N. Asakura

Chapter 4: Power and particle control

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

ELM energy and particle losses and their extrapolation to burning plasma experiments

Abstract Analysis of Type I ELMs from present experiments shows that ELM energy losses decrease with increasing pedestal plasma collisionality ( ν ∗ ped ) and/or increasing τ Front ∥ , where ( τ ∥ Front =2π Rq 95 / c s ,ped ) is the typical ion transport time from the pedestal to the divertor target. ν ∗ ped and τ Front ∥ are not the only parameters that affect the ELMs, also the edge magnetic shear influences the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. ‘Minimum’ Type I ELMs, with energy losses acceptable for ITER, where there is no change in the plasma temperature profile during the ELM, are observed for some conditions in JET and DIII-D. The duration of the divertor ELM power pulse is well correlated with τ Front ∥ and not with the duration of the ELM-associated MHD activity. Similarly, the time scale of ELM particle fluxes is also determined by τ Front ∥ . The extrapolation of present experimental results to ITER is summarised.

Edge localized mode physics and operational aspects in tokamaks

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs.High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high βp are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and βp.The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in terms of their relevance for ITER. Successful plasma edge control, at high triangularity (~0.5) and high density (~0.7nGR), in ITB scenarios in JET is reported.Active control of ELMs by edge current, pellet injection, impurities and external magnetic perturbations creating an ergodic zone localized at the separatrix are discussed for present-day experiments and from the perspective of future reactors.

Investigations of impurity seeding and radiation control for long-pulse and high-density H-mode plasmas in JT-60U

Rseduction of heat loading appropriate for the plasma facing components such as the divertor is crucial for a fusion reactor. Power handling by large radiative power loss has been studied in long pulse ELMy H-mode discharges on JT-60U (τd = 30–35 s). Case 1 is argon (Ar) seeding into standard ELMy H-mode plasmas, where large radiation loss in the confined region of the main plasma caused a change in ELM characteristics from Type-I to Type-III. Case 2 is a combination of Ar and nitrogen (Ne) gas seeding into Type-I ELMy H-mode plasmas with an internal transport barrier (ITB). For case 1, large radiation loss both from the main plasma and from the divertor was produced, and operation of Type-III ELMs was preferable to a reduction in ELM energy loss fraction (WELM/Wdia) to 0.15%. Both transient and steady-state heat loadings were reduced. Relatively good energy confinement (HH98y2 = 0.87 − 0.75) with large frad (Prad/Pabs > 0.8) and divertor plasma detachment was sustained continuously for 13.5 s. For case 2, with reduced Ar seeding to the main plasma and increased divertor radiation with Ne seeding, the ELMy H-mode plasma with an ITB had better energy confinement (HH98y2 = 0.95 − 0.8), which was sustained continuously for 12 s. The radiated power was increased primarily in the divertor ( ), which was produced both by seeded Ne ions and by carbon influx due to transient (ELM) and steady-state heat loadings in the attached divertor. Reduction in the heat loading was not enough, thus enhancement of the radiated power in the divertor will be necessary for the formation of the divertor detachment.

Initial Boronization of JT-60U Tokamak Using Decaborane

A decaborane-based boronization system has been installed in the JT-60U tokamak in order to reduce the influx of impurities during plasma discharges. Boronization has been performed under a glow discharge using a helium-decaborane gas mixture. The properties of the boron films deposited through boronization and the effects of boronization on the tokamak discharges were investigated. It was found that the deposition of a boron layer with high purity was achieved with few impurities other than hydrogen through boronization, and that the present boronization deposited toroidally nonuniform boron film. It was also found that the decaborane-based boronization resulted in good plasma performance similar to that of conventional boronization.

Recent progress in understanding the behavior of dust in fusion devices

It has been known for a long time that microscopic dust appears in plasmas in fusion devices. Recently it was shown that dust can be responsible for the termination of long- discharges. Also, in ITER-scale experiments dust can pose safety problems related to its chemical activity, tritium retention and radioactive content. In particular, the presence of dust in the vacuum chamber of ITER is one of the main concerns of the ITER licensing process. Here we review recent progress in the understanding of different experimental and theoretical aspects of the physics of dust dynamics and transport in fusion plasmas and discuss the remaining issues.

Measurement of the chemical sputtering yields of CH4/CD4 and C2Hx/C2Dx at the carbon divertor plates of JT-60U

The chemical sputtering yields of CH4/CD4 and C2Hx/C2Dx have been measured at the divertor plates of JT-60U. Spectroscopic measurements for CH/CD and C2 spectral bands are applied to estimate the CH4/CD4 and the C2Hx/C2Dx flux. At the surface temperatures of 380, 440 and 560 K, the CH4 yield is, respectively, ~0.8%, 1-2% and 2-3%, the C2Hx yield 1-2%, 3-4% and 4-5%, and the total sputtering yield by hydrogen ions 3-4%, ~8% and ~10%. With increasing ion flux to the divertor plates (Γion), the sputtering yields (Y) decrease, i.e. Y∝Γion(-0.05 to -0.40). With increasing electron temperature (Te), the sputtering yields increase, i.e. Y∝Te0.5. It is concluded from the result of regression analysis of Y∝Te0.5 that the negative dependence of the yields on the ion flux is attributed to the incident ion energies to the carbon plates. The ratio of the sputtering yields by deuterium ions and hydrogen ions is estimated to be ≥1.5 based on the ion flux measurement by Hα/Dα intensity. The C2Hx/C2Dx sputtering yield accounts for ~80% of the total number of sputtered carbon atoms.

Role of divertor geometry on detachment and core plasma performance in JT60U

Experimental results related to the divertor geometry such as divertor plasma detachment, neutral transport and plasma energy confinement, were compared in the open and W-shaped divertors. The ion flux near the outer strike point was larger than in the open divertor, and the electron temperature at the target, T e div , was reduced. Divertor detachment and x-point MARFEs occurred at n e 10-20% lower than that for the open divertor. Although the leakage of neutrals from the divertor to the main chamber decreased, a neutral source in the main chamber due to an interaction of the outer scrape-off layer (SOL) plasma to the baffle plates became dominant above the baffle. Degradation in the enhancement factor of the energy confinement was observed similarly in the open and W-shaped divertors. The neutral density inside the separatrix was estimated to be a factor of 2-3 smaller, which did not affect the energy confinement.

Impurity and particle recycling reduction by boronization in JT-60U

Abstract In JT-60U boronization using decaborane was carried out. Boronization reduced the oxygen impurity in OH discharges and shortened the wall conditioning after the vacuum vessel vent and consequently enabled JT-60U to produce clean plasmas easily except for NB heated plasmas. After boronization, particle recycling was reduced drastically in OH and NB discharges. High confinement plasmas were obtained including high β p mode and H-mode discharges. In the latest boronization part of divertor plates were replaced with B 4 C coated tiles with a B 4 C thickness ∼ 300 μm. After introducing B 4 C divertor tiles, an explosive generation of boron particles from the tiles was observed. By the combined effects of boronization with decaborane and boron generation from B 4 C tiles, oxygen impurity was so low that oxygen line signals were reduced to noise levels after the latest boronization. On the other hand, boron burst from the B 4 C tiles restricted the operation of JT-60U.