Future Fusion Devices Research Articles

Assessment of the JET ICRH system performance since 2000

Abstract The paper provides an assessment of the ion-cyclotron resonance heating (ICRH) system performance on JET since the year 2000. The vast amount of collected data offer an insight into the historical challenges and trends in the ICRH system performance encompassing the transition from carbon (JET-C) to beryllium & tungsten ITER-like wall (JET-ILW) operations, the deuterium–tritium experiments (DTE2 & DTE3) and introduction of new RF antenna & matching systems. The best achieved operational parameters are reported and statistics on the RF plant reliability and performance is analysed. Antenna-plasma coupling is identified as the dominant factor critical to all the aspects of the ICRH system behaviour; parametric dependencies of coupling resistance on plasma parameters and the RF plant settings are discussed and the key role of local electron concentration profiles close to the antennas is highlighted. Following confident antenna performance at high RF voltages over the recent decade, observations are presented suggesting improved electrical strength of the RF vacuum components after the JET-C to JET-ILW transition; this is tentatively attributed to the reduction of dust levels in the JET vessel. Statistics on application rate and typical origins of the RF amplifier failures and protection power limits is presented indicating that the amplifier issues noticeably affected the high-power ICRH operations. Performance comparison is provided for different RF antenna & matching systems installed at JET since 2000 including the original system, two load-tolerant systems based on the 3 dB hybrid and external conjugate-T power-splitters, and the ITER-like antenna. The paper could be of interest both as a summary of technical challenges, constraints and achievements related to the ICRH application on JET and as a reference for design and operations of high-power RF systems in future fusion devices.

Fuel recycling feedback control via real-time boron powder injection in EAST with full metal wall

Abstract A feedback control of fuel recycling via real-time boron powder injection, addressing the issue of continuously increasing recycling in long-pulse plasma discharges, has been successfully developed and implemented on EAST tokamak. The feedback control system includes four main parts: the impurity powder dropper (IPD), a diagnostic system measuring fuel recycling level represented by Dα emission, a plasma control system (PCS) implementing the Proportional Integral Derivative (PID) algorithm, and a signal converter connecting the IPD and PCS. Based on this control system, both active control and feedback control experiments have recently been performed on EAST with a full metal wall. The experimental results show that the fuel recycling can be gradually reduced to lower level as PCS control voltage increases. In the feedback control experiments, it is also observed that the Dα emission is reduced to the level below the target Dα value by adjusting boron injection flow rate, indicating successful implementation of the fuel recycling feedback control on EAST. This technique provides a new method for fuel recycling control of long pulse and high parameter plasma operations in future fusion devices.

Numerical simulation of [formula omitted] drift effects on fueling pellet ablation and transport in EAST

The pellet injection technique is considered a primary method for fueling future fusion devices. The inclusion of the E×B particle drift effect is essential in the simulation to reproduce the actual particle transport of injected pellets and the changes in energy flux in the divertor region. In this study, the dynamic neutral gas shield model is coupled with the edge plasma code SOLPS-ITER, utilizing initial data referencing from EAST experiments. The investigation focuses on the effects of the E×B drift on the ablation of injected fuel pellets, particle transport, and heat flux in the divertor region. The simulation results show that the poloidal component of the E×B drift velocity causes the dispersion of ions and electrons produced by the ionization of ablated atoms from the pellet, which would accumulate around the pellet if the E×B drift is not considered. Under unfavorable toroidal magnetic field conditions, a significant accumulation of ions ▪ produced by the ionization of D atoms introduced by the injected pellet is observed near the outer target. Moreover, both the change and duration of energy flux variations in the outer target region are higher when compared with the case without E×B drift.

Real-time plasma boundary shape reconstruction using visible camera on EAST tokamak

Abstract Accurate plasma shape reconstruction is crucial for the stable operation of tokamak devices. In this study, a plasma optical boundary reconstruction system is constructed on the Experimental Advanced Superconducting Tokamak (EAST), and real-time shape reconstruction based on an optical method is realized for the first time on EAST. First, a boundary extraction algorithm based on gray features is designed to extract the optical boundary from plasma optical images. Then, to reconstruct the optical boundary in the tokamak coordinate system, a camera calibration algorithm is developed based on feature point matching, and a coordinate mapping algorithm is designed based on plasma geometric features. Then, the system reconstructs the plasma boundary shape based on images acquired by a camera. Finally, the system is deployed in real-time on EAST, and its reconstruction rate meets the requirement of the EAST plasma shape control system. This study validates the feasibility of using an optical boundary shape reconstruction method on a tokamak device for real-time plasma shape reconstruction, providing a new shape reconstruction approach during steady-state discharge in future fusion devices.

