Confinement Improvement Factor Research Articles

Evaluation of NBI absorption and fast ion stored energy to improve thermal energy confinement time calculation in EAST

The absorption of neutral beam power and the fast ion stored energy in EAST plasmas with neutral beam injection (NBI) is analyzed to improve the calculation of thermal energy confinement time. The neutral beam power absorption and fast ion stored energy are systematically calculated using the TRANSP code, through the investigation of global parameters including plasma current, line averaged density and beam energy. Results have shown that scaling laws for the NBI absorption coefficient and fast ion energy rate are obtained through statistical analysis. A comparison of the confinement improvement factor with these new scaling laws against those assuming fixed coefficients is given.

Lower hybrid wave absorption study to improve the calculation of energy confinement time in EAST

The auxiliary heating power absorption of 2.45 GHz and 4.6 GHz lower hybrid (LH) waves in EAST is analysed in order to improve the determination of the energy confinement time. The LH power absorption coefficient is obtained by using the time derivative of the total stored energy including the plasma kinetic energy and the poloidal magnetic field energy. Results on EAST have shown that, for both frequencies, the fraction of LH power absorbed in the plasma bulk decreases with the increase in the plasma density, and with the decrease in the toroidal magnetic field. Scaling laws for the density dependence of the LH absorption coefficient at different fields and frequencies have been statistically obtained. Comparison of the confinement improvement factor H89 obtained with these new scaling laws against those assuming an absorption coefficient equal to the launched k//wave spectrum directivity is made.

The GOL-NB program: further steps in multiple-mirror confinement research

Physical and technical details of the GOL-NB project are presented. GOL-NB is a medium-scale multiple-mirror trap that is under development in the Budker Institute, Novosibirsk, Russia. This device will be created in several years as a deep conversion of the existing GOL-3 facility. It will consist of a central trap with two 0.75 MW neutral beams, two multiple-mirror solenoids, two expander tanks and a plasma gun that creates the start plasma. The central trap with the neutral beam injection-heated plasma is a compact gas-dynamic system. The multiple-mirror sections should decrease the power and particle losses along the magnetic field. The confinement improvement factor depends on plasma parameters and on the magnetic configuration in the multiple mirrors. The main physical task of GOL-NB is direct demonstration of the performance of multiple-mirror sections that will change equilibrium plasma parameters in the central trap. In this paper we discuss results of the scenario modeling and progress in the hardware.

A control method of divertor plasma start-up assisted by tritium-ratio control for Demo-CREST

In this paper, core plasma operation flexibility by the tritium-ratio(T-ratio) control is proposed in order to control the fusion power keeping a high plasma density for the divetor heat-handling during the commissioning phase. The operational space of the core plasma for the Demo-CREST concept is expanded to the high density region by reducing T-ratio. Especially, the expansion of operation space for the low beta operation is large. For such operation expansion, the critical parameter is the confinement improvement factor HH, and a high HH similar to that of the ITER steady state (ITER-SS) operation is required for Demo-CREST. The resultant condition of the SOL density and the impurity radiation power required for the divertor heat load less than 10 MW/m 2 is analyzed, and those operation conditions for the initial operation point of the fusion power 0 MW during the commissioning phase are found to be similar to those of the ITER-SS operation.

Response of fusion gain to density in burning plasma simulation on JT-60U

A burning plasma simulation scheme has been developed with consideration for temperature dependence of the DT fusion reaction rate in JT-60U. The heating power for the simulation of alpha particle heating was calculated using real-time measurements of density and ion temperature. The response of a simulated fusion gain to the density was investigated in this scheme where a constant heating power was used for the simulation of external heating, in order to understand burn controllability by the fuel density in a fusion reactor. When temperature dependence of the fusion reaction rate was assumed as a square of ion temperature, density dependence of the simulated fusion gain stronger than the square of density was observed. Transport analysis using a 1.5 dimension transport code indicated that the strong density dependence is induced by changes in a confinement improvement factor and changes in a pressure profile.

On the extrapolation to ITER of discharges in present tokamaks

An expression for the extrapolated fusion gain G = Pfusion/5Pheat (Pfusion being the total fusion power and Pheat the total heating power) of ITER in terms of the confinement improvement factor (H) and the normalized beta (βN) is derived in this paper. It is shown that an increase in normalized beta can be expected to have a negative or neutral influence on G depending on the chosen confinement scaling law. Figures of merit like should be used with care, since large values of this quantity do not guarantee high values of G, and might not be attainable with the heating power installed on ITER.

Study of Impurity Injection Scenario in Burning Plasmas

Accumulation of impurity injected for reduction of heat load to the divertor plates was of great concern with a peaked density profile. Applicability of impurity injection to a burning plasma with a peaked density profile was investigated for various impurity accumulation levels using the A-SSTR2 design parameters. Impurity transport analysis indicated that the argon density profile twice as peaked as the electron density profile can yield acceptable radiation profile even with a peaked density profile. The required confinement improvement factor over the IPB98(y,2) scaling slightly increased from 1.4 with the flat density profile to 1.5 with the peaked electron density profile at ne(r/a = 0)/ne(r/a = 0.7) ~ 3. When the argon density profile was determined by neoclassical transport, the radiation loss in the core plasma intensively increased with the peaked density profile, which requires higher confinement enhancement factor of 1.9 at ne(r/a = 0)/ne(r/a = 0.7) ~ 3.

