Steady-state Operation Scenarios Research Articles

Electron ITB formation in EAST high poloidal beta plasmas under dominant electron heating

Plasma confinement and transport in tokamaks play a crucial role in the development of high poloidal beta steady-state operation scenarios. Therefore, it is very important to study the relevant mechanism of internal transport barriers (ITBs), which can help plasmas to obtain better confinement in order to achieve higher fusion gain. This paper mainly introduces the analysis of the characteristics of electron heat transport of discharges with ITB in high βP operation regime in Experimental Advanced Superconducting Tokamak (EAST). Based on the statistical analysis of stable discharges with βP > 1.5, it is found that there is an obvious bifurcation of the normalised electron temperature gradient (ETG) ( R/LTe ) in the range of βP = 2–2.2. Then the discharges of lower βP ( βP < 2, where the value of βP is below the bifurcation threshold) and of higher βP ( βP > 2.2, where the value of βP is above the threshold) were selected for analysis. The diagnostic data provided by Thomson scattering, x-ray crystal spectrometry and charge exchange recombination spectroscopy are used to provide reliable parameter profiles and then combined with the data of external magnetic probe measurements and the polarisation interferometer diagnosis system to fully reconstruct the balance. A relevant plasma current calculation model is used to calculate and analyse the current density profiles and power deposition, and then the transport analysis is carried out. Interestingly, in the higher βP discharges, it is found that the turbulence intensity provided by the CO2 laser collective scattering system gradually decreases and the normalised ETG R/LTe gradually increases. It is also found that the electron heat transport coefficient decreases in the discharge with higher βP and the growth rate of the electron-scale turbulence calculated by transport gyro-Landau fluid (TGLF) is significantly reduced. Meanwhile, similar conclusions are also obtained in the discharges when the βP is further increased.

Analysis and control of split-source current-type inverter for grid-connected applications

Current source inverters (CSIs) have been widely used for renewable energy sources integration with the utility grid. However, traditional CSIs provide only voltage-boost power conversion, which can be considered a limitation for grid-interfacing applications when consistent tracking of a widely varying input voltage is necessary. Hence, an additional front-end converter is usually utilized in an attempt to step down the input source voltage. Nevertheless, adding the dc-dc bucking stage would reduce the system reliability and its overall efficiency, increase the converter cost, and complicate the control scheme. In order to address the aforementioned shortcomings, this paper proposes a novel three-phase single-stage inverter, suitable for low-power applications, called split-source current-type inverter (SSCTI). The proposed converter can achieve high current boosting capability with lower component count and high-quality output current in comparison with other single-stage topologies. In addition, it utilizes the conventional modulation scheme of the CSI, while maintaining low current stresses on the inverter switches. Comprehensive analysis, modelling, and closed-loop control scheme of the proposed converter are developed. Also, a comparison between the proposed topology and several single-stage topologies is carried out. Moreover, an experimental prototype is built to validate the proposed converter analysis under steady-state operation and different transient scenarios.

Direct Multivariable Control for Modular Multilevel Converters

Modular multilevel converters (MMCs) have enabled new and demanding applications for power electronics in the high voltage/high power range. In these applications, very fast and tight control of all variables of the MMC is of prime importance. This article presents a new multivariable control (MVC) concept, which enables independent and extremely fast control of all essential variables. For each variable, a tolerance band can be specified, which is strictly met even under unexpected severe transients and disturbances. The minimized dead times achieved by the control are more than one order of magnitude smaller than the state of the art. The MVC has been implemented based on FPGA-hardware and digital signal processing of the measured values. Experimental results of the control performance and the robustness against parameter variation are presented, using a down scaled MMC with 96 submodules. These results include the analysis of steady-state operation, dynamic performance, and fault scenarios.

Fluid turbulence simulations of divertor heat load for ITER hybrid scenario using BOUT++

The BOUT++ six-field turbulence code is used to simulate the ITER 11.5 MA hybrid scenario and a brief comparison is made among ITER baseline, hybrid and steady-state operation (SSO) scenarios. Peeling–ballooning instabilities with different toroidal mode numbers dominate in different scenarios and consequently yield different types of ELMs. The energy loss fractions (ΔW ped/W ped) caused by unmitigated ELMs in the baseline and hybrid scenarios are large (∼2%) while the one in the SSO scenario is dramatically smaller (<1%), which are consistent with the features of type-I ELMs and grassy ELMs respectively. The intra ELM divertor heat flux width in the three scenarios given by the simulations is larger than the estimations for inter-ELM phase based on Goldston’s heuristic drift model. The toroidal gap edge melting limit of tungsten monoblocks of divertor targets imposes constraints on ELM energy loss, giving that the ELM energy loss fraction should be smaller than 0.4%, 1.0%, and 1.2% for ITER baseline, hybrid and SSO scenarios, correspondingly. The simulation shows that only the SSO scenario with grassy ELMs may satisfy the constraint.

