CORSICA: A comprehensive simulation of toroidal magnetic-fusion devices. Final report to the LDRD Program

Crotinger, J.A.; LoDestro, L.L.; Pearlstein, L. D.; Tarditi, Alfonso G.; Casper, T. A.; Hooper, E. B.

doi:10.2172/522508

Cited by 89 publications

(95 citation statements)

References 22 publications

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“…The simulations have been performed using three free-boundary transport simulation codes, TRANSP/TCS, 58 CORSICA, 47 and DINA. The simulations use a prescribed density profile evolution and the Coppi-Tang 41 thermal transport model to evolve the plasma for the baseline 15 MA inductive scenario.…”

Section: Progress In Scenario Modelling a Iter Baseline Operatinmentioning

confidence: 99%

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Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Sips

Giruzzi²,

Ide

et al. 2014

Physics of Plasmas

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The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ∼ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ∼ 1.0 only for βN = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to ne/nGW ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.

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Section: Progress In Scenario Modelling a Iter Baseline Operatinmentioning

confidence: 99%

“…In preparation for operation in hydrogen and helium, several simulations have been performed with CORSICA 47 and JINTRAC. 48 The time-dependent simulations were performed for reduced operation at 7.5 MA and 2.65 T as well as at 15 MA and 5.3 T. Key in these simulations is the access to H-mode, without the a-heating from DT reactions.…”

Section: -13mentioning

confidence: 99%

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Sips

Giruzzi²,

Ide

et al. 2014

Physics of Plasmas

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“…These results include ITPA joint experiments coupled with ITER modeling and model validation, and suggest that improving the vertical control capability of the ITER baseline design may be important in order to provide robustness comparable to that of presently operating devices. Modeling and simulation includes use of the LLNL Caltrans code [3], the GA TokSys environment [4], and the MIT Alcasim environment [5]. The present study focuses on "machineindependent" performance metrics that describe the proximity to practical controllability limits rather than ideal stability boundaries.…”

Section: Introductionmentioning

confidence: 99%

Experimental vertical stability studies for ITER performance and design guidance

et al. 2009

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Abstract. Operating experimental devices have provided key inputs to the design process for ITER axisymmetric control. In particular, experiments have quantified controllability and robustness requirements in the presence of realistic noise and disturbance environments, which are difficult or impossible to characterize with modeling and simulation alone. This kind of information is particularly critical for ITER vertical control, which poses the highest demands on poloidal field system performance, since the consequences of loss of vertical control can be

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“…The coupling between the core transport solver and the external components usually occurs at memory level in view of the frequent calls that the core solver has to make to its modules while integrating the transport equations in time. A non-exhaustive list of these codes are ONETWO [1,2], TRANSP/PTRANSP [3], TOP-ICS [4], ASTRA [5], CRONOS [6], CORSICA [7], FAS-TRAN [8], FACETS [9].…”

Section: Transport-based Frameworkmentioning

confidence: 99%

Integrated Modeling of Tokamak Experiments with OMFIT

Meneghini

Lao

2013

Plasma and Fusion Research

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One Modeling Framework for Integrated Tasks (OMFIT) is a framework that allows data to be easily exchanged among different codes by providing a unifying data structure. The main idea at the base of OMFIT is to treat files, data and scripts as a uniform collection of objects organized into a tree structure, which provides a consistent way to access and manipulate such collection of heterogeneous objects, independent of their origin. Within the OMFIT tree, data can be copied/referred from one node to another and tasks can call each other allowing for complex compound task to be built. A top-level Graphical User Interface (GUI) allowing users to manage tree objects, carry out simulations and analyze the data either interactively or in batch. OMFIT supports many scientific data formats and when a file is loaded into the framework, its data populates the tree structure, automatically endowing it with many potential uses. Furthermore, seamless integration with experimental management systems allows direct manipulation of their data. In OMFIT modeling tasks are organized into modules, which can be easily combined to create arbitrarily-large multi-physics simulations. Modules inter-dependencies are seamlessly defined by variables referencing tree locations among them. Creation of new modules and customization of existing ones is encouraged by graphical tools for their management and an online repository. High level Application Programmer Interfaces (APIs) enable users to execute their codes on remote servers and creation application-specific GUIs. Finally, within OMFIT it is possible to visualize experimental and modeling data for both quick analysis and publication purposes. Examples of application to the DIII-D tokamak are presented.

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CORSICA: A comprehensive simulation of toroidal magnetic-fusion devices. Final report to the LDRD Program

Cited by 89 publications

References 22 publications

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Experimental vertical stability studies for ITER performance and design guidance

Integrated Modeling of Tokamak Experiments with OMFIT

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