Deuterium Plasma Research Articles

Turbulence and E × B flow correlations across the L-H transition in DIII-D deuterium and hydrogen plasmas.

The isotope dependence of the low- to high confinement-mode (L- to H-mode) transition power threshold PLH presents significant challenges for the initial (non-nuclear) hydrogen operations phase of the international thermonuclear experimental reactor (ITER). Here, we examine the isotope dependence of turbulence and E × B flow correlation properties in the L-mode edge plasma, leading up to the L-H transition. At marginal auxiliary power (near PLH), turbulence is initially suppressed periodically during limit cycle oscillations (LCO) that precede the transition to sustained H-mode confinement. We present evidence that the long-range (toroidal) correlation of the E × B edge plasma flow across the LCO phase is much weaker in hydrogen than in deuterium in the DIII-D tokamak, congruent with the higher threshold power PLH required to access LCO and H-mode in hydrogen. Concomitantly, the time required to initially quench edge turbulence via localized edge E × B flow shear is significantly longer in hydrogen (1-1.5 ms) than in deuterium (approx. 100 µs). No toroidal long-range correlation of the turbulence amplitude is observed, in agreement with expectations based on the relatively short poloidal turbulence correlation length. Radial edge turbulence and flow correlation lengths are longer in deuterium than in hydrogen plasmas as one would expect from 'naïve' gyro-Bohm isotope transport scaling, despite the substantially higher thermal flux across the last closed flux surface in hydrogen before the L-H transition. This article is part of a discussion meeting issue 'H-mode transition and pedestal studies in fusion plasmas'.

Deuterium retention in CVD diamond: Combined experimental and computational study

Diamond’s intrinsic hardness, excellent thermal conductivity and low atomic number make it a highly promising candidate as a plasma facing material. However, as with the previously used graphite, concerns over tritium retention and resultant chemical etching have so far limited research interest in the use of synthetic diamond. In order to study tritium retention, the DELPHI facility at the Culham Centre for Fusion Energy was used to expose polycrystalline diamond samples to a deuterium plasma. Deuterium ions were accelerated to an energy of 0.2 keV to 1 keV for a 5 h exposure time, achieving a fluence of approximately 5.5 × 1021 D m−2. Exposed samples were analysed using Thermal Desorption Spectroscopy. Increasing implantation energy resulted in additional D2 release peaks observed in the 800–1100 K temperature range that were not seen at lower energies. These peaks were interpreted as an additional bonding mechanism, a likely candidate for which is inter-grain deuterium. Experimental work was complemented with molecular dynamics simulations on the University of Bristol’s high performance computer—Blue Crystal Phase 4. In these simulations, both varying implantation energy and the presence of grain boundaries were explored. A two-step etching mechanism was observed, in which the surface initially swelled before carbon removal. No significant differences could be observed on the inclusion of a grain boundary at the energies tested.

ECRH experiments on Tokamaks SST-1 & Aditya-U and ECRH upgradation plan for SST-1

A 42GHz-500kW ECRH system [1-6] is used to carry out various experiments related to plasma breakdown and ECR heating on tokamaks SST-1 and Aditya-U. The system has been upgraded with new anode modulator power supply to launch two ECRH pulses to carry out breakdown and heating simultaneously. In SST-1, ECRH system is used routinely for plasma breakdown at fundamental harmonic, approximately 150kW power is launched for 70ms to 150ms duration and consistent plasma start-up is achieved in SST-1. In the recent experiments, second EC pulse is also launched at the plasma flat-top to heat the plasma, some heating signatures are seen but more experiments will be carried out to confirm the plasma heating with ECRH. In Aditya-U tokamak, simultaneous plasma breakdown and heating experiments are carried out successfully [2]. In the first pulse around 100kW power in fundamental O-mode is launched for 70ms duration for the breakdown at low-loop voltage and around 150kW ECRH power for 50ms duration is launched in second EC pulse to heat the plasma. In case of Aditya-U, plasma heating is observed clearly as soft X-ray signal increases sharply with ECRH. In AdityaU tokamak, deuterium plasma experiments have been carried out and ECRH launched at the flat-top of deuterium plasma current. In deuterium plasma also ECR heating is observed as soft X-ray signal increases with ECH power. For SST-1, ECRH system is being upgraded with another 82.6GHz system, this system would be used to carry out plasma heating and start-up at second harmonic. The 82.6GHz system is already connected with the SST-1 tokamak, the old 82.6GHz-200kW Gyrotron will be upgraded to 400kW system to carry out effective heating experiments on SST-1 at higher ECRH power. The paper discusses the recent results of ECRH experiments carried out on tokamaks SST-1 & Aditya-U and presents the upgradation plan of EC system for SST-1.

