is shown to both significantly reduce free-surface waves and the amount of spray due to turbulent breakup. Overall, boundary-layer cutting appears to have the greatest benefit when used on a “well-conditioned” turbulent liquid sheet.

Design status and development strategy of China liquid lithium–lead blankets and related material technology

A series of fusion reactors (named FDS series) have been designed and assessed in China, with four types of liquid lithium lead blankets including the RAFM steel-structured He-cooled quasi-static LiPb tritium breeder (SLL) blanket, the RAFM steel-structured He–LiPb dual-cooled (DLL) blanket, the RAFM steel-structured refractory material thermally-insulated high temperature LiPb (HTL) hydrogen production blanket and the RAFM steel or optionally the austenitic stainless steel-structured He–LiPb dual-cooled high level waste transmutation (DWT) blanket. To demonstrate and validate the feasibility of the candidate blankets for fusion energy application, the three-phases-strategy of TBM (test blanket module) development, i.e. material R&D and out-of-pile experimental mockup, EAST-TBM and ITER-TBM have been proposed. A brief overview of the four types of LiPb blanket designs and their goals are given. Material technology requirement and development strategy are also presented in this paper.

Current status and critical issues for development of SiC composites for fusion applications

Silicon carbide (SiC)-based ceramic composites have been studied for fusion applications for more than a decade. The potential for these materials have been widely discussed and is now understood to be (1) the ability to operate in temperature regimes much higher than for metallic alloys, (2) an inherent low level of long-lived radioisotopes that reduces the radiological burden of the structure, and (3) perceived tolerance against neutron irradiation up to high temperatures. This paper reviews the recent progress in development, characterization, and irradiation effect studies for SiC composites for fusion energy applications. It also makes the case that SiC composites are progressing from the stage of potential viability and proof-of-principle to one where they are ready for system demonstration, i.e., for flow channel inserts in Pb–Li blankets. Finally, remaining general and specific technical issues for SiC composite development for fusion applications are identified.

Conceptual design of the fusion-driven subcritical system FDS-I

The fusion-driven subcritical system (named FDS-I) was previously proposed as an intermediate step toward the final application of fusion energy. A conceptual design of the FDS-I is presented, which consists of the fusion neutron driver with relatively easy-achieved plasma parameters, and the He-gas/liquid lithium–lead Dual-cooled subcritical Waste Transmutation (DWT) blanket used to transmute long-lived radioactive wastes and to generate energy on the basis of self-sustainable fission and fusion fuel cycle. An overview of the FDS-I is given and the specifications of the design analysis are summarized.

Neutronics safety analysis in severe transients of the dual-cooled waste transmutation blanket for the FDS-I

The conceptual design of the fusion-driven subcritical system FDS-I with the multifunctional subcritical dual-cooled waste transmutation (DWT) blanket proposed as a middle step toward the final application of fusion energy was presented previously. Safety is a key concern for the innovative conceptual system. The potential safety characteristic is expected as one of the advantages of FDS-I for the transmutation and incineration of nuclear waste compared with the critical reactor. With the intent of evaluating the inherent and passive safety features of FDS-I, the reactivity coefficients (e.g. the coolant density/void effect and the Doppler feedback), which are similar to those in critical reactors, and the kinetics quantities (e.g. neutron generation time and β eff), which are influenced by the external neutron source of the source-driven system, are calculated and analyzed. In addition, the specific transient scenarios for FDS-I, which show the effects of the source power perturbation on the safety parameters, is presented.

Analysis on fuel cycle schemes in the dual-cooled waste transmutation blanket for the FDS-I

The fusion-driven subcritical system (FDS-I) which consists of the fusion neutron driver with feasible and easy-achieved plasma parameters and the multifunctional subcritical blanket (DWT) was proposed as an intermediate step toward the final application of fusion energy. In this contribution, fuel configuration and chemical form to be adopted in the FDS-I and possible fuel cycle schemes are clarified and analyzed based on the general survey. The influences of initial fuel loadings on the fuel cycle parameters are presented. Optimal analysis on the in-pile operation length is discussed and some recommendations are given to guide the design of optimized fuel cycle scheme on the basis of comprehensively considering the balance of the spent fuel loadings and the transmutation capacity.

