Simulation and Analysis of the Hybrid Operating Mode in ITER

Kessel, C.E.; Budny, R.; Indireshkumar, K.

doi:10.1109/fusion.2005.252943

Cited by 3 publications

(3 citation statements)

References 16 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A similar result for Hybrid scenario plasmas in ITER was reported in [50]. It was found that very large pedestal temperatures ( 10 keV) are needed to produce peaked-density Hybrid plasmas with high β n ( 3).…”

Section: Peaked Densitysupporting

confidence: 79%

Predictions of H-mode performance in ITER

Budny¹,

André²,

Bateman³

et al. 2008

Nucl. Fusion

121

View full text Add to dashboard Cite

Time-dependent integrated predictive modelling is carried out using the PTRANSP code to predict fusion power and parameters such as alpha particle density and pressure in ITER H-mode plasmas. Auxiliary heating by negative ion neutral beam injection and ion-cyclotron heating of He 3 minority ions are modelled, and the GLF23 transport model is used in the prediction of the evolution of plasma temperature profiles. Effects of beam steering, beam torque, plasma rotation, beam current drive, pedestal temperatures, sawtooth oscillations, magnetic diffusion and accumulation of He ash are treated self-consistently. Variations in assumptions associated with physics uncertainties for standard base-line DT H-mode plasmas (with I p = 15 MA, B TF = 5.3 T and Greenwald fraction = 0.86) lead to a range of predictions for DT fusion power P DT and quasi-steady state fusion Q DT (≡P DT /P aux ). Typical predictions assuming P aux = 50-53 MW yield P DT = 250-720 MW and Q DT = 5-14. In some cases where P aux is ramped down or shut off after initial flat-top conditions, quasi-steady Q DT can be considerably higher, even infinite. Adverse physics assumptions such as the existence of an inward pinch of the helium ash and an ash recycling coefficient approaching unity lead to very low values for P DT . Alternative scenarios with different heating and reduced performance regimes are also considered including plasmas with only H or D isotopes, DT plasmas with toroidal field reduced 10% or 20% and discharges with reduced beam voltage. In full-performance D-only discharges, tritium burn up is predicted to generate central tritium densities up to 10 16 m −3 and DT neutron rates up to 5 × 10 16 s −1 , compared with the DD neutron rates of 6 × 10 17 s −1 . Predictions with the toroidal field reduced 10% or 20% below the planned 5.3 T and keeping the same q 98 , Greenwald fraction and β n indicate that the fusion yield P DT and Q DT will be lower by about a factor of two (scaling as B 3.5 ).

show abstract

Section: Peaked Densitysupporting

confidence: 79%

Predictions of H-mode performance in ITER

Budny¹,

André²,

Bateman³

et al. 2008

Nucl. Fusion

121

View full text Add to dashboard Cite

show abstract

“…The simulations using the GLF23 core transport model show in tables 1 and 2 that for the ITER hybrid, a 5 keV pedestal assumption leads to β N value in the range 2.1-2.6, while present hybrid experiments indicate that values closer to 3.0 can be achieved, just below the n = 1 external kink no wall β limit [3,4]. Previous ITER hybrid simulations showed that temperature pedestal values of 7-9 keV, with GLF23 core transport, were required to reach a β N value of about 3.0 [48]. These pedestal values are considered outside the range of the pedestal database predictions for ITER [45].…”

Section: Iter Hybrid Issuesmentioning

confidence: 59%

Simulation of the hybrid and steady state advanced operating modes in ITER

et al. 2007

View full text Add to dashboard Cite

Abstract. Integrated simulations are done to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in ITER. Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS, and ASTRA. The hybrid operating mode is simulated using the GLF23 energy transport model. The injected powers are limited to the negative ion neutral beam (NBI), ion cyclotron (ICRF), and electron cyclotron (EC). Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison among the simulations. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources for the hybrid with the possible addition of lower hybrid (LH). These simulations will be presented and compared with particular focus on code to code results when using the same energy transport model, and within the same code when using different energy transport models

show abstract

“…The boundary-evolving and predictive capabilities of TSC are used to develop the full time evolution of an ITER discharge, and TRANSP is used for correcting the heating and current drive profiles with more sophisticated source models and for computing fast ion parameters such as grad(β α ). Results for ITER hybrid plasmas have been published [338] and have been submitted to the ITPA profile database. For example the effects of applying the ITER NBI system at various different injection angles into a hybrid plasma has been simulated and the results shown in figure 46.…”

Section: Main Results From Simulations For Itermentioning

confidence: 99%

Chapter 6: Steady state operation

Gormezano¹,

et al. 2007

View full text Add to dashboard Cite

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

show abstract

Simulation and Analysis of the Hybrid Operating Mode in ITER

Cited by 3 publications

References 16 publications

Predictions of H-mode performance in ITER

Predictions of H-mode performance in ITER

Simulation of the hybrid and steady state advanced operating modes in ITER

Chapter 6: Steady state operation

Contact Info

Product

Resources

About