Optimization of helicon wave off-axis current drive in CFETR tokamak

The China Fusion Engineering Testing Reactor (CFETR) is a tokamak reactor and aims to eventually reach DEMO relevant fusion power level of 1 GW. A lower single null configuration of the divertor design is proposed for the machine. To obtain the high performance operation of the plasma, both the hybrid and steady-state operating scenarios are suggested. In these scenarios, the plasma beta of 〈βe〉∼1.6% has been generally reached. In the paper, helicon wave propagation and current drive in CFETR are studied through GENRAY/CQL3D code. An analysis of the wave damping factor in CFETR plasma indicates that the electron Landau damping is dominant for the core plasma. According to the dimensionless parameter ξe and βe, a strong wave damping region is found. Based on the GENRAY simulation, the single pass power absorptions are generally obtained in both scenarios. Scanning of generated current on the launched poloidal angle indicates that high current drive efficiency is produced at θ∼300. For helicon wave with frequency f=700 MHz, calculations of the current drive show that off-axis current drive about 30 kA/MW and 50 kA/MW are obtained in the hybrid and steady-state scenarios respectively. Simultaneously, the lower plasma density and high temperature promote the high current generation. Moreover, the obtained current drive and its peak position are shown to be sensitive to the launched n//. Finally, the helicon wave with frequency f=700 MHz and launched n//∼2.0–3.0 is proposed to be a promising scheme for effective off-axis current drive in CFETR.

Joint DIII-D/EAST research on the development of a high poloidal beta scenario for the steady state missions of ITER and CFETR

Experimental and modeling investigations on the DIII-D and EAST tokamaks show the attractive transport and stability properties of fully noninductive, high poloidal-beta (βP) plasmas, and their suitability for steady-state operating scenarios in ITER and CFETR. A key feature of the high-βP regime is the large-radius (ρ > 0.6) internal transport barrier (ITB), often observed in all channels (ne, Te, Ti, rotation), and responsible for both excellent energy confinement quality and excellent stability properties. Experiments on DIII-D have shown that, with a large-radius ITB, very high βN and βP values (both ≥ 4) can be reached by taking advantage of the stabilizing effect of a nearby conducting wall. Synergistically, higher plasma pressure provides turbulence suppression by Shafranov shift, leading to ITB sustainment independent of the plasma rotation. Experiments on EAST have been used to assess the long pulse potential of the high-βP regime. Using RF-only heating and current drive, EAST achieved minute-long fully noninductive steady state H-mode operation with strike points on an ITER-like tungsten divertor. Improved confinement (relative to standard H-mode) and steady state ITB features are observed with a monotonic q-profile with qmin ∼ 1.5. Separately, experiments have shown that increasing the density in plasmas driven by lower hybrid wave broadens the q-profile, a technique that could enable a large radius ITB. These experimental results have been used to validate MHD, current drive, and turbulent transport models, and to project the high-βP regime to a burning plasma. These projections suggest the Shafranov shift alone will not suffice to provide improved confinement (over standard H-mode) without rotation and rotation shear. However, increasing the negative magnetic shear (higher q on axis) provides a similar turbulence suppression mechanism to Shafranov shift, and can help devices such as ITER and CFETR achieve their steady-state fusion goals.

Electrical characterisation of proton exchange membrane fuel cells stack using grasshopper optimiser

In this study, optimum values of unknown seven parameters of proton exchange membrane fuel cells (PEMFCs) stack are generated for the sake of appropriate modelling. An objective function is adopted to minimise the sum of square errors (SSE) between the experimental data and the corresponding estimated results. A novel application of grasshopper optimisation algorithm (GOA) is engaged to minimise the SSE subjects to set of inequality constraints. Three study cases of typical commercial PEMFCs stacks are demonstrated and verified under various steady-state operating scenarios. Necessary subsequent comparisons to new results by others found in updated state-of-the-art are made. Sensitivity analysis of defined parameters is carried out. It is found that the PEMFC model is susceptible to the deviations of optimised parameters as the errors are substantially disturbed which signifies the value of the GOA-based method. In addition, performance measures to indicate the robustness of the GOA-based methodology are pointed out. At this moment, dynamic model of the stack is addressed and incorporated to demonstrate its dynamic response. Detailed MATLAB/SIMULINK simulation model is implemented to study the PEMFC dynamic performance. The simulated test cases emphasise the viability and effectivity of the GOA-based procedure in steady-state and dynamic simulations.

