The recent development of novel refractory materials for fusion applications with better ductile properties has led, in the recent years, to rethink the helium cooled divertor design under development at KIT. In particular, the availability of pipes and plates made of tungsten laminates showing very good mechanical properties over an extended temperature range has triggered a new search for...
Tungsten (W) is currently considered the preferred plasma-facing material (PFM) for future magnetic confinement thermonuclear fusion reactors. This is mainly due to the fact that W exhibits a high threshold energy for sputtering by hydrogen isotopes as well as a low retention of tritium within the material. From an engineering point of view, however, W is a difficult metal to work with, as it...
Interaction between the tungsten, used as plasma facing material, and the D/He plasma is still under investigation, in particular for the long duration pulses (> 30s) and high heat fluxes (>10MW/m²). One of the main goal of WEST experiment is to study the behavior of an ITER-like divertor in a tokamak environment, and in particular its impact on the operation, for this type of...
Total flux expansion, a divertor magnetic topology design choice embodied in the Super-X divertor, is predicted through simple analytic models [1] and SOLPS calculations [3] to reduce the plasma and impurity density detachment thresholds as the outer divertor target strike point position, Rt, is increased. Since the total magnetic field, |B| ~ 1/R, |B| at the target is lowered as Rt is...
A major challenge facing the design and operation of future high-power steady-state fusion devices is to develop a robust boundary solution with an order-of-magnitude increase in power handling capability relative to present experience, while having acceptable erosion at the surface of the plasma facing components (PFCs) to ensure an adequate reactor lifetime. Recently, a small angle slot...
Active handling of excessively high heat load and tungsten sputtering on divertor targets is of critical challenge for EAST and future fusion devices like ITER and CFETR. It is acknowledged by the fusion community that divertor detachment is the most promising means for steady state plasma-wall interaction control.
Significant progresses on the active feedback control of H-mode detachment...
Future fusion reactors require a safe, steady state divertor operation. In the detached regime, the power and particle fluxes to the divertor targets are sufficiently reduced to meet the material limits. In H-mode operation at the full-tungsten ASDEX Upgrade tokamak (AUG), this is achieved by injection of significant amounts of nitrogen into the divertor volume. This increases dominantly the...
Plasma-surface interactions (PSI) span diverse physical processes as well as many decades of time and length scales (ps–s and Å–m). Correspondingly, comprehensive modeling of PSI must accurately target each scale and mechanism. Here, we present an integrated model designed to capture the multi-physics nature of interactions between the edge plasma and the divertor surfaces in a fusion tokamak....
This work was carried out within the framework of EUROfusion/PPPT SAE (Safety and Environment) project. Activity and decay heat values were calculated for the DEMOnstration power plant (DEMO) 2015 baseline model divertor. Two irradiation scenarios were considered lasting for 5.2 and 14.8 calendar years respectively. Each irradiation scenario describes continuous irradiation with exception of...
Operating a tokamak fusion reactor in a highly dissipative detached divertor regime, whilst maintaining sufficient core confinement, remains a major challenge. It is likely that advanced divertor solutions will be required. Proof of principle experiments of such ideas, as well as the underlying physics processes, are being explored on the TCV tokamak. Previous studies in L-mode revealed a...
The divertor is one of the key components of the EU-DEMO reactor. The development of a reliable solution for the power and impurity particle exhaust is recognized as a major challenge towards the realization of DEMO. The pre-conceptual design activities for the EU-DEMO divertor are carried on considering two project areas: the ‘Target development’, focusing on the design of the vertical...
Divertor with tungsten act as plasma facing material is expected apply for future fusion reactors. How to exhaust high heat load deposit on tungsten is key issue. Tungsten monoblock structure which applied for ITER divertor is a very good solution. Is there any other tungsten bonding structure which can handling 10MW/m2 heat load or even more?
A kind of flat heat sink with special cooling...
Plasma exhaust is a crucial issue for future fusion reactors. The high-power level across the separatrix needed to ensure H-mode operation and the narrow Scrape-off Layer (SOL) width make the task of staying within acceptable target heat loads extremely challenging, probably necessitating operation in a detached regime. In the past few years, significant efforts have been devoted towards the...