Deuterium gas-driven permeation and retention in ZrC dispersion-strengthened W and pure W

ZrC dispersion-strengthened W (WZrC) exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. However, the hydrogen isotope behaviors in WZrC are not well understood. In this study, experiments of deuterium gas-driven permeation (GDP) through 0.5 wt.% ZrC dispersion-strengthened W and pure W in the temperature range of 923–1123 K have been performed. The deuterium permeability of WZrC is close to pure W. In comparison, the deuterium diffusivity of WZrC is significantly lower than pure W in the experimental temperature range. Besides, thermal desorption spectroscopy (TDS) was used to investigate the deuterium retention properties of both W materials after static deuterium gas charging. The results reveal two deuterium desorption peaks with detrapping energies of 0.91 ± 0.09 eV and 1.07 ± 0.28 eV for WZrC, while one peak of 1.12 ± 0.31 eV for pure W. The deuterium retention of WZrC is approximately one order of magnitude higher than pure W. The study further discusses the impact of microstructural features on the transport and retention properties of deuterium in these materials.

Liquid lithium divertor analysis using coupled plasma material interaction model

A liquid lithium divertor can improve performance of future fusion devices by creating efficient power exhaust and improving the energy confinement via pumping of the hydrogen isotopes. In addition, significantly higher heat fluxes can be handled if controlled vapor shielding is used to redistribute the divertor heat flux over a wider area. Design and optimization of such a system calls for an analysis model which includes a strong two-way coupling between the plasma and divertor material. The incoming plasma heat and particle flux will affect the divertor surface temperature, which is a defining factor of the lithium evaporative and sputtered flux going into the plasma. Results of the coupled model based on the plasma edge code SOLPS-ITER and the computational fluid dynamics (CFD) code ANSYS-CFX will be presented for different configurations. An analytical slab flow model is used as a heat transfer boundary condition for SOLPS, defining particle flux from the wall via calculation of the surface temperature. At the final step, results of the SOLPS analysis are verified using a 3D CFD magnetohydrodynamics (MHD) analysis which uses heat and particle flux from SOLPS as a boundary condition. In addition to plasma heat flux, both analytical and CFD temperature models include several plasma material interaction effects, such as lithium evaporation, condensation and sputtering based on deuterium target flux. New adatom sputtering model based on the available experimental data is presented. Analytical model is expanded to include free surface axisymmetric configurations. Results of parametric studies of the divertor configurations with different lithium inlet temperature and velocity will be presented leading to the optimal design resulting in the lowest possible lithium contamination in the core.

Toroidal distribution of heat load on castellated plasma facing components for lower divertor target in EAST

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

J-TEXT achievements in turbulence and transport in support of future device/reactor

Following the reconstruction of the TEXT tokamak at Huazhong University of Science and Technology in China, renamed as J-TEXT, a plethora of experimental and theoretical investigations has been conducted to elucidate the intricacies of turbulent transport within the tokamak configuration. These endeavors encompass not only the J-TEXT device’s experimental advancements but also delve into critical issues pertinent to the optimization of future fusion devices and reactors. The research includes topics on the suppression of turbulence, flow drive and damping, density limit, non-local transport, intrinsic toroidal flow, turbulence and flow with magnetic islands, turbulent transport in the stochastic layer, and turbulence and zonal flow with energetic particles or helium ash. Several important achievements have been made in the last few years, which will be further elaborated upon in this comprehensive review.

An integral approach to plasma-wall interaction modelling for EU-DEMO

Abstract The integral approach to plasma-wall interaction (PWI) modelling for DEMO is presented, which is a part of EUROfusion Theory and Advanced Simulation Coordination (E-TASC) activities that were established to advance the understanding and predictive capabilities for modelling of existing and future fusion devices using a modern advanced computing approach. In view of DEMO design, the aim of PWI modelling activities is to assess safety-relevant information regarding erosion of plasma-facing components (PFC), including its impact on plasma contamination, dust production, fuel inventory, and material response to transient events. This is achieved using a set of powerful and validated computer codes that deal with particular PWI aspects and interact with each other by means of relevant data exchange. Steady state erosion of tungsten PFC and subsequent transport and re-deposition of eroded material are simulated with the ERO2.0 code using a DEMO plasma background produced by dedicated SOLPS-ITER simulations. Dust transport simulations in steady state plasma also rely on the respective SOLPS-ITER solution and are performed with the MIGRAINe code. On the way to improved simulations of tungsten erosion in the divertor of DEMO, relevant high density sheath models are being developed based on Particle-in-Cell (PIC) simulations with the state-of-the-art BIT code family. PIC codes of the SPICE code family, in turn, provide relevant information on multi-emissive sheath physics, such as semi-empirical scaling laws for field-assisted thermionic emission. These scalings laws are essential for simulations of material melting under transient heat loads that are performed with the recently developed MEMENTO code, the successor of MEMOS-U. Fuel retention simulations assess tritium retention in tungsten and structural materials and permeation to coolant accounting for neutron damage, in particular for divertor monoblocks of different sizes, using the FESTIM code, and for the first wall, in particular in view of tritium self-sufficiency, using the TESSIM code. Respective code-code dependencies and interactions, as well as modelling results achieved to date are discussed in the contribution.