Development of a systematic, self-consistent algorithm for the K-DEMO steady-state operation scenario

An optimum plasma pressure/current density profile and corresponding heating/current drive (H/CD) determination scheme is newly developed by integrating equilibrium, stability, confinement, and H/CD, self-consistently subject to maximize the fusion gain for Korean fusion demonstration reactor (K-DEMO) steady-state operation scenarios. The integrated plasma modeling package, FASTRAN/IPS, is adopted for the integrated numerical apparatus. The target pressure profile with a pedestal structure is investigated by varying its peaking, pedestal height and width as a first step. Formation of stable equilibria is evaluated by solving the Grad–Shafranov equation and checking linear MHD stability. For the case of potentially stable equilibrium, required external heating distribution is calculated by considering both power balance and external current drive alignment to reproduce the pressure profile of the stable equilibrium. Electron/ion temperature and poloidal flux evolutions are solved with the derived heating configuration to find a steady-state scenario and achieve self-consistent plasma profiles. A self-consistent target steady-state pressure and current profile parameters are proposed through designed systematic algorithm with fusion power PF = 2070 MW, fusion gain Q = 19.7, and normalized beta βN = 2.8 at toroidal field BT = 7.4 T and plasma current IP = 15.5 MA. Feasibility of fusion power PF = 3000 MW operation is also explored with enhanced density and temperature limit assumption.

Gaseous swelling of U3Si2 during steady-state LWR operation: A rate theory investigation

Rate theory simulations of fission gas behavior in U3Si2 are reported for light water reactor (LWR) steady-state operation scenarios. A model of U3Si2 was developed and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. The reliability of the model was examined by performing sensitivity analysis of the key parameters. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U3Si2 swelling behavior between 390K and 1190K. Under typical steady-state LWR operating conditions where U3Si2 temperature is expected to be below 1000K, intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U3Si2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.

Self-consistent modeling of CFETR baseline scenarios for steady-state operation

Integrated modeling for core plasma is performed to increase confidence in the proposed baseline scenario in the 0D analysis for the China Fusion Engineering Test Reactor (CFETR). The steady-state scenarios are obtained through the consistent iterative calculation of equilibrium, transport, auxiliary heating and current drives (H&CD). Three combinations of H&CD schemes (NB + EC, NB + EC + LH, and EC + LH) are used to sustain the scenarios with qmin > 2 and fusion power of ∼70–150 MW. The predicted power is within the target range for CFETR Phase I, although the confinement based on physics models is lower than that assumed in 0D analysis. Ideal MHD stability analysis shows that the scenarios are stable against n = 1–10 ideal modes, where n is the toroidal mode number. Optimization of RF current drive for the RF-only scenario is also presented. The simulation workflow for core plasma in this work provides a solid basis for a more extensive research and development effort for the physics design of CFETR.

Fast-ion transport in qmin&amp;gt;2, high-β steady-state scenarios on DIII-D

Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-qmin confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing βN and the noninductive current drive. However, in scenarios with qmin&amp;gt;2 that target the typical range of q95= 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. This enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable βN. In contrast, similar plasmas except with qmin just above 1 have approximately classical fast-ion transport. Experiments that take qmin&amp;gt;3 plasmas to higher βP with q95= 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-qmin scenario, the high βP cases have shorter slowing-down time and lower ∇βfast, and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, βN, and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q95, high-qmin plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes.

Simulation of Lower Hybrid Current Drive for DEMO

Lower hybrid (LH) wave is not only convenient to generate a flat or reversed magnetic shear profiles, but also helps one to explore scenarios for steady-state tokamak operation with improved confinement. Here with LSC code (lower hybrid simulation code), we calculate density and temperature profiles, relative power of injected wave and current wave lunch for two options of DEMO at the launched LH wave frequency 5 GHz. Two plasma scenarios pertaining to two different DEMO options, known as pulsed (option 1) and steady-state (option 2) models, have been analyzed. We perceive that power deposition by using lower hybrid wave injection mainly takes place near the edge of plasma and approximately in more peripheral region for both of options but has approximately higher efficiency for option 1 compared to option 2. About current wave lunch, a major part of that is close to the plasma edge for both of options. We have some considerable parts that reach to internal layers for option 1 and then current drive mainly takes place in a wider, more peripheral region for option 1.