A numerical study is conducted to explore the thermal efficiency of cooling streams in geometries relevant for fusion reactor high heat flux components. A tile-type structure is considered employing one or more cooling channels within the heatsink, based on recent investigations showing the benefits of this approach and advances in additive manufacturing. Various flow configurations are...
A gas baffle is being installed in the vessel of the tokamak à configuration variable (TCV) [1], in order to improve the closure of the divertor region. This upgrade has been envisaged, along with a foreseen increase in the available input power, in order to facilitate the access to detached divertor regimes at lower plasma collisionality, namely in more ITER-relevant conditions. It is...
A spontaneous break of the up-down symmetry of the divertor plasma parameters in a symmetric transport model (symmetric double-null divertor configuration and boundary conditions, as well as the absence of drifts) in the presence of impurity seeding was found in computational modelling [1], [2]. The effect was attributed to the radiation-condensation instability (RCI) that amplifies the...
In the future fusion reactor, huge power has to be exhausted through the divertor due to the high fusion power. Therefore, it is critical to find an appropriate way to reduce the heat load onto divertor target, which has an engineering limit of 10 MW/m2. In snowflake configuration [1], second order null is introduced to increase the flux expansion as well as connection length, which benefits...
The continuing rapid evolution of a number of advanced technologies being strongly pursued for major non-fusion applications, is potentially transformative for the divertor physics performance requirements of reactors:
Advanced Manufacture, e.g. 3D printing, holds promise to increase the power handling capability of solid divertor targets significantly above the present limit for total power...
In reactor-size fusion devices, such as ITER and DEMO, are expected the critical loads on the divertor plates both during steady state and at transient events (disruption, VDE, ELMs, runway electrons). High heat plasma load leads to enhanced erosion and destruction of material surface accompanied by enhanced absorption of tritium in erosion products.
The paper introduce the experimental dates...
The divertor is a fundamental component of fusion power plants, being primarily responsible for power exhaust and impurity removal via guided plasma exhaust. Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m2), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its...
Resonant magnetic perturbation (RMP) fields applied for control of edge-localized modes (ELMs) break the axisymmetry of the plasma boundary in tokamaks. With RMP fields applied, a striation of the divertor target heat and particle flux pattern is detected which proves existences of helical magnetic fingers reaching from the X-point outward to the divertor target. A threedimensional plasma...
For the reactor-scale fusion devices such as ITER or DEMO, control of the divertor target power loading, both in steady state and during ELMs, is particularly challenging with regard to tungsten target lifetime. It should be preferably reduced below a certain value so that the divertor target cooling capability ensures a planned long-term replacement period of the targets. It is widely...
The ITER divertor has been designed for axisymmetric configurations, yet symmetry breaking resonant magnetic perturbations (RMPs) will be applied for control of edge localized modes (ELMs). Recently,the numerical capability to investigate the predicted detached divertor scenario at ITER with such 3-D deformations has been made available after stabilization of the iterative framework [H....
The island divertor (ID) concept investigated experimentally in the Wendelstein 7-AS (W7-AS) and the superconducting Wendelstein 7–X (W7-X) stellarator devices have so far proven to be extremely successful, and shown a favourable tendency towards improved detachment performance from W7-AS to W7-X. Stable detachment, which could be achieved only partially in W7-AS and required additional...
This paper discusses advantage and disadvantage of the LHD heliotron divertor in terms of divertor functions, such as neutral compression, impurity transport, and divertor heat load control.
The divertor plasma density in heliotron divertor stays at low values, ~1x10^19 m^-3, which never exceeds upstream density, i.e. at the stochastic layer, where ne can reach up to ~10^20 m^-3. This is...
The controlled particle and heat exhaust is one of the most challenging aspects towards the realisation of a commercial fusion power plant. However, despite this importance, it is difficult to extrapolate the expected divertor heat load for a future fusion power plant. In the tokamak community, credible models for the divertor heat load in DEMO have long been missing and still largely rely on...
