Burning Plasma Experiment Research Articles

Next-step spherical torus experiment and spherical torus strategy in the course of development of fusion energy

A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading to more attractive Demo and power plant scale reactors. To provide the physical basis for the CTF an intermediate step needs to be taken, which we refer to as the ‘next-step spherical torus’ (NSST) device and which we examine in some detail herein. NSST is a ‘performance extension’ stage ST with a plasma current of 5–10 MA, R = 1.5 m, BT ⩽ 2.6 T and the possibility of varying physical parameters. The mission of NSST is to (1) provide a sufficient physical basis for the design of a CTF; (2) explore advanced operating scenarios with high bootstrap current fraction and high performance which can be utilized by CTF, Demo, and power plants; and (3) contribute to the general science of high β toroidal plasmas. The NSST is designed to utilize a TFTR-like site to minimize the cost and time required for design and construction.

Key Issues in Diagnostics for Burning Plasma Experiments

The development of diagnostics for a next step burning plasma device (BPX) will be a major challenge. A starting point is the development and justification of the requirements for the measurement of quantities for plasma control, machine protection, and physics evaluation. The design of diagnostic systems and components, to perform these measurements in a BPX, is complicated by the fact that they need to operate reliably for a long time in a very hostile environment. A range of radiation-induced effects may deteriorate the characteristics of electrical and optical components, and R&D programs are focused on finding suitable materials. This paper first addresses the measurement requirements and then focuses on a number of important R&D issues.

Predicting the behaviour of magnetic reconnection processes in fusion burning plasma experiments

Critical stability issues involving magnetic reconnection that are likely to influence the successful operation of burning plasma experiments are addressed. In particular, we discuss: (1) the stabilization of drift-tearing modes in weakly collisional regimes, with particular reference to nonlinear stability, including neo-classical effects; (2) the present understanding of sawtooth oscillations, in particular the interpretation of peculiar sawtooth features in FTU experiments with pellet injection; (3) reconnection near the X-points of magnetic separatrices.

Performance of ITER as a burning plasma experiment

Recent performance analysis has improved confidence in achieving Q (= fusion power/auxiliary heating power)⩾ 10 in inductive operation in ITER. Performance analysis based on empirical scalings shows the feasibility of achieving Q ⩾ 10 in inductive operation, particularly with improved modelling of helium exhaust. Analysis has also indicated the possibility that ITER can potentially demonstrate Q ∼ 50, enabling studies of self-heated plasmas. Theory-based core modelling indicates the need for a high pedestal temperature (3.2–5.3 keV) to achieve Q ⩾ 10, which is in the range of projections with presently available pedestal scalings. Pellet injection from the high-field side would be useful in enhancing Q and reducing edge localized mode (ELM) heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirements on confinement improvement and beta (HH98(y,2) ⩾ 1.3 and βN ⩾ 2.6). Stabilization of the resistive wall modes (RWMs), required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structures taken into account. Recent analysis shows a potential of high power steady state operation with a fusion power of 0.7 GW at Q ∼ 8. Achievement of the required βN ∼ 3.6 by RWM stabilization is a possibility. Further analysis is also needed on reduction of the divertor target heat load.

Burning plasma projections using drift-wave transport models and scalings for the H-mode pedestal

The GLF23 and multi-mode core transport models are used along with models for the H-mode pedestal to predict the fusion performance for the International Thermonuclear Experimental Reactor, Fusion Ignition Research Experiment, and IGNITOR tokamak designs. Simulations using combinations of core and pedestal models have also been compared with experimental data for H-mode profiles in DIII-D, JET, and Alcator C-Mod. Power-independent (ballooning mode limit) and power-dependent pedestal scalings lead to very different predictions when used with the core models. Although the two drift-wave transport models reproduce the core profiles in a wide variety of tokamak discharges, they differ in their projections to burning plasma experiments for the same pedestal parameters. Differences in the core transport models in their response to the ion temperature gradient (i.e. their stiffness) and impact of the power dependence of the H-mode pedestal on fusion performance predictions are discussed.

Resistive Wall Mode Feedback Stabilization in Burning Plasma Experiments

A simple model is used to analyze the systems for feedback stabilization of the resistive wall mode (RWM) in proposed burning plasma experiments. In ITER, the presence of several conducting structures close to the control coils, but far from the plasma, leads to a slow feedback response time compared to the time scale of the RWM growth. In FIRE, the copper shell passive stabilizer sets a relatively long time scale for the RWM growth; therefore, the effects of higher resistivity structures close to the coils and far from the plasma are nearly negligible. RWM feedback control should be able to raise the stable βN up to near the ideal-wall limit in FIRE with moderate requirements on the feedback electronics bandwidth.