Low-recycling conditions and improved core confinement in steady-state operation scenarios in JET (Joint European Torus)

In this paper, we discuss the phenomena that link particle recycling from the vessel walls in the L-mode during discharge start-up and the core confinement in the H-mode during the subsequent main heating phase. We consider available data of JET experiments that aimed at approaching fully non-inductive ITER-relevant steady-state conditions and show that the high electron temperature produced at the edge by a low recycling during start-up tends to favour the build-up of high normalized β(βN) regimes in the H-mode, the confinement being improved in a large plasma volume. To provide an insight into this complex phenomenon we have modelled the relation between particle recycling in the scrape-off layer and the evolution of plasma transport, plasma current density and shear as well as the stability properties for such experimental conditions. The results confirm the existence of a link between the confinement in the H-mode phase and the values at the edge of electron temperature, bootstrap current density and local magnetic shear during start-up. Such a link could favour these regimes to be self-sustained in time.

Implementation of IEEE-1588 timing and synchronization for ATCA control and data acquisition systems

Control and data acquisition (C&DA) systems for Fusion experiments are required to provide accurate timing and synchronization (T&S) signals to all of its components. IPFN adopted PICMG's Advanced Telecommunications Computing Architecture (ATCA) industry standard to develop C&DA instrumentation. ATCA was chosen not only for its high throughput characteristics but also for its high availability (HA) features which become of greater importance in steady-state operation scenarios. However, the specified ATCA clock and synchronization interface may be too limited for the timing and synchronization needs in advanced Physics experiments. Upcoming specification extensions, developed by the “xTCA for Physics” workgroups, will contemplate, among others, a complementary timing specification, developed by the PICMG xTCA for Physics IO, Timing and Synchronization Technical Committee. The IEEE-1588 Precision Time Protocol (PTP) over Ethernet is one of the protocols, proposed by the Committee, aiming for precise synchronization of clocks in measurement and control systems, based on low jitter and slave-to-slave skew criteria.The paper presents an implementation of IEEE-1588 over Ethernet, in an ATCA hardware platform. The ATCA hardware consists of an Advanced Mezzanine Card (AMC) quad-carrier front board with PCI Express switching. IEEE-1588 is to be implemented on a Virtex-6 FPGA. Ethernet connectivity with the remote master clock is located on the rear transition module (RTM). The generated synchronized clock and absolute time in IRIG-B format are distributed to all systems endpoints by a cross-point switch which is also implemented on the FPGA.

Real-Time Control of Advanced Scenarios for Steady-State Tokamak Operation

This lecture was given at the 4th ITER International Summer School in Austin, Texas, May 31–June 4, 2010. It reviews the recent experimental and modeling progress made to design real-time kinetic and magnetic profile control of advanced scenarios for steady-state tokamak operation. The lecture addresses four challenging issues that need to be resolved and that are open to future research activities: (a) how to operate a tokamak in a continuous manner, (b) how to control the core kinetic and magnetic profiles of tokamak plasmas, (c) how to control the fusion burn in plasmas with dominant self-generated bootstrap non-inductive current and fusion-born alpha heating, and (d) how to control simultaneously core and edge plasma parameters.

Discrete Alfvén eigenmodes in international thermonuclear experimental reactor operations with negative magnetic shear

Discrete Alfvén eigenmodes in steady-state operation scenarios with negative magnetic shear in the international thermonuclear experimental reactor are investigated in this paper. These magnetohydrodynamic eigenmodes are trapped by the α-induced potential wells along the magnetic field line. Here α = –q2Rdβ/dr with q being the safety factor, β the ratio between plasma and magnetic pressures, and R the major radius, and r the minor radius. Due to negligible continuum damping via wave energy tunneling, these Alfvén eigenmodes could be readily destabilized by energetic particles.