A fast-flowing liquid metal (e.g. Lithium, Tin) divertor (FFLMD) is an attractive option that can take all (or almost all) the heat flux coming to the PFCs. A generic fusion reactor divertor with “fast” flow generally requires a ~1-20 m/s speed with approximately mm to cm thickness. Balancing the heat flow into the divertor and carrying capacity of the liquid metal (LM) flow is the main...
The crucial stepping stone between ITER and a fusion power plant is generally foreseen as a demonstration power plant (DEMO). The European approach foresees only a modest upscaling in dimensions from ITER but due to the large increase in fusion power and subsequently strongly increased power crossing the separatrix [1] this implies increased challenges for power exhaust. As a risk mitigation...
The divertor for a practical fusion power producing facility very likely must dissipate the intense heat flux emerging from the plasma core volumetrically, rather than allowing it to strike a material surface directly. We have proposed [1, 2] that a dense cloud of lithium vapor be contained in the divertor region by local evaporation from, and condensation onto, capillary porous structures...
Addressing the effect of E×B on closure diveror detachment onset by SOLPS
Hailong Du, Guoyao Zheng, Jiaxian Li
Southwestern Institute of Physics, PO Box 432, Chengdu 610041, People’s Republic of China
Email: duhl@swip.ac.cn
Abstract
The closed divertor (such as small angle slot-SAS, C-Mod)[1] can well trap neutral (D, D2) and carbon impurity from erosion particles in...
Liquid metals have the potential to mitigate several issues inherent to solid divertor targets, e.g., problems arising from erosion, embrittlement due to neutron irradiation and crack formation under fast transient loads. As possible choice for such a liquid metal tin ($T_{melt} = 505$ K) was identified, which promises low physical sputtering yields and a large operational temperature range...
High power magnetically confined fusion devices have very high heat and particle loads on the plasma facing components. Liquid metals (LM) mock-ups were proposed as alternative of full tungsten divertor for DEMO. Extrapolation of the disruptions/ELMs erosion effects obtained at the present-day tokamaks to the transient peak loads of next step fusion devices (ITER and DEMO) remains uncertain....
Helical systems inherently have a suitable feature as a future fusion power plant in terms of steady state operation because of no need of the plasma current drive. Among several configurations, conceptual design study of the LHD-type helical fusion reactor has been conducted and the design of the commercial scale power plant FFHR-d1, which can be operated with a fusion power of 3 GW, has been...
The kinetic trajectory simulation method has been employed to study the plasma-wall interaction in the magnetized plasma with two species of positive ions exposed to the tungsten (W)-surface. The multi-component plasma interacts with W-surface through non-neutral plasma sheath formed near the Plasma Facing Materials (PFMs). It is found that the ion velocity distribution functions have a...
Liquid metal (LM) plasma facing components (PFC) are considered an attractive design choice for fusion devices including pilot plants. Several liquid metal concepts for the divertor region are currently under development. Lithium or lithium eutectics have a high affinity for tritium and deuterium at low operating temperatures, and provide a low-recycling boundary condition for the core plasma,...
This paper reports divertor heat load patterns and plasma responses observed in the Ne, N, and Kr seeding experiments in LHD.
The previous study showed that the edge stochastic magnetic field layer in LHD, where Te changes from ~30eV at the divertor legs to 300-500eV at the LCFS, provides the main radiation contribution rather than the divertor legs. Understanding and control of impurity...
Turbulent transport has two critical impacts on the operational domain of tokamak reactors: it sets the core confinement performances through limitation of kinetic gradients from the very centre of the confined plasma to the magnetic separatrix, and it sets the condition of power exhaust by the tokamak wall, through the size of the heat flux wetted area. Experiments across a variety of...
Neutral particles play a key role in addressing the power exhaust challenge in magnetically confined fusion devices. The presence of neutral gas, together with impurities, allows reaching tolerable plasma temperature in front of the divertor targets, and helps reducing peak heat fluxes through power spreading. In fact, the neutral gas pressure in the divertor is often considered to be a key...
For reactor-scale tokamaks, the core plasma operational scenario imposes a number of boundary conditions on the divertor and SOL plasma. The most critical of these, upstream power density flowing into the divertor, and upstream separatrix electron and impurity density exert high leverage over divertor designs, and may even determine the need for advanced divertor designs to safely exhaust the...