X-ray sources for in situ wavelength calibration of x-ray imaging crystal spectrometers.

X-ray sources for a range of wavelengths are being considered for in situ calibration of X-ray Imaging Crystal Spectrometers (XICSs) and for monitoring line shifts due to changes in the crystal temperature, which can vary during experimental operation over a day [A. Ince-Cushman etal., Rev. Sci. Instrum. 79, 10E302 (2008), L. Delgado-Aparicio etal., Plasma Phys. Control. Fusion 55, 125011 (2013)]. Such crystal temperature dependent shifts, if not accounted for, could be erroneously interpreted as Doppler shifts leading to errors in plasma flow-velocity measurements. The x-ray sources encompass characteristic x-ray lines falling within the wavelength range of 0.9-4.0Å, relevant for the XICSs on present and future fusion devices. Several technological challenges associated with the development of x-ray sources for in situ calibration are identified and are being addressed in the design of multiple x-ray tubes, which will be installed inside the spectrometer housing of the XICS for the JT-60SA tokamak. These x-ray sources will be especially useful for in situ calibration between plasma discharges. In this paper, laboratory experiments are described that were conducted with a Cu x-ray source, a heated quartz (102) crystal, and a pixelated Pilatus detector to measure the temperature dependent shifts of the Cu Kα1 and Kα2 lines at 1.5405 and 1.5443Å, respectively, and to evaluate the 2d-lattice constant for the Bragg reflecting crystal planes as a function of temperature, which, in the case of in situ wavelength calibration, would have to be used for numerical analysis of the x-ray spectra from the plasma.

Overview of physics results from the ADITYA-U tokamak and future experiments

The ADITYA upgrade (ADITYA-U), a medium-sized (R0=75 cm,a=25 cm) conventional tokamak facility in India, has been consistently producing experiments findings by using circular and shaped-plasmas. Recognizing the plasma parameters aligning closely with the design parameters of circular limited plasmas, ADITYA-U shifted its focus toward exploring the operational regime for experimentation on saw-tooth and MHD phenomena. Moreover, ADITYA-U has made consistent advancements toward conducting preliminary plasma shaping experiments through the activation of top and bottom divertor coils utilizing hydrogen as well as deuterium fuels. Confinement is improved by a factor of ∼1.5 in D2 plasmas when compared to H 2 plasmas of ADITYA-U. Further, ADITYA-U operations emphasize preventing disruptions and runaway electrons (REs) to ensure safe operations for future fusion devices. Significant suppression of REs has been achieved in ADITYA-U with the application of pulsed localized vertical magnetic field (LVF) perturbation, thereby establishing the technique’s independence from the tokamak device. The successful RE mitigation requires a critical threshold of LVF pulse magnitude, which is approximately 1% of the toroidal magnetic field, and a minimum duration of ∼5 ms. Apart from this, several novel findings have been achieved in the ADITYA-U experiments, including the modification of sawtooth duration through gas-puff, the emergence of MHD-induced geodesic acoustic mode-like oscillations, the propagation of fast heat pulses induced by MHD activity, the control of RE dynamics through Gas-puffs, the propagation of pinch-driven cold-pulses, the transport and core accumulations of argon impurities, the mass dependency of plasma toroidal rotation and the detection of ‘RICE’ scaling, as well as the characterization of edge plasma using wall conditioning methods, such as glow discharge cleaning using a combination of Ar-H 2 mixture, localized wall cleaning by electron cyclotron resonant plasma, and the development of machine learning-based disruption predictions, will be discussed in this paper.

