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Since tungsten (W) was considered as the most promising plasma facing materials (PFMs) in fusion reactors, there has been extensive research on the physical performance of W-PFMs. It is found that under the extreme conditions in a fusion reactor, W-PFMs should be in a nonequilibrium state of high electronic temperature and low ionic temperature. This leads to the possibility of non-thermal phase transitions, where the crystal structure of the tungsten material may change from body-centered cubic (bcc) phase to hexagonal close-packed (hcp) phase or face-centered cubic (fcc) phase. Consequently, it is necessary to investigate the relevant physical properties of hcp-W and fcc-W under the electron-excited state. In this work, the fundamental physical properties, including atomic structures, electronic structures, elastic constants, and vacancy formation energies, of bcc-W, hcp-W and fcc-W, were theoretically calculated at various electronic temperatures. The mechanical stability of these three phases was also systematically analyzed under varying electronic temperatures. The results of this research are expected to provide a certain guidance in the optimization of W-PFMs in future fusion reactors.

The necessity of a neutron source for fusion materials research was identified already in the 70s. Though neutrons induced degradation present similarities on a mechanistic approach, thresholds energies for crucial transmutations are typically above fission neutrons spectrum. The generation of He via 56Fe (n,α) 53Cr in future fusion reactors with around 12 appm/dpa will lead to swelling and structural materials embrittlement. Existing neutron sources, namely fission reactors or spallation sources lead to different degradation, attempts for extrapolation are unsuccessful given the absence of experimental observations in the operational ranges of a fusion reactor. Neutrons with a broad peak at 14 MeV can be generated with Li(d,xn) reactions; the technological efforts that started with FMIT in the early 80s have finally matured with the success of IFMIF/EVEDA under the Broader Approach Agreement. The status today of five technological challenges, perceived in the past as most critical, are addressed. These are: 1. the feasibility of IFMIF accelerators, 2. the long term stability of lithium flow at IFMIF nominal conditions, 3. the potential instabilities in the lithium screen induced by the 2 × 5 MW impacting deuteron beam, 4. the uniformity of temperature in the specimens during irradiation, and 5. the validity of data provided with small specimens. Other ideas for fusion material testing have been considered, but they possibly are either not technologically feasible if fixed targets are considered or would require the results of a Li(d,xn) facility to be reliably designed. In addition, today we know beyond reasonable doubt that the cost of IFMIF, consistently estimated throughout decades, is marginal compared with the cost of a fusion reactor. The less ambitious DEMO reactor performance being considered correlates with a lower need of fusion neutrons flux; thus IFMIF with its two accelerators is possibly not needed since with only one accelerator as the European DONES or the Japanese A-FNS propose, the present needs > 10 dpa/fpy would be fulfilled. World fusion roadmaps stipulate a fusion relevant neutron source by the middle of next decade, the success of IFMIF/EVEDA phase is materializing this four decades old dream.

For the development of the tritium monitoring system in ITER the hydrogen isotope release by Laser-Induced Desorption (LID) from Be layers is studied to determine the laser parameters for a high desorption efficiency while minimising dust production and surface modifications is also pursued. Be layers of 1 µm thickness with 25–30 at% D and 3 × 1022 D/m2 comparable to JET-ILW areal concentrations [1] have been produced by High Power Impulse Magnetron Sputtering (HiPIMS) on ITER grade W. Laser pulses of 1, 5 and 10 ms duration heat the layer in vacuum in the Fuel REtention DIagnostic Setup (FREDIS) and release the retained D thermally. By mass spectrometry in FREDIS and subsequent Nuclear Reaction Analysis (NRA) inside the laser spot the desorbed and remaining D is quantified. While a pulse duration of 1 ms cannot fully desorb the deuterium, it is found that a single 5 or 10 ms laser pulse with an absorbed energy density of ca. 1.5 MJ/m2 corresponding to a heat flux factor around 20 MW√s/m2 leads to nearly complete desorption of the retained D. This encourages the development of a useful tritium monitoring system, although the present layers produce some dust due to local delamination of the layer on at least 11% of the heated surface (at 1.4 MJ/m2 absorbed energy within 5 ms) and lead to unavoidable crack formation. Keywords: Fuel retention, Beryllium, Tritium monitoring, Laser, Desorption, FREDIS