Fueling a future fusion reactor and the effects of fueling upon the pedestal structure is an open physics question and the U.S fusion research program has initiated a multi-institutional effort to develop a physics basis to address this question. In contrast to existing devices, future reactor-scale tokamaks the edge pedestal will be opaque to neutrals and no longer be fueled primarily by edge...
Power exhaust research at DIII-D is addressing the need for fusion reactors to integrate high $\beta_N$, high confinement plasmas with a radiating mantle and divertor that are compatible with stringent requirements on fuel dilution, high confinement mode power threshold, plasma stability, and wall heat flux limits. This research program encompasses studies from diagnostic optimized divertor...
While prediction for the divertor heat-flux width in the wide range of the poloidal magnetic field in attached NSTX, DIII-D, NSTX and JET plasmas matches the Eich-14 scaling formula within the regression error bar, the XGC prediction for the full-current ITER showed over six times wider heat-flux width than the Eich value. There were questions from the community if this difference in the...
The BOUT++ code has been used to simulate edge plasma electromagnetic (EM) turbulence and transport, and to study the role of EM turbulence in setting the scrape-off layer (SOL) heat flux width λ_q. More than a dozen tokamak discharges from C-Mod, DIII-D, EAST, ITER and CFETR have been simulated with encouraging success. The parallel electron heat fluxes onto the target from the BOUT++...
The current baseline EU-DEMO, as designed by the EUROfusion Power Plant Physics & Technology Department (PPPT), considers an ITER-like LSN divertor for particle and power exhaust. Various modelling activities have been undertaken in the past years to assess the performance of this key machine component, both concerning plasma physics and engineering design. Goal of the present work is to...
Recent analysis using the GA systems code (GASC) has provided new insights into the power exhaust requirements for future fusion power systems. This analysis was enabled by improvements in the underlying models for power exhaust, magnet technology limits, bootstrap current, and costing as well an improved optimization algorithm capable of identifying optimum solutions for a range of...
$~~$ Power exhaust scenario for the feasible DEMO plasmas and the divertor design have been studied with a high priority in the steady-state Japanese (JA) DEMO with the fusion power of 1.5 GW-level and the major radius of 8 m-class. The power exhaust concept requires large power handling in the SOL and divertor, i.e. $P_{sep}$~250 MW, and $P_{sep}/R_{p}$~30 $\rm MWm^{-1}$ corresponds to 1.8...
The uncertainties surrounding the physics of plasma exhaust and its centrality in reactor design require a thorough evaluation of promising exhaust configurations, so EUROfusion established a project to assess alternative divertors for reactor relevant devices and DEMO in particular. An alternative here is any divertor solution that cannot be qualified by ITER and it includes, but is not...
Abstract
The experimental and modeling have shown that the advanced snowflake divertor can well mitigate heat flux loading onto target surfaces due to the smaller perpendicular incident angle and larger magnetic expansion compared with conventional divertor [1,2]. But, the large magnetic expansion of snowflake may lead to a serious problem of particle exhaust, which results in the core...
Divertor is one of the key components in Tokamak. The control of heat flux and erosion of the divertor target is one of the grand challenges facing the design and operation of next-step high-power steady-state fusion. It is essential to efficiently dissipate power in the divertor to ensure the maximum steady-state power load at the divertor target below 10 ~ 15 $MW/m^2$. In addition, adequate...
The design of a DEMO divertor is an important task which defines the reflux of fuel and helium neutral particles to the plasma and finally determines the particle exhaust and pumping efficiency. For the conventional divertor, optimization of the dome height and its effect on neutral compression, position of the pumping ports as well as the effect of neutral gas screening by electrons in the...
The main objective for JT-60SA is to study magnetically confined plasma in near-fusion scenarios in support of ITER and DEMO. One of the major open issues is to demonstrate divertor heat and particles handling in ITER-like plasma conditions. One of the options for JT-60SA divertor in the Integrated Research phase II is to use Tungsten as plasma facing component[1], while in the initial...
The development of a reliable solution for the power and particle exhaust in a reactor is recognized as a major challenge towards the realization of DEMO [1]. Alternative magnetic configurations such as Double Null, Snowflake, X and Super-X divertors are considered as a promising solution to reduce the heat load on the divertor targets even if their scalability on a DEMO size device is a...