Pre-conceptual design and proof of principle assessments of self-cooled Toroidally symmetric lead-lithium (TSLL) blanket

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

Development of high impact toughness Cu/ODS-Cu joints using HIP bonding process for the preparation of W/Cu/ODS-Cu monoblock divertor

Compared to CuCrZr, oxide dispersion strengthened copper alloy (ODS-Cu) exhibits higher stability of properties under irradiation and exposure to elevated temperatures, demonstrating broad application prospects in divertor components. Oxygen-free high thermal conductivity copper (Cu-OFHC) has been frequently employed as an interlayer between W and Cu-based alloy in the fabrication of W/Cu divertor components. This study investigates the effect of joining temperature together with the addition of a Ni interlayer on the interface microstructure and mechanical properties of the Cu-OFHC/ODS-Cu joints. As the joining temperature increased from 680 ℃ to 900 ℃, the interface bonding ratio of the Cu-OFHC/ODS-Cu joints improved from 40.4% to 95.8%, and the impact toughness increased from 23.4 J/cm2 to 133.1 J/cm2. With the addition of a Ni interlayer, the interface bonding ratio increased from 40.4% to 90.3%, and the impact toughness improved from 23.4 J/cm2 to 122.5 J/cm2. Increasing the joining temperature or adding a Ni interlayer effectively reduced interfacial voids, enhanced the interface bonding ratio, and consequently improved the impact toughness of Cu-OFHC/ODS-Cu joints. Then, the W/Cu/ODS-Cu monoblock mock-ups with good interfacial bonding were successfully fabricated under two conditions: at a joining temperature of 900 ℃ without an interlayer and at 680 ℃ with a Ni interlayer. These results provide a fundamental understanding for achieving high-quality Cu-OFHC/ODS-Cu joints and offer technical support for the engineering preparation of W/Cu/ODS-Cu components in future fusion devices.

Design and implementation of intelligent control system for non-evaporable getter pumps

Abstract An intelligent control system for Non-Evaporable Getter (NEG) pumps was designed and developed. The main functions of the system include remote monitoring and remote control of key parameters of the NEG power supply, intelligent control of NEG working status according to NEG saturation status and vacuum chamber status, especially intelligent warning and protection of NEG from vacuum leaks and NEG supersaturation. The intelligent control system is proven stable and reliable, which provides a feasible logic architecture and control design for NEG pump application in future fusion devices.

Depth-resolved deuterium retention analysis in displacement-damaged tungsten using laser-induced breakdown spectroscopy

Fuel retention in plasma-facing components (PFCs) is a critical issue in future nuclear fusion reactors operating with Deuterium-Tritium (DT) regarding nuclear safety and fulfillment of the T cycle. However, during DT plasma operation, highly energetic neutrons will induce damage in the lattice of W PFCs causing enhanced fuel retention in defects or traps. Laser-Induced Breakdown Spectroscopy (LIBS) is a potential tool to monitor the T-content in situ in PFCs of future nuclear fusion devices. This article presents an ex situ study on pre-damaged W material after D plasma exposure to qualify the method and mimic conditions expected in a reactor. ITER grade W samples were displacement-damaged by 10.8 MeV W ions to a damage dose of 0.23 dpa and exposed to low temperature deuterium plasma at low energy in PlaQ. The resulting deuterium concentration was analyzed by using 3He Nuclear Reaction Analysis (depth resolution of ≈150 nm) as a well-established method, and LIBS (picosecond laser pulses, depth resolution of 15 nm). The sample with the highest deuterium concentration showed a deuterium-rich zone up to a depth of 1.13 μm using both techniques. This is close to the expected W ion-induced damage depth of ≈1 μm. The results imply that LIBS as an in situ technique for tritium monitoring could be a viable option for a reactor.

System size scaling of triangularity effects on global temperature gradient-driven gyrokinetic simulations

In this work, we explore the triangularity effects on turbulent transport employing global gyrokinetic simulations performed with the ORB5 code. Numerous experiments on the Tokamak á Configuration Variable (TCV) and, more recently, on the DIII-D machine, have demonstrated superior confinement properties in L-mode of negative triangularity (NT) over positive triangularity (PT) configuration. This presents a particularly attractive scenario, as L-mode operation eliminates or significantly mitigates the presence of hazardous edge-localized modes (ELMs). However, a full theoretical understanding of all these observations remains elusive. Specifically, questions remain about how NT improvements can extend to the core where triangularity is very small, and whether these improvements can scale to larger devices. This paper addresses these two questions. Our analysis is divided into two parts: we first demonstrate that the confinement enhancement in NT configurations arises from the interdependent edge-core dynamics, and then we present the results of a system size scan. Crucially, we find that the relative turbulent transport reduction of NT over PT appears not to be contingent on machine dimensions or fluctuation scales and is moreover robust with respect to variations in plasma profiles. This insight underscores the fundamental nature of the NT confinement advantage and paves the way for its potential application in future fusion devices, regardless of their size.

Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices. In this work a new divertor model is proposed, which is processed by 3D-printing technology to accurately control the size of the internal capillary structure. At a steady-state heat load of 10 MW m−2, the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation. In order to evaluate the wicking ability of the capillary structure, the wicking process at 600 °C was simulated by FLUENT. The result was identical to that of the corresponding experiments. Within 1 s, liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface. During the wicking process, the average wicking mass rate of lithium should reach 0.062 g s−1, which could even supplement the evaporation requirement of liquid lithium under an environment > 950 °C. Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device (SCU-PSI), and the evolution of the vapor cloud during plasma irradiation was analyzed. It was found that the target temperature tends to plateau despite the gradually increased input current, indicating that the vapor shielding effect is gradually enhanced. The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure. These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid−solid combined target designs.

Implications of T loss in first wall armor and structural materials on T-self-sufficiency in future burning fusion devices

Future fusion reactors will have to breed enough tritium (T) to sustain continuous operation and to produce excess T to power up other fusion reactors. Therefore, T is a scarce resource that must not be lost inside the fusion power plants systems. The factor that describes the T production is the ‘tritium breeding ratio’ (TBR) which is the ratio of the breading rate in atoms per second to the burn rate in atoms per second. Its value is calculated from neutronics analyses of the breeding process in the blanket and coupled dynamics of the T processing plant. However, these calculations generally ignore the T transport and loss in the first wall by assuming essentially instantaneous recycling of the impinging T in-flux. In this paper the transport and retention of T in the main chamber first wall of a future EU-DEMO reactor is investigated based on the available material data and expected particle loads onto the wall. Two breeding blanket concepts are compared WCLL (water cooled lithium lead) and HCPB (helium cooled pebble bed) and the resulting wall-loss probabilities are compared with a simple balance model that describes the maximum allowable wall loss given a TBR to achieve T-self-sufficiency.

Long-term plasma exposure of ITER-like W/Cu monoblocks with pre-damaged surfaces in EAST experiments

In the ITER and future fusion devices, W/Cu monoblocks will be used as divertor target which are exposed to both steady state heat load and transient heat flux. Especially, the transient heat flux up to 10 GW m−2 during plasma disruption, is expected to induce the shallow surface damages, such as melting, and even boiling of W/Cu monoblocks. Thus, the performance of damaged W/Cu monoblocks under subsequent long-term plasma discharges is a key concern that needs to be verified and tested on existing tokamaks. Since 2022, a new type of main limiter composed of ITER-like W/Cu monoblocks has been installed and tested in EAST. The surface of W/Cu monoblocks of the limiter was damaged by the transient heat flux during the early stages of plasma construction. Subsequently, they were subjected to long-term plasma discharges over 2600 shots in normal plasma discharge conditions. This circumstance conveniently facilitates the discussion of the performance of W/Cu monoblocks with damaged surfaces especially a melting edge with hill structure under prolonged exposure to plasma. In general, the shallow damage resulting from transient heat flux on W/Cu monoblocks appears to have minimal impact on the heat exhaust capacity under steady-state heat loads, as indicated by both experimental monitoring and numerical simulation results. However, shallow melting, leading to a change in surface structure and the formation of hills, could theoretically increase local temperatures, creating potential hot spots. This phenomenon requires further validation through dedicated experiments. Moreover, the brittleness of the near-surface layer may give rise to brittle destructions, such as cracks and even dust particles, posing an additional concern. These findings yield unique qualitative conclusions that can be referenced for ITER and other fusion devices.

A full wave solver integrated with a Fokker–Planck code for optimizing ion heating with ICRF waves for the ITER deuterium–tritium plasma

Efficient ion heating is crucial for future fusion devices, and the only way to heat ions directly is ion cyclotron resonance heating. Reported here is a full wave solver integrated with a Fokker–Planck code for optimizing ion heating with ion cyclotron range of frequency waves for the International Thermonuclear Experimental Reactor deuterium–tritium plasma. Both the direct absorption of minority ions and the power transfer to bulk ions via collisions are considered, while also accounting for the edge effects on ion absorption near the core. The simulation results show that the appropriate scrape-off layer density profile and parallel wave number lead to enhanced edge coupling and broaden the absorption region with moderate absorption intensity of the minority ions, which is very important for ion heating. More power from the heated ions is transferred to bulk ions than to electrons through collisions in our simulation via optimization, and reducing the total RF power results in a significant increase of the absorbed fraction of bulk ions.