Divertor Design Research Articles

Front-Face RH Compatible Structure Design and Thermal Analysis for CFETR Divertor Dome

The China Fusion Engineering Test Reactor (CFETR) is a device developed to verify the engineering feasibility of a fusion reactor. For CFETR, the divertor is an important plasma-facing component, whose main function is to exclude impurities and remove plasma heat. In addition, the requirement for remote handling (RH) maintenance must be satisfied because of the level of radioactivity in the vacuum vessel after shutdown. The dome is an important component of the divertor, whose main function is to isolate impurity particles as well as to improve the ability of excluding particles. In the optional dome design, a hybrid divertor-blanket concept, a front-face RH compatible structure in plasma-facing units (PFUs), and a RH maintenance scheme for the main bolt are proposed. The vulnerable targets can be replaced directly and thus reduce the RH maintenance time. The dome needs to withstand the heat flux of 10 MW/m2 and nuclear heat in the condition of 1.5 GW of fusion power in the engineering design requirements. Because of the RH compatible structure, higher requirements are demanded for the design of the dome cooling system. In this study, the cooling system and the customized heat transfer structure of dome PFUs are designed to guarantee the maximum heat removal level. The steady-state thermal analysis shows that the cooling system fulfills the design requirements. The concept of the hybrid divertor-blanket and the front-face RH compatible structure for the divertor target have certain reference significance and value for the engineering design and RH maintenance research for the fusion reactor divertor in the future.

Recent progress of plasma exhaust concepts and divertor designs for tokamak DEMO reactors

The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

Comparing lithium vapor box designs in a high heat flux scenario using SOLPS-ITER

SOLPS calculations of lithium vapor box divertor designs on NSTX-U are presented. Predictive high power simulations (Pheat=10 MW, qpeakunmitigated∼65 MW/m2) are used to compare and contrast two divertor designs. Specifically a baffled “box” divertor, where a region of neutral density is allowed to build up, is compared to a more typical slot divertor geometry. It is found that significant differences in lithium containment lead to profoundly different viability of the two designs. These differences are seen to be due to far SOL flow patterns that change based on the presence of baffling as well differences in efficiency of the lithium evaporator. Outer-midplane (OMP) separatrix lithium content is found to be strongly detrimental to upstream temperature when nLi/ne>0.1 is reached. This regime of high upstream contamination is avoided via baffling. The reduction in upstream lithium allows access to low heat flux solutions below 5 MW/m2 with very little reduction to upstream temperature from the unmitigated, 65 MW/m2 solution. The slot is able to reach sub-10 MW/m2 heat fluxes though raising the evaporation rate much further reduces the upstream temperature, such that the range of stable evaporation rates with low heat flux to the target is small. Higher performance solutions (low heat flux and low upstream lithium content) are accessible by controlling recycling coefficients of deuterium on the walls above the box.

Improvement in the simulation tools for heat distribution predictions and control of baffle and middle divertor loads in Wendelstein 7-X

In the first divertor campaign in Wendelstein 7-X (W7-X), unexpected significant heat loads were observed at particular plasma-facing components (e.g. baffle tiles and middle divertor part) which were not designed to receive high heat flux. In a prior investigation, it was concluded that the previous diffusive field line tracing (DFLT) model used for divertor design in W7-X cannot reproduce these loads, due to the missing physics in simulating the heat transport in the shaded flux tubes. To tackle this issue, two new efficient codes (DFLT_rev and EMC3-Lite) are introduced and validated against various experimental heat distributions in different magnetic configurations. The new tungsten baffle tiles have been designed with these codes and mounted in the machine, aiming for mitigated heat loads in the upcoming campaign.