The divertor configuration defines the power exhaust capabilities of DEMO as one of the major key design parameters and sets a number of requirements on the tokamak layout, including port sizes, PF coil positions, and size of TF coils. It also requires a corresponding configuration of plasma-facing components and a remote handling scheme to be able to handle the cassettes and associated...
In a next step fusion device like ITER or DEMO, the unmitigated power loads at the divertor targets will considerably exceed the allowed material limits which are foreseen to be in the order of $5-10\,\mathrm{MWm}^{-2}$. Therefore, to prevent severe damage of plasma facing components and erosion of target material, a significant amount of power has to be exhausted via impurity radiation. For...
Power exhaust scenario for the feasible DEMO plasmas and the divertor design have been studied with a high priority in the steady-state Japanese (JA) DEMO with the fusion power of 1.5 GW-level and the major radius of 8 m-class. The power exhaust concept requires large power handling in the SOL and divertor, i.e. Psep~250 MW, and Psep/Rp~30 MWm-1 corresponds to 1.8 times larger than ITER. Long...
Operating at 12 tesla on axis with a plasma current of 7.5 MA and total fusion power of 100 MW, SPARC [1] is projected to have a power exhaust heat flux width of 0.2 mm with an unmitigated parallel heat flux of up to 30 GW m-2 entering the divertor. While recent UEDGE modelling of other high-field tokamaks designs, ADX and ARC, indicates that long-leg divertors can dramatically improve...
The fusion power of China Fusion Engineering Test Reactor (CFETR) [1] is proposed to achieve the level of gigawatt, which implies the critical issue of power exhaust by divertor. Impurity radiation is an effective way to reduce the heat load onto the divertor target. For CFETR, to reduce the tritium retention and increase the lifetime of plasma-facing components, full-tungsten wall would be...
Systems codes, such as PROCESS, model all systems of a power plant to investigate large numbers of design points. They are used for scoping studies and to identify areas of feasible design points.
Multi-dimensional modelling of the plasma Scrape-Off-Layer (SOL), divertor and seeded impurities is too computationally intensive to be incorporated directly into a systems code. Divertor...
The EUROfusion roadmap [1] has recognized the exhaust of large heat loads as one of the most critical issue to solve for the generation of electrical power with a Demonstration Fusion Power Plant (DEMO) by 2050. This condition is particularly challenging in the material facing the plasma in the divertor, where detached conditions must be guaranteed for safe operations of the machine. According...
On the eve of component procurement and with a substantially updated Research Plan [1] describing the pathway to achievement of inductive and steady state burning plasma operation on ITER, the present physics basis for the first ITER tungsten (W) divertor is outlined, focusing on the main design and operational driver: steady state and transient target power fluxes [2]. With the dimensions...
Since 2017, significant efforts have been made by Chinese community for the physics and engineering design of Chinese Fusion Engineering Testing Reactor (CFETR) [1], which is proposed to bridge the gap between ITER and DEMO. One of the key challenges is that the divertor solution for CFETR must meet requirements beyond that of ITER. For the standard CFETR operation with the fusion power up to...
Predictions for the scrape-off layer (SOL) in future fusion devices based on empirical scalings imply extremely large parallel heat flux, q|| ~10 GW/m2, which is exacerbated for high-field concepts that may enable a Compact Pilot Plant (CPP) as recommended by a recent US strategic planning assessment. Here we discuss the framework of a proposed ~10 year program to more firmly establish the...
Dealing with power exhaust is one of the most difficulty tasks in foreseen fusion reactor based on magnetic confinement. It is commonly recognized that the standard Single Null Divertor (SND) configuration can face some difficulties in providing a scalable solution based on H-mode tokamak operation compatible with present solid divertor target technological solutions. To provide a safer...
The divertor target is the most intense plasma-surface interaction area in the tokamak. To keep the lifetime of the device and maintain long pulse discharges, the power load and particle removal control become to be the critical issues. The lower graphite divertor of EAST tokamak is the main limitation to the achievement of further high-power long-pulse discharges [1]. To solve this problem,...
