Facteur de gain d'énergie de fusion nucléaire

vignette| L'explosion de la bombe à hydrogène Ivy Mike. La bombe à hydrogène est le seul dispositif actuellement capable d'atteindre un facteur de gain d'énergie de fusion nettement supérieur à 1. En technologie des réacteurs à fusion nucléaire, le facteur de gain d'énergie de fusion, que l'on exprime couramment à l'aide du symbole Q, est le rapport entre la puissance de fusion produite par le réacteur et la puissance auxiliaire fournie pour maintenir le plasma en régime stationnaire. La situation pour laquelle Q = 1, lorsque la puissance libérée par les réactions de fusion est égale à la puissance de chauffage auxiliaire, est appelée seuil de rentabilité (« breakeven » dans la littérature anglo-saxonne). Certains ouvrages parlent parfois de seuil de rentabilité scientifique (« scientific breakeven » dans la littérature anglo-saxonne). Une partie de l'énergie produite par les réactions de fusion peut être capturée au sein du combustible, entraînant un auto-échauffement. Le reste de l'

Free-Boundary Simulations of ITER Advanced Scenarios

Karim Besseghir

The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced sce- narios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH&CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000s, with a flat-top plasma current of 12MA, and a fusion gain of Q ∼= 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10MA, and with a fusion gain of Q ∼= 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady- state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal pressure.

EPFL2013

Self-consistent simulation of plasma scenarios for ITER using a combination of 1.5D transport codes and free-boundary equilibrium codes

Karim Besseghir, Javier García Hernández, Jonathan Bryan Lister

Self-consistent transport simulation of ITER scenarios is a very important tool for the exploration of the operational space and for scenario optimization. It also provides an assessment of the compatibility of developed scenarios (which include fast transient events) with machine constraints, in particular with the poloidal field coil system, heating and current drive, fuelling and particle and energy exhaust systems. This paper discusses results of predictive modelling of all reference ITER scenarios and variants using two suites of linked transport and equilibrium codes. The first suite consisting of the 1.5D core/2D SOL code JINTRAC (Wiesen S. et al 2008 JINTRAC-JET modelling suite JET ITC-Report) and the free-boundary equilibrium evolution code CREATE-NL (Albanese R. et al 2003 ISEM 2003 (Versailles, France); Albanese R. et al 2004 Nucl. Fusion 44 999), was mainly used to simulate the inductive D-T reference Scenario-2 with fusion gain Q = 10 and its variants in H, D and He (including ITER scenarios with reduced current and toroidal field). The second suite of codes was used mainly for the modelling of hybrid and steady-state ITER scenarios. It combines the 1.5D core transport code CRONOS (Artaud J.F. et al 2010 Nucl. Fusion 50 043001) and the free-boundary equilibrium evolution code DINA-CH (Kim S.H. et al 2009 Plasma Phys. Control. Fusion 51 105007).

Iop Publishing Ltd2013

Rapid optimization of stationary tokamak plasmas in RAPTOR: demonstration for the ITER hybrid scenario with neural network surrogate transport model QLKNN

Federico Alberto Alfredo Felici, Olivier Sauter, Simon Van Mulders

This work presents a fast and robust method for optimizing the stationary radial distribution of temperature, density and parallel current density in a tokamak plasma and its application to first-principle-based modeling of the ITER hybrid scenario. A new solver is implemented in the RAPTOR transport code, enabling direct evaluation of the stationary solution to which the radial plasma profiles evolve. Coupled to a neural network emulation of the quasi-linear gyrokinetic QuaLiKiz transport model (QLKNN-hyper-10D), a first-principle-based estimate of the stationary state of the core plasma can be found at unprecedented computational speed (typically a few seconds on standard hardware). The stationary state solver is then embedded in a numerical optimization scheme, allowing the optimization of tokamak plasma scenarios in only a few minutes. The proposed method is applied to investigate the performance of ITER hybrid scenarios at different values of total plasma current, plasma density and pedestal height and for different power contributions in a heating mix consisting of electron cyclotron and neutral beam heating. Optimizing the radial distribution of electron cyclotron current drive (ECCD) deposition, the q profile is tailored to maximize the fusion gain Q, by maximizing the energy confinement predicted through the first-principles-based transport model, while satisfying q > 1, avoiding sawtooth oscillations. It is found that optimal use of ECCD in ITER hybrid scenarios is to deposit power as close to the core as possible, while maintaining sufficient off-axis current drive to keep q above 1. Upper limits for the fusion gain Q are shown to be constrained either by minimum power requirements for the separatrix power flow to maintain H-mode or by minimum current drive requirements for q profile tailoring. Finally, it is shown that the ITER hybrid scenario operating window is significantly extended by an upgrade of the electron cyclotron power to 40 MW.