ELM sputter erosion modeling of a tungsten coated small angle slot divertor in DIII-D

We modeled plasma edge localized mode (ELM) sputter erosion for a Small Angle Slot divertor with a tungsten coated region (SAS-VW), designed for experiments in the DIII-D tokamak, and proposed for use in future advanced tokamaks. The simulations use a free-streaming, 1000 eV, C+6 and D+1 ELM impingement model, with SOLPS-ITER, ITMC-DYN, and REDEP/WBC code packages for background plasma, material response, and erosion/redeposition respectively. The results show ELM’ing plasma gross and net tungsten erosion fluxes of the mixed-material C/W surface peaking at the slot entrance region, and an order of magnitude higher than for non-ELMs. The per-pulse erosion, however, remains low, of order 0.5 nm, due to expected moderate ELM frequencies and duration in DIII-D. The ELMs result in a ∼25x higher peak sputtered W current leaving the divertor slot region, towards the core plasma, compared to the ELM-free plasma case. The time-integrated escape current, however, may not significantly affect core plasma high-Z contamination concerns, for a 1% ELM duty factor, but may be an issue for higher frequency ELMs. In general, the modeling results appear favorable for effective testing of the SAS-VW divertor in DIII-D, and extrapolation to innovative divertor designs in future ITER-like and DEMO fusion devices.

Design and thermal-hydraulic analysis for CFETR divertor Dome

The main function of China Fusion Engineering Test Reactor (CFETR) Divertor Dome is to isolate neutral particles and impurity particles as well as improve the ability of particle exclusion. For the CFETR Divertor Dome, a new integral hybrid divertor-blanket structure was proposed in previous study. The new Dome should also be compatible with the Remote handling (RH) requirements due to neutron activation of in-vessel components, thus special RH channels are designed on the front of the plasma-facing units (PFUs). Due to direct bombardment or sputtering of plasma, the Dome will suffer heat flux up to 10 MW/m2 and nuclear heat under 1.5 GW fusion power during steady-state operation, which puts forward critical and tricky requests for the design of cooling system. While for the RH channels, the situation is even direr due to its location and poor cooling conditions. In this paper, cooling system of the Dome is proposed, and cooling channels with special thermal transfer structure are designed in the PFUs to guarantee them with heat removal ability up to 10 MW/m2. Special PFUs and cooling system are designed to adapt to the RH, to shadow them from the direct bombardment of plasma and provide enough heat removal capability. Thermal-hydraulic analysis is carried out to verify the heat transfer ability of the Dome cooling system. The results will provide supports to the electromagnetic and structure analysis of Dome and promote the final design of CFETR water-cooled divertor.

Review of magnetic islands from the divertor perspective and a simplified heat transport model for the island divertor

Magnetic islands in toroidal confinement devices are reviewed from the viewpoint of their divertor potential. Divertor-relevant geometric parameters are derived analytically, and the relationships among them are revealed. We explain how the island geometry limits the target length and demonstrate the importance of an appropriate numerical tool to minimize the risk of thermal overload of plasma-facing components in the divertor design. The currently available three-dimensional (3D) models are briefly discussed, and their strengths and weaknesses are evaluated. The highlight will be the introduction of a new energy transport model recently developed within the framework of the EMC3 code (Feng et al 2004 Contrib. Plasma Phys. 44 57)—the so-called EMC3-Lite version—primarily for the design and optimization of 3D divertors involving thermal overload concerns. While still undergoing experimental validation with the current graphite divertor of W7-X, it is already being used to develop a subsequent tungsten divertor for W7-X.

Simulation of Liquid Lithium Divertor Geometry Using SOLPS-ITER

Plasma-facing component (PFC) geometries are evaluated for a Fusion Nuclear Science Facility (FNSF) design to find solutions that are compatible with flowing liquid lithium components while satisfying requirements from the perspective of plasma and neutral particle transport. Flowing liquid metal (LM) divertor systems offer important advantages due to continual regeneration of the surface material, but may require modifications from standard optimized tokamak divertor designs. Scrape Off Layer Plasma Simulator for the International Thermonuclear Experimental Reactor (SOLPS-ITER) simulations are used to compare standard vertical target, open, and balanced baffled geometries. It is found that the open geometry offers little control of neutral particles and has narrow operating windows where the divertor fluxes can be lowered to acceptable levels, while maintaining the required upstream density. The addition of neutral baffles allows greater control over the divertor conditions with less impact on the upstream density. The divertor neutral pressure in the baffled geometry is slightly higher than the open divertor, but not as large with a vertical target. A first-pass coupling of the SOLPS solution with sheath and implantation/erosion models to compute the lithium emission shows that the lithium is largely confined to the near-surface region due to the small ionization mean free path and large main ion flow toward the surface. The lithium source was varied by three orders of magnitude. At the highest rate, when the sourced lithium approaches the level of the main ion recycling flux, a high core lithium concentration (~10%) is observed with only moderate power dissipation to affect the divertor plasma.