Burning Plasma Confinement Projections and Renormalization of the GLF23 Drift-Wave Transport Model

Fusion power predictions are presented using the GLF23 drift-wave transport model for several next-step tokamak designs including ITER, FIRE, and IGNITOR. The GLF23 model has been renormalized using recent gyrokinetic simulations and a database of nearly 50 H-mode discharges from three different tokamaks. The renormalization reduces the ion temperature gradient/trapped electron mode (ITG/TEM)-driven transport by a factor of 3.7 while electron temperature gradient (ETG) mode transport is increased by a factor of 4.8 with respect to the original model. Using the renormed model, the fusion power performance is uniformly assessed, and the pedestal requirements are summarized for each device. The renormed model is still quite stiff and yields somewhat more optimistic predictions for next-step burning plasma experiments. The consequences of stiff transport in the plasma core are discussed. A fusion fit formula is derived whereby the GLF23 results follow a universal stiff model curve for the normalized fusion power versus pedestal temperature.

Integrated predictive modelling simulations of burning plasma experiment designs

Models for the height of the pedestal at the edge of H-mode plasmas (Onjun T et al 2002 Phys. Plasmas 9 5018) are used together with the Multi-Mode core transport model (Bateman G et al 1998 Phys. Plasmas 5 1793) in the BALDUR integrated predictive modelling code to predict the performance of the ITER (Aymar A et al 2002 Plasma Phys. Control. Fusion 44 519), FIRE (Meade D M et al 2001 Fusion Technol. 39 336), and IGNITOR (Coppi B et al 2001 Nucl. Fusion 41 1253) fusion reactor designs. The simulation protocol used in this paper is tested by comparing predicted temperature and density profiles against experimental data from 33 H-mode discharges in the JET (Rebut P H et al 1985 Nucl. Fusion 25 1011) and DIII-D (Luxon J L et al 1985 Fusion Technol. 8 441) tokamaks. The sensitivities of the predictions are evaluated for the burning plasma experimental designs by using variations of the pedestal temperature model that are one standard deviation above and below the standard model. Simulations of the fusion reactor designs are carried out for scans in which the plasma density and auxiliary heating power are varied.

Safety in the Design of Three Burning Plasma Experiments

The 2002 Snowmass Fusion Energy Sciences Summer Study required a uniform assessment of the safety design goals for three candidate burning plasma experiments: the Fusion Ignition Research Experiment (FIRE), the IGNITOR compact tokamak, and the International Thermonuclear Experimental Reactor (ITER). The main assessment criterion was an objective judgment of each design’s ability to obtain a generalized regulatory approval. A brief overview of environmental impact, safety, and health results from the uniform assessment of safety are given in this paper. As safety documentation was reviewed for each design, several issues became apparent. This paper also documents these specific issues. Each of these three designs could obtain a general regulatory approval based on their safety design practices.

New Directions in Fusion Plasma Diagnostic Development

ABSTRACTDiagnostic development is an important and vital part of the fusion program. There is close coupling between the physics needs for measurements and the diagnostic development that receives funding. This paper gives brief technical descriptions of ongoing developments in a few areas currently receiving attention. These areas include Te diagnostics, j(R) diagnostics applicable to new regimes, and diagnostics to probe edge and core turbulence. The underlying physics motivation and enabling technologies involved in these developments are also mentioned. A brief status of diagnostic work in support of a burning plasma experiment (BPX) is also described, since this effort may become more significant worldwide. Several diagnostic R&D topics, identified as crucial for a BPX, are discussed.

Creep Effects in the Toroidal Field Coils of Fire and Other Burning Plasma Tokamaks