Numerical simulation on edge localized mode control capability of resonant magnetic perturbation in the KSTAR tokamak

Numerical simulations are carried out to investigate the applicability of resonant magnetic perturbation (RMP) to KSTAR plasmas for a possible control of edge localized mode (ELM) to suppress or mitigate its damages to divertor materials. For the verification of the feasibility of RMP application, magnetic island configurations, resonant normal fields, magnetic island widths and Chirikov parameters are calculated for two types of KSTAR operation scenarios: steady state and hybrid. Field error correction (FEC) coils in KSTAR are considered to produce externally perturbed magnetic fields for RMP, and the directions of coil currents determine the toroidal mode n and the parity (even or odd). The RMP configurations are described by vacuum superposition of the equilibrium magnetic fields and the perturbed ones induced by FEC coils. The numerical simulations for n = 2 toroidal mode in both operation scenarios show that when the pitches of the equilibrium and perturbed magnetic fields are well aligned, magnetic islands are formed for a series of m poloidal modes and the adjacent islands are overlapped to generate a stochastic layer in the edge region. Even parity turns out to be more effective in making the magnetic islands overlapped to become stochastic field lines in the steady-state operation, while odd parity in the hybrid operation. The formation of the stochastic layer is verified by the calculated Chirikov parameters, which also give basic information on the current requirement of FEC coils. Additionally, lobe structures of stochastic field lines are found in the edge region extended to the divertor plate in the hybrid scenario. Based on the standard vacuum criteria for RMP, the simulation results indicate that the FEC coils will be feasible for control of ELMs and mitigation of divertor heat load by RMP in both steady-state and hybrid operation scenarios.

Cyclic scenarios for steady-state operation of tokamak reactors

A new concept of steady-state scenario for tokamak reactors is proposed. It is based on cyclic operations, alternating phases of positive and negative loop voltage with no magnetic flux consumption on average. Localized non-inductive current drive by electron cyclotron waves is used to trigger and sustain an internal transport barrier (ITB), whereas neutral beam current drive is used to periodically recharge the tokamak transformer. The fact of operating in cycles relaxes the hard constraint of simultaneous fusion performance maximization and full non-inductive operation, within the MHD stability limits. Integrated modelling simulations are performed to apply this concept for the ITER steady-state regime. A linear MHD analysis of the instabilities that could appear in this type of scenario is performed, showing that MHD stability would be strongly improved with respect to a steady regime with a strong ITB.

Simulations of KSTAR high performance steady state operation scenarios

We report the results of predictive modelling of high performance steady state operation scenarios in KSTAR. Firstly, the capabilities of steady state operation are investigated with time-dependent simulations using a free-boundary plasma equilibrium evolution code coupled with transport calculations. Secondly, the reproducibility of high performance steady state operation scenarios developed in the DIII-D tokamak, of similar size to that of KSTAR, is investigated using the experimental data taken from DIII-D. Finally, the capability of ITER-relevant steady state operation is investigated in KSTAR. It is found that KSTAR is able to establish high performance steady state operation scenarios; βN above 3, H98(y, 2) up to 2.0, fBS up to 0.76 and fNI equals 1.0. In this work, a realistic density profile is newly introduced for predictive simulations by employing the scaling law of a density peaking factor. The influence of the current ramp-up scenario and the transport model is discussed with respect to the fusion performance and non-inductive current drive fraction in the transport simulations. As observed in the experiments, both the heating and the plasma current waveforms in the current ramp-up phase produce a strong effect on the q-profile, the fusion performance and also on the non-inductive current drive fraction in the current flattop phase. A criterion in terms of qmin is found to establish ITER-relevant steady state operation scenarios. This will provide a guideline for designing the current ramp-up phase in KSTAR. It is observed that the transport model also affects the predictive values of fusion performance as well as the non-inductive current drive fraction. The Weiland transport model predicts the highest fusion performance as well as non-inductive current drive fraction in KSTAR. In contrast, the GLF23 model exhibits the lowest ones. ITER-relevant advanced scenarios cannot be obtained with the GLF23 model in the conditions given in this work. Finally, ideal MHD stability is investigated for the ITER-relevant advanced scenarios in KSTAR. The methods and results presented in this paper are expected to contribute to improving the ITER and beyond ITER predictive simulations.

Steady state ECRH for next step superconducting fusion devices

Controlled plasma start-up, heating towards thermonuclear fusion temperatures and steady state discharge control in advanced configurations is a major challenge for both, stellarators and tokamaks. Electron Cyclotron Resonance Heating (ECRH) and current drive (ECCD) plays a key role in the steady state operation scenarios of the ITER and JT-60 SA tokamaks as well as for the W7-X stellarator. The physics demands as well as the key technology of the different ECRH-systems, which are similar in frequency and have continuous wave (cw) capability, are presented. Advanced solutions for future ECRH-systems are discussed.