E × B flow driven electron temperature bifurcation in a closed slot divertor with ion B × ∇B away from the X-point in the DIII-D tokamak

An electron temperature bifurcation is observed in the small angle slot divertor, which has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Experiments in the DIII-D tokamak and associated SOLPS-ITER modeling with full drifts find a strong interplay between drifts and divertor geometry on divertor dissipation. The coupling of divertor geometry and drift flows can strongly affect the path towards divertor detachment onset as the plasma density is raised. With the strike point on the inner slanted surface and ion B × ∇B away from the magnetic X-point, bifurcative transitions were observed with sharp decrease of T e towards detachment onset both experimentally and computationally. This differs from the situation for the open divertor where the T e cliff was only observed for ion B × ∇B towards the X-point. SOLPS-ITER modeling with full drifts demonstrates that the magnitude of the E × B drift flow is comparable with the main plasma flow. The reversal of both the poloidal and radial E × B flows near the strike point leads to rapid density accumulation right near the separatrix, which results in bifurcative step transition of divertor conditions with cold plasma across the entire divertor target plate. These results indicate that the interplay between geometry and drifts should be fully taken into account in future fusion reactor divertor designs.

Design of a multi-configurations divertor for the DTT facility

The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation. The new high field superconducting divertor tokamak test facility (DTT) [1] is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs. Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered. The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume. In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.

Divertor geometry modeling with the SOLPS-ITER code for reactor concepts with liquid metal divertors

• The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). • Two flat divertor configurations are investigated and we found that simplified open geometry presents challenges for neutral control and impurity retention, but balanced geometry with baffling appears promising. • Lithium (Li) is sourced into the balanced configuration and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. • At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the Private flux region (PFR) and divertor surfaces. • At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). Two divertor configurations are investigated to determine puff levels that provide robust operational windows that meet the FNSF design requirements on upstream density and divertor flux with a flat divertor for flowing liquid metal (LM). Neon (Ne) is introduced to control the heat flux to the divertors while Deuterium (D 2 ) puffing is performed to maintain sufficiently high upstream density. We found that a simplified open geometry presents challenges for neutral control and impurity retention, but a balanced geometry with baffling is found to have acceptable puff operational windows. Lithium (Li) is sourced into the balanced configuration uniformly along the divertor surfaces and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the private flux region (PFR) and divertor surfaces. At moderate sourcing level ( 1 × 10 23 < ϕ Li ≤ 1 × 10 24 / s ), the Li starts affecting the upstream and divertor plasmas, and Li radiation and electron cooling are almost linearly increasing with the Li sourcing level. At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. At high sourcing level, Li reaches above the x-point that can result in a substantial Li radiation in the core and sufficient dissipation to strongly affect the divertor and upstream plasmas.

The development of the MHD film flow numerical simulation platform

<p indent="0mm">In this paper, magnetic confinement fusion is introduced focusing on its fluid mechanics relevant to a liquid lithium divertor. The main research problem is the flow stability of a liquid metal film flow under the influence of a magnetic field. Therefore, a linear stability analysis numerical calculation platform and a direct numerical simulation platform are developed to understand the effect of magnetohydrodynamics on the flow stability of thin films. These platforms will be used to reveal the physical mechanism of a magnetic field on the film flow stability and to guide the design and implementation of a liquid lithium divertor.