ABSTRACTAll three burning plasma experiments discussed at Snowmass during the summer of 2002, use preloaded structures to resist some component of the operating loads. For the resistive pulsed reactors, it is the preloads which introduce the most noticeable creep responses because these loads are applied for much longer than the operating loads. If the preloads are maintained during shut-down and maintenance periods, then the structure experiences the preload stresses at room temperature. OFHC copper has significant creep behavior, predominantly at high stress and high temp, but copper experiences finite creep even at cryogenic temperatures. The Beryllium copper used in the FIRE inner leg has better creep properties than OFHC copper.The purpose of these analyses is to characterize the influence of creep on the magnets of the Fusion Ignition Research Reactor (FIRE) and compare it with the creep response of the other proposed burning plasma experiments. The concern is that the desirable features provided by coil preloads will be lost over the lives of the experiments. Structural finite element models of FIRE and IGNITOR are used with creep equations derived from NIST[6] data to explore the structural sensitivity of the machines to creep. For both FIRE and IGNITOR, copper coil material, creep has been found to have a minimal effect on magnet performance. IGNITOR’s generally lower stresses (with respect to FIRE’s BeCu TF stresses) and the use of active as well as passive preload systems helps reduce creep to acceptable levels. FIRE’s structure is more sensitive to creep due to the free standing wedged TF coil, but the BeCu used in FIRE’s inner TF legs has a much lower creep behavior than ETP or OFHC copper. This reduces creep to acceptable levels. For FIRE, however, there is some creep in the horizontal legs which relaxes some of the support of the inner leg. Recommendations are presented to support the OFHC copper horizontal legs more effectively. More work is needed to address the multiple load-unload cycling effects on creep.

Stationary high-performance discharges in the DIII-D tokamak

Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak under stationary conditions at relatively low plasma current (q95>4). A figure of merit for fusion gain (βNH89/q952) has been maintained at values corresponding to Q = 10 operation in a burning plasma for >6 s or 36τE and 2τR. The key element is the relaxation of the current profile to a stationary state with qmin>1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall β limit. Feedback control of the energy content and particle inventory allow reproducible, stationary operation. The particle inventory is controlled by gas fuelling and active pumping; the wall plays only a small role in the particle balance. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. In addition, the pulse length capability is greatly increased, which is essential for a technology testing phase of a burning plasma experiment where fluence (duty cycle) is important.

Magnetic diagnostics for ITER/BPX plasmas (invited)

The measurement requirements and expected performance of the International Thermonuclear Experimental Reactor (ITER) magnetics system are summarized. The constraints of a burning plasma experiment (BPX) on tokamak magnetic diagnostic system designs are listed and illustrated using the parameters and design of sensors being developed for the ITER device. Design principles to meet the constraints are discussed, with an emphasis on radiation effects, reliability, and redundancy for sensors and subsystems. The remaining research and development for the ITER system is outlined.

Diagnostic exploitation of complex heavy elements in tokamak plasmas

In future burning plasma experiments, a promising method for the diagnosis of first wall erosion is to detect the onset of emission from nonintrinsic materials embedded in the plasma facing components as they are revealed and enter the confined plasma. Likely high Z species were introduced into the Joint European Torus plasmas by laser ablation. Results from targets of single species (W and Hf) and dual species (W/Hf composite) are presented. The composite target experiment has shown that it is possible to distinguish reliably between emission from near neighbor heavy elements, even with instrument resolutions of Δλ/λ∼0.001. Forward prediction to power plant conditions necessitates a model of the complex spectra and a system for generating the required atomic data and producing spectral feature emissivity coefficients is described. An important consideration is to enable exploration of different instrumentation so the spectral resolution of the compound features are tuned to the diagnostic capability.

System Analysis of a Compact Next Step Burning Plasma Experiment

A new burning plasma systems code has been developed for analysis of a next step compact burning plasma experiment with copper-alloy magnet technology. Two classes of configurations are considered: type A, with the toroidal field (TF) coils and ohmic heating (OH) coils unlinked, and type B, with the TF and OH coils linked. Curves of the minimizing major radius as a function of aspect ratio R(A) are obtained for each configuration type for typical parameters. These curves represent, to first order, cost-minimizing curves. The type B curves always lie below the type A curves for the same physics parameters, indicating that they lead to a more compact design. However, the fact that the type A OH and TF magnets are not linked presents fewer engineering challenges and should lead to a more reliable design. Both the type A and type B curves have a minimum in major radius R at a minimizing aspect ratio A typically above 2.8 and at high values of magnetic field B above 10 T. The minimizing A occurs at larger values for longer pulse and higher performance devices. The larger A and higher B design points also have the feature that the ratio of the discharge time to the current redistribution time is largest so that steady-state operation can be more realistically prototyped. A sensitivity study is presented for the baseline type A configuration showing the dependence of the results on the parameters held fixed for the minimization study.