Impact of the edge pedestal characteristics on the integrated performance in advanced tokamak operation modes in JT-60U

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

Impurity-seeded ELMy H-modes in JET, with high density and reduced heat load

Experiments performed at JET during the past two years show that, in high triangularity H-mode plasmas with Ip = 2.5 MA, ne/nGr ≈ 1.0, it is possible to radiate separately up to ≈40% of the total injected power on closed flux surfaces in the pedestal region (argon seeding) and up to ≈50% of the injected power in the divertor region (nitrogen seeding), while maintaining the confinement improvement factor at the value required for ITER, H98(y, 2) ≈ 1.0. The total radiated power fraction achieved in both cases (65–70%) is close to that required for ITER. However, Type I ELMs observed with impurity seeding have the same characteristics as that observed in reference pulses without seeding: decreasing plasma energy loss per ELM with increasing pedestal collisionality. One has to reach the Type III ELM regime to decrease the transient heat load to the divertor to acceptable values for ITER, although at the expense of confinement. The feasibility of an integrated scenario with Type-III ELMs, and q95 = 2.6 to compensate for the low H factor, has been demonstrated on JET. This scenario would meet ITER requirements at 17 MA provided that the IPB98 scaling for energy content is accurate enough, and provided that a lower dilution is obtained when operating at higher absolute electron density.

Steady Burning Criteria for D-3He Spherical Tokamak Reactor with Internal Transport Barrier

Steady burning criteria for a D-3He fusion reactor have been investigated by using the 0-D power and particle balance equations and the IPB98(y,2) scaling law. A spherical tokamak (ST) configuration with the internal transport barrier (ITB) and the bootstrap current fraction of 100 % are assumed. As main results, the hot ion mode (Ti/Te > 1) and the high central beta are essential to realize a D-3He fusion reactor, and the confinement improvement does not reduce the requirement of the hot ion mode. The reactor size, the magnetic field, confinement improvement factor and the heat flux may be attainable if the hot ion mode is realized.

Forthcoming Break-Even Conditions of Tokamak Plasma Performance for Fusion Energy Development

The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value ΒN), confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fnGW), the electric break-even condition requires the simultaneous achievement of 1.2 < ΒN < 2.7, 0.8 < HH, and 0.3 < fnGW < 1.1 under the conditions of a maximum magnetic field on the TF coil Btmax = 16 T, thermal efficiency ηe = 30 %, and current drive power PNBI < 200 MW. It should be noted that the relatively moderate conditions of ΒN ˜ 1.8, HH ˜ 1.0, and fnGW ˜ 0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric break-even condition. The second condition is the economic break-even condition, which is required for fusion energy to be selected as an alternative energy source in the energy market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of Btmax = 16 T, ηe = 40 %, plant availability 60 %, and a radial build with/without CS coil, the economic break-even condition requires ΒN ˜ 5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with ΒN ˜ 3.0 in the ITER project leads to the upper region of the break-even price in the present world energy scenario, which implies that it is necessary to improve the plasma performance beyond that of the ITER advanced plasma operation.

6. Feasibility of Advanced Fuel Fusion Reactor

Feasibility of D-3He fusion reactor has been investigated by using 0-dimensional power and particle balance equations and the IPB98(y,2) scaling. A self burning spherical tokamak (ST) D-3He reactor requires the high Ti mode and low impurity contamination. The beta value, the confinement improvement factor (HH98y2) and the superconducting magnet should be improved from the present level.

Plasma Performance Required for a Tokamak Reactor to Generate Net Electric Power.

Using a system analysis code, plasma parameter ranges required for a Tokamak reactor to generate net electric power (Penet) were investigated. The investigation revealed that in order for a Tokamak to produce a net electric power of 0 MW under foreseeable engineering conditions, it is required to achieve simultaneously the following ranges of parameters, i.e., normalized beta value 1.2 ≤ βN ≤ 2.7, confinement improvement factor for H-mode scaling 0.8 ≤ HH, and the ratio of Greenwald density limit 0.4 ≤ fnGW ≤ 1.1, for a major radius R ≤ 8.5 m. It also revealed that a reactor to produce a net electric power of 1,000 MW requires the simultaneous achievement of βN ≥ 3.0, HH ≥ 0.9 and fnGW ≥ 0.9.

Parameter Optimization of the Inductively Operated Day-Long Tokamak Reactor

The plasma and machine parameters of a pulsed tokamak reactor with a day-long operation period have been studied, where engineering constraints such as maximum toroidal field strength are preserved at International Thermonuclear Experimental Reactor (ITER) levels so as to realize a fusion reactor with only a short-range extension of currently available technology. To provide the magnetic flux necessary to sustain a plasma current inductively for 1 day or longer, plasma with a major radius of R > 9.5 m are necessary, and a plasma with an aspect ratio as high as A>5 should be employed. Typical parameters are as follows: major radius R = 10 m, minor radius a = 1.85 m, plasma elongation [kappa] = 1.8, plasma current I[sub p] = 12.2 MA, toroidal field on axis B[sub t] = 7.56 T, and safety factor at the plasma surface q[psi] = 3. A plasma volume V [approximately] 1200 m[sup 3] is comparable with that of ITER, even though the major radius of a day-long operation reactor is relatively large. A very small amount of heating power ([approximately]15 MW) with a heating time of only a few tens of seconds is sufficient to achieve the ignition condition. This is wellmore » within the capacity of auxiliary heating systems currently used in large tokamak devices. A confinement improvement factor (from L mode) of [line integral][sub L] > 1.7 is required to design a reactor with a reasonable machine size and a day-long pulse duration. The operation temperature is chosen to be [r angle]T[l angle] = 20 keV with a toroidal beta [beta][sub t] = 2.6% (Troyon factor g = 3), which gives a fusion power P[sub fus] = 2.5 GW even for an alpha-particle dilution n[sub [alpha]]/n[sub e] of 10%. The bootstrap current fraction is 50% or more of the total current, and current profile needed for the beta limit could be achieved with a combination of ohmic current in the plasma center region and bootstrap current in the outer region. 23 refs., 8 figs., 2 tabs.« less