Design of a Novel Variable Geometry Divertor for Tokamaks

The divertor is a key component of fusion reactors, allowing exhaust of gas, impurities, and helium ash to preserve plasma purity. The divertor geometry strongly affects plasma performance, and it is designed to be compatible with different plasma shapes in present-day fusion experiments. We present a novel concept for a variable geometry divertor, in which the divertor baffle tiles are reorientable by external actuation. Implementation of this concept in a medium-sized research tokamak would uniquely provide the flexibility to tailor divertor geometry to the plasma configuration and also enable study of the effect of divertor closure on plasma performance. To ensure compatibility with typical tokamak operations, the adjustable divertor must withstand the effects of significant mechanical and thermal stresses such as MW/m2-scale heat fluxes and large electromagnetic fields, e.g., disruption forces. The technological solutions for actuation mechanisms, cooling system, gas baffling and plasma-facing components are assessed. A functional reduced-scale model with movable outer divertor target baffle tiles is developed and the actuation mechanism is tested.

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

As an important component of tokamaks, the divertor is mainly responsible for extracting heat and helium ash, and the targets of the divertor need to withstand high heat flux of 10 MW m−2 for steady-state operation. In this study, we proposed a new strategy, using microchannel cooling technology to remove high heat load on the targets of the divertor. The results demonstrated that the microchannel-based W/Cu flat-type mock-up successfully withstood the thermal fatigue test of 1000 cycles at 10 MW m−2 with cooling water of 26 l min−1, 30 °C (inlet), 0.8 MPa (inlet), 15 s power on and 15 s dwell time; the maximum temperature on the heat-loaded surface (W surface) of the mock-up was 493 °C, which is much lower than the recrystallization temperature of W (1200 °C). Moreover, no occurrence of macrocrack and ‘hot spot’ at the W surface, as well as no detachment of W/Cu tiles were observed during the thermal fatigue testing. These results indicate that microchannel cooling technology is an efficient method for removing the heat load of the divertor at a low flow rate. The present study offers a promising solution to replace the monoblock design for the EAST divertor.

The physics of turbulence localised to the tokamak divertor volume

Fusion power plant designs based on magnetic confinement, such as the tokamak design, offer a promising route to sustainable fusion power but require robust exhaust solutions capable of tolerating intense heat and particle fluxes from the plasma at the core of the device. Turbulent plasma transport in the region where the interface between the plasma and the materials of the device is handled - called the divertor volume - is poorly understood, yet impacts several key factors ultimately affecting device performance. In this article a comprehensive study of the underlying physics of turbulence in the divertor volume is conducted using data collected in the final experimental campaign of the Mega Ampere Spherical Tokamak device, compared to high fidelity nonlinear simulations. The physics of the turbulence is shown to be strongly dependant on the geometry of the divertor volume - a potentially important result as the community looks to advanced divertor designs with complex geometry for future fusion power plants. These results lay the foundations of a first-principles physics basis for turbulent transport in the tokamak divertor, providing a critical step towards a predictive understanding of tokamak divertor plasma solutions.

Development and Application of SONIC Divertor Simulation Code to Power Exhaust Design of Japanese DEMO Divertor

An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evaluate the power exhaust scenarios with Psep = 230–290 MW. The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak qtarget ≤ 10 MWm−2 was determined in the low nesep of 2 − 3 × 1019 m−3 under severe conditions such as reducing radiation loss fraction, i.e., f*raddiv = (Pradsol + Praddiv)/Psep and diffusion coefficients (χ and D). The divertor geometry and reference parameters (f*raddiv ~ 0.8, χ = 1 m2s−1, D = 0.3 m2s−1) were consistent with the low nesep operation of the JA DEMO concepts. For either severe assumption of f*raddiv ~ 0.7 or χ and D to their half values, higher nesep operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral–neutral collision) under the DEMO relevant SOL and divertor condition are presented.

Benefits and Challenges of Advanced Divertor Configurations in DEMO

The divertor design and configuration define the power exhaust capabilities of DEMO, and act as a major design driver. They set a number of requirements on the tokamak layout, including port sizes, poloidal field coil positions, and size of toroidal field coils. It also requires a corresponding configuration of plasma-facing components (PFCs) and a remote handling scheme to be able to handle the cassettes and associated in-vessel components the configuration requires.There is a risk that the baseline ITER-like single-null (SN) divertor configuration cannot meet the PFC technology limits regarding power exhaust while achieving the target plasma performance requirements of DEMO or a future fusion power plant. Alternative magnetic configurations (AMCs) – for example, double-null, snowflake, and super-X – exist and potentially offer mitigation solutions to these risks and a route to achievable power handling in DEMO. However, these options impose significant changes on machine architecture, increase the machine complexity and affect remote handling and plasma physics and so an integrated approach must be taken to assessing the feasibility of these options.This paper describes work carried out to define a set of design limitations that any alternative divertor configuration for DEMO must consider, and assesses the feasibility of integration of a set of potential AMCs for DEMO.