JET beamline with integrated x-ray, VUV, and visible spectrometers, for burning plasma experiments

We report several developments to the JET Tokamak survey spectroscopy beamline, which now supports a suite of x-ray, VUV, and visible instruments, located in a radiation-shielded bunker outside the torus hall. The instruments have vacuum systems that are integrated into the tritium circuit, and are available for D–T experiments. The x-ray and VUV diagnostics have real-time data reduction. The proposed plasma impurity monitor for ITER is based on these developments.

ELM energy and particle losses and their extrapolation to burning plasma experiments

Analysis of Type I ELMs from present experiments shows that ELM energy losses decrease with increasing pedestal plasma collisionality ( ν ∗ ped ) and/or increasing τ Front ∥, where ( τ ∥ Front=2π Rq 95/ c s,ped ) is the typical ion transport time from the pedestal to the divertor target. ν ∗ ped and τ Front ∥ are not the only parameters that affect the ELMs, also the edge magnetic shear influences the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. ‘Minimum’ Type I ELMs, with energy losses acceptable for ITER, where there is no change in the plasma temperature profile during the ELM, are observed for some conditions in JET and DIII-D. The duration of the divertor ELM power pulse is well correlated with τ Front ∥ and not with the duration of the ELM-associated MHD activity. Similarly, the time scale of ELM particle fluxes is also determined by τ Front ∥. The extrapolation of present experimental results to ITER is summarised.

Feasibility Study of a Compact Ignition Tokamak Based upon GyroBohm Scaling Physics

The design of a reduced size (R = 4.45 m, BT = 5.04 T) ignition tokamak (Q = ∞) with superconducting coils using a standard ELMing H-mode plasma appears to be feasible. This effective size (BT2/3R5/6) is smaller than current proposals for Q = 10 burning (D-T) plasma experiments. The good confinement required for ignition with this small effective size is obtained by operating along a gyroBohm scaling path starting from the existing tokamak database at high beta (β = 4.1%) so that the loss power from core transport exceeds the H-mode threshold power. Using a design that can achieve a high normalized current (Ip/aBT = 1.63) also helps to decrease the size of the machine. The design of this relatively compact ignition tokamak satisfies reasonable engineering constraints on the superconducting toroidal field coils and central solenoid, and allows for a sufficiently long burn time for the plasma current to relax to its final state.

Physics basis and simulation of burning plasma physics for the fusion ignition research experiment (FIRE)

The FIRE design for a burning plasma experiment is described in terms of its physics basis and engineering features. Systems analysis indicates that the device has a wide operating space to accomplish its mission, both for the ELMy H-mode reference and the high bootstrap current/high-β advanced tokamak regimes. Simulations with 1.5D transport codes reported here both confirm and constrain the systems projections. Experimental and theoretical results are used to establish the basis for successful burning plasma experiments in FIRE.

Role of Fusion Energy in a Sustainable Global Energy Strategy

Fusion energy is one of only a few truly long-term energy options. Since its inception in the 1950s, the vision of the fusion energy research program has been to develop a viable means of harnessing the virtually unlimited energy stored in the nuclei of light atoms--the primary fuel deuterium is present as one part in 6,500 of all hydrogen. This vision grew out of the recognition that the immense power radiated by the sun is fueled by nuclear fusion in its hot core. Such high temperatures are a prerequisite for driving significant fusion reactions. The fascinating fourth state of matter at high temperatures is known as plasma. It is only in this fourth state of matter that the nuclei of two light atoms can fuse, releasing the excess energy that was needed to separately bind each of the original two nuclei. Because the nuclei of atoms carry a net positive electric charge, they repel each other. Hydrogenic nuclei, such as deuterium and tritium, must be heated to approximately 100 million degrees Celsius to overcome this electric repulsion and fuse. There have been dramatic recent advances in both the scientific understanding of fusion plasmas and in the generation of fusion powermore » in the laboratory. Today, there is little doubt that fusion energy production is feasible. For this reason, the general thrust of fusion research has focused on configuration improvements leading to an economically competitive product. The risk of conflicts arising from energy shortages and supply cutoffs, as well as the risk of severe environmental impacts from existing methods of energy production, are among the reasons to pursue these opportunities [1]. In this paper we review the tremendous scientific progress in fusion during the last 10 years. We utilize the detailed engineering design activities of burning plasma experiments as well as conceptual fusion power plant studies to describe our visions of attractive fusion power plants. We use these studies to compare technical requirements of an attractive fusion system with present achievements to identify remaining technical challenges for fusion. We discuss scenarios for fusion energy deployment in the energy market.« less