Oxide dispersion strengthened tungsten (ODS-W) is a potential candidate for plasma-facing materials (PFMs) in future fusion reactors. In this work, deuterium (D) retention and surface blistering in W-1 wt% La2O3 (W-La2O3) have been investigated after exposure to low-energy (40 eV) D plasma with various exposure temperatures (400–600 K) and fluences (3.6 × 1024–1.4 × 1025 D/m2). Surface blistering and D retention exhibit a strong dependence on the exposure temperature and fluence. The most pronounced effect is found at 500 K. The blister-induced defects including dislocations and vacancies are considered to dominate the D retention. At 400 K and 600 K, the D retained in W-La2O3 is governed by unique intrinsic defects including interfaces, micro-pores, and unoxidized La particles. Regarding the exposure fluence, as expected, surface blistering and D retention are positively correlated with it, in which two dominant stages of nucleation and growth for blistering are identified from the changes in area density and size of blisters. Based on the results obtained from W-La2O3, comparisons with W are performed with the exposure condition (500 K, 1.4 × 1025 D/m2) where the blistering and D retention is most pronounced. Although the area density of blisters is similar between the two materials, the average size of blisters is larger in W-La2O3. Notably, an additional high-temperature D desorption shoulder appears in the release spectra of W-La2O3, which is probably due to the particular defects such as interfaces, micro-pores and La particles, and finally resulting in a higher D retention in W-La2O3 than that in W.

The behavior of ITER-like W/Cu plasma facing components under complex conditions in current tokamaks is one of the main concerns for ITER. EAST has installed a full upper W divertor with the ITER-like W/Cu monoblocks as targets since 2014. A melting failure of CuCrZr cooling tube of W/Cu monoblocks has occurred during the plasma campaigns in 2019. Due to the loss of cooling water, the leading edge-induced thermal loading can lead to the melting of both W armor and CuCrZr cooling tube, which has been confirmed by thermal simulation and analysis. With movement and migration of the melted Cu through gaps, the structure and function of the CuCrZr cooling tube for W/Cu monoblocks was severely destroyed. As a result, those failed W/Cu monoblocks had to be replaced. Such case in EAST just simulates the extreme condition of accidental loss of coolant in future devices. The melting failure of CuCrZr cooling tube was caused by operation without coolant water at largely misaligned monoblocks, which is a key lesson referenced to other tokamaks which adopt or plan to apply such type of W/Cu monoblocks.

Plasma facing materials (PFMs) are subjected to long-duration high-energy particle streams and radiation in tokamak devices. The PFMs of EAST have been upgraded several times and Titanium-Zirconium-Molybdenum (TZM) tiles were installed into EAST as its first wall since 2011. However, with the gradually increasing of plasma parameters, several unexpected TZM melting phenomena were found at the high field side by post mortem inspection after each EAST plasma experimental campaign since 2017. The resolidified melted surface is general in wave shape with unobvious motion of melting layer. Three different grain shapes, i.e., columnar grain, isometric crystal and original rolled crystal from surface to deep region are found by means of metallurgical analysis, in which the superficial layer columnar grain is very thin with a thickness of 100 ∼ 200 μm and the thickness of intermediate isometric crystal is also small only about 300 ∼ 400 μm, strongly indicating there was a large temperature gradient near surface when melting occurred. Combined with plasma operation parameters and temperature evolution, the melting of TZM tiles were concluded to be induced by the transient heat flux during plasma disruption. These results imply the transient heat flux during plasma disruption in EAST can severely destruct the metal PFMs and should not be ignored, suggesting the active mitigation of plasma disruption is necessary for future long pulse and high parameters operation.

Simultaneous control of transient heat load induced by large-amplitude edge-localized modes (ELMs) and steady-state heat load on divertor targets under metal wall environment is crucial for steady-state operation of future tokamak fusion reactors, such as ITER and the China Fusion Engineering Test Reactor (CFETR). In the recent experiments, sustained partial energy detachment without confinement degradation has been achieved in the Experimental Advanced Superconducting Tokamak (EAST) in high-performance grassy-ELM H-mode with q95 ~ 5.9 by a newly developed detachment feedback control scheme, in which we first used electron temperature (Tet) measured by divertor Langmuir probes to identify the onset of energy detachment, and then the system switched to the feedback control of total radiation power measured by absolute extreme ultraviolet (AXUV) system. Tet around the upper outer strike point was successfully maintained less than 8 eV with seeding of 80% Ne and 20% D2 mixture from upper outer divertor, and the total radiation power was maintained ~1.4 MW, around 52% of injected power. There was no significant decrease of the plasma stored energy and H98,y2 factor (~1) over the entire detachment feedback control process. These experiment results demonstrate good compatibility of the high-performance grassy-ELM regime with radiative divertor. In order to confirm the compatibility in a wider range, stable partial energy detachment in grassy-ELM H-mode with relatively lower q95 (~5.4) was also achieved in EAST through the newly developed integrated-feedback-control technique. The new detachment feedback control without confinement degradation in grassy-ELM H-mode provides a candidate mode for EAST long-pulse operation in the future with well control of ELM-induced transient and steady heat fluxes on the divertor target.