SOLEDGE3X full vessel plasma simulations for computation of ITER first‐wall fluxes

AbstractITER Pre‐Fusion Power Operation 1 (PFPO‐1) phase low‐power plasmas are simulated with the SOLEDGE3X‐EIRENE code, a multi‐fluid edge plasma solver that can describe plasma conditions up to the machine's first wall, including a kinetic description of neutrals. Here, SOLEDGE3X is used in 2D mean‐field mode. A throughput scan is performed, reproducing a similar scan from the ITER SOLPS simulations database. The physics assumptions for these simulations are presented, in particular the plasma–neutral interaction model where molecule charge‐exchange and ion‐molecule elastic collisions have newly been added. Challenges encountered in such up‐to‐the‐wall simulations are discussed, along with remaining open questions. Results are presented with two foci: on targets, and on the rest of the first wall. On targets, conditions transition from attached to partially detached as expected, and toroidally symmetric heat fluxes (i.e. without 3D effects) remain below 5 MW/m2, within the divertor design limits. An increase in the parallel heat flux decay length λq with throughput is also observed. On the rest of the first wall, increasing puff rate does not only spread power deposited on the target but also increases heat load on the first wall. These fluxes remain however small (&lt;100 kW/m2).

New proposed ITER divertor design using carbon insert on tungsten to mitigate ELMs and secondary radiation effects on nearby components

Building a successful device for the magnetic fusion energy production is a great challenge. ITER is an international project of the tokamak based magnetic fusion design being developed for the demonstration of the feasibility of thermonuclear technologies for future realization of successful commercial fusion energy. A key obstacle to a successful magnetic fusion energy production is however, the performance during abnormal events including plasma disruptions and edge-localized modes (ELMs). A credible reactor design must tolerate at least a few of these transient events without serious consequences such as melting of the structure. This paper investigates and compares the performance of the current ITER tokamak design during two types of transient events, i.e., ELMs occurring at normal operation and disruptions during abnormal operation. We simulated the divertor components response using our integrated 3D HEIGHTS package. The simulations include self-consistent modeling of the interaction of the released core plasma particles with the initial solid divertor material, energy deposition processes, vaporization of divertor material, secondary plasma formation and MHD evolution, incident core particles collisions and scattering from this dense secondary plasma, photon radiation of secondary plasma, and the resulting heat loads on nearby components. Our simulations showed that using a small carbon insert around the strike point can significantly reduce the overall expected damage on the tungsten dome structure, reflector plates, and prevent tungsten vaporization and its potential core plasma contamination.

SOLPS-ITER simulations of high power exhaust for CFETR divertor with full drifts

One of the major challenges for the Gigawatt-class Chinese Fusion Engineering Testing Reactor (CFETR) is to efficiently handle huge power fluxes on plasma-facing components , especially the divertor targets. This work investigates the effects of two candidate radiation impurity species, argon (Ar) and neon (Ne), with two different divertor geometries (baseline and long leg divertor geometry) on the reduction of steady-state power load to divertor targets in CFETR by using the SOLPS-ITER code package with full drifts and kinetic description of neutrals. The modeling results show clearly that increasing the seeding rate of Ar or Ne with fixed fueling gas D2 injection rate reduces the target electron temperature and heat flux density for the baseline divertor geometry, which can be reduced further by higher D2 injection rate. With a high impurity seeding rate, partial detachment with steady-state power load at the divertor target below the engineering limit of 10 MW m−2 is demonstrated. In addition, the radiation efficiency for Ar is better than that for Ne. Increasing the divertor leg length reduces the electron temperature and heat load at the targets. This modeling, therefore, suggests that a long leg divertor design with Ar seeding impurity is appropriate to meet the CFETR divertor requirements.