High-density and high-ρR fuel assembly for fast-ignition inertial confinement fusion

Scaling relations to optimize implosion parameters for fast-ignition inertial confinement fusion are derived and used to design high-gain fast-ignition targets. A method to assemble thermonuclear fuel at high densities, high ρR, and with a small-size hot spot is presented. Massive cryogenic shells can be imploded with a low implosion velocity VI on a low adiabat α using the relaxation-pulse technique. While the low VI yields a small hot spot, the low α leads to large peak values of the density and areal density. It is shown that a 750kJ laser can assemble fuel with VI≃1.7×107cm∕s, α≃0.7, ρ≃400g∕cc, ρR≃3g∕cm2, and a hot-spot volume of less than 10% of the compressed core. If fully ignited, this fuel assembly can produce high gains of interest to inertial fusion energy applications.

Liquid Vortex Shielding for Fusion Energy Applications

Swirling liquid vortices can be used in fusion chambers to protect their first walls and critical elements from the harmful conditions resulting from fusion reactions. The beam tube structures in heavy ion fusion (HIF) must be shielded from high energy particles, such as neutrons, x-rays and vaporized coolant, that will cause damage. Here an annular wall jet, or vortex tube, is proposed for shielding and is generated by injecting liquid tangent to the inner surface of the tube both azimuthally and axially. Its effectiveness is closely related to the vortex tube flow properties. 3-D particle image velocimetry (PIV) is being conducted to precisely characterize its turbulent structure. The concept of annular vortex flow can be extended to a larger scale to serve as a liquid blanket for other inertial fusion and even magnetic fusion systems. For this purpose a periodic arrangement of injection and suction holes around the chamber circumference are used, generating the layer. Because it is important to match the index of refraction of the fluid with the tube material for optical measurement like PIV, a low viscosity mineral oil was identified and used that can also be employed to do scaled experiments of molten salts at high temperature.

Views on Neutronics and Activation Issues Facing Liquid-Protected IFE Chambers

Thin and thick liquid walls provide an attractive solution to the challenging material issues facing the heavy-ion applications of the inertial fusion energy (IFE) concept. Given the many advantages of liquid-protected chambers, there are several nuclear-related concerns that are discussed in detail for the thick liquid wall option in particular. These are the ability to protect the steel-based structure from radiation damage and high activation, the feasibility of rewelding the structure, and the pulse-related problems. These issues have a profound impact on the ARIES-IFE thick-liquid protected chamber design.

Compact Toroid and Advanced Fuel - Together to the Moon?!

Magnetic geometries like the field reversed configuration and spheromak are compact toroids, where plasma is enclosed in a linear external magnetic field geometry. To burn advanced fuels requires a high fusion power density compact system and a D-3He field reversed configuration will be considered here. Such innovative confinement concept together with remarkable advantages of compact system and advanced fuels, including direct energy conversion, full lifetime fusion core, fusion space propulsion, nonelectrical application of D-3He fusion energy and space resources including helium-3 fusion fuel is very attractive.

Alternate Applications of Fusion—Production of Radioisotopes

A major effort to find near-term, non-electric applications of fusion energy has shown that the production of radioisotopes is attractive. The use of the D3He fusion reaction to produce Positron Emission Tomography (PET) isotopes is described. An Inertial Electrostatic Confinement (IEC) device is particularly well suited to produce low levels of high-energy (14.7 MeV) protons, which in turn, can produce short-lived PET isotopes. The IEC device at the University of Wisconsin has been modified to investigate the potential of this process to be commercially attractive.