Investigation ofTemeasurements discrepancies between ECE and Thomson diagnostics in high-performance plasmas in JET

For high-temperature JET and TFTR discharges, electron cyclotron emission (ECE) measurements of central electron temperature were systematically found to be up to 20% higher than those taken with Thomson scattering. In recent high-performance JET discharges, centralTemeasurements, performed with LIDAR Thomson scattering and the X-mode ECE interferometer, have been studied in a large database, including deuterium (DD), and deuterium-tritium plasmas (DT). Discrepancies betweenTemeasurements have been observed outside of the experimental uncertainties. ECE measurements, at highTe, have been found to be higher or lower than those of LIDAR, depending on the specific plasma scenario. In addition, discrepancies between the peaks of the second and third harmonic ranges of the ECE spectrum have been interpreted as evidence for the presence of non-Maxwellian features in the electron distribution function. These comparisons seem to suggest that such features can be found in most of the high-performance scenarios selected in this JET database.

Investigation of the surface temperature in contact with plasma by two-color pyrometry

A high-speed two-color pyrometer has been developed that allows measuring the intensity of thermal radiation in temperature range of 100-3500oC with 2 μs time resolution. The analysis of the main factors affecting the measurement of surface temperature in the conditions of its interaction with plasma is carried out. It is shown that at plasma temperature of more than 10 eV and moderate concentration of particles, the wall plasma can be considered transparent to thermal radiation. Experimental studies on non-contact measurement of temperature of surface interacting with plasma on Globus-M2 tokamak and plasma gun test bench have been carried out. Sawtooth temperature fluctuations were detected on the wall of tokamak chamber near the exit of outer separatrix branch. The period of these flashes was 3-4 ms, and duration was 2 ms. The maximum wall temperature measured by a high-speed pyrometer exceeded the temperature measured by infrared camera. Measurements of temperature of the front surface of W-Li sample were carried out at plasma gun test bench under conditions of deuterium plasma jet cyclic action under loads simulating regimes of plasma current disruption in tokamak. The sample temperature reached more than 3000^oC during <1 ms. The thermal conductivity of the sample degraded with an increase in number of plasma exposure cycles, which was expressed in progressive decrease in target cooling rate after the pulse. Using a high-speed pyrometer, it is proposed to control the formation of heat-insulating foils on surface interacting with plasma Keywords: two-color pyrometer, thermal radiation, surface temperature, materials irradiation.

Исследование температуры поверхности, контактирующей с плазмой, методом двухцветной пирометрии

A high-speed two-color pyrometer has been developed that allows measuring the intensity of thermal radiation in temperature range of 100-3500 0C with 2 µs time resolution. The analysis of the main factors affecting the measurement of surface temperature in the conditions of its interaction with plasma is carried out. It is shown that at plasma temperature of more than 10 eV and moderate concentration of particles, the wall plasma can be considered transparent to thermal radiation. Experimental studies on non-contact measurement of temperature of surface interacting with plasma on Globus-M2 tokamak and plasma gun test bench have been carried out. Sawtooth temperature fluctuations were detected on the wall of tokamak chamber near the exit of outer separatrix branch. The period of these flashes was 3-4 ms, and duration was 2 ms. The maximum wall temperature measured by a high-speed pyrometer exceeded the temperature measured by infrared camera. Measurements of temperature of the front surface of W-Li sample were carried out at plasma gun test bench under conditions of deuterium plasma jet cyclic action under loads simulating regimes of plasma current disruption in tokamak. The sample temperature reached more than 3000 0C during < 1 ms. The thermal conductivity of the sample degraded with an increase in number of plasma exposure cycles, which was expressed in progressive decrease in target cooling rate after the pulse. Using a high-speed pyrometer, it is proposed to control the formation of heat-insulating foils on surface interacting with plasma.

Effect of the E × B drift on the redistribution of the divertor particle flux in the HL-2A ECRH plasmas

Double-peaked distribution (DPD) of particle flux has only been observed on the outer divertor target in electron cyclotron resonance heating deuterium plasmas with B×∇B toward the X-point in the HL-2A tokamak using high spatiotemporal Langmuir probe arrays. The experimental results demonstrate that the formation of the DPD is mainly due to the enhanced poloidal E×B drift flow stimulated in the divertor region, which is dependent on the plasma density, heating power, and divertor structure. The experimental results are qualitatively consistent with the SOLPS simulation. The experiment also shows that the formation of the DPD might be related to the enhanced cross field transport in the far scrape-off layer. This experimental findings presented here reveal the crucial role played by the synergistic effect of poloidal E × B drift flow and the closed divertor structure in the redistribution of the particle flux, which provides a potential way for the control of high heat flux in future fusion devices.

Ion cyclotron heating at high plasma density in Proto-MPEX

The physics of ion cyclotron heating (ICH) relevant to the steady-state linear machine MPEX (Material Plasma Exposure eXperiment) has been explored in its predecessor, short-pulse device: Proto-MPEX. MPEX will utilize fundamental ICH to increase heat flux at the target and produce ion temperatures and velocity distributions with improved fidelity to those found in a tokamak divertor region, in comparison to those produced by substrate biasing. In the experiments on Proto-MPEX described here, bulk ion temperatures up to ∼ 15 eV have been achieved with 20 kW net ICH power at 6.5 MHz, using ICH heating of a deuterium plasma produced by a helicon plasma source. The heat flux at the target has been observed to increase throughout the plasma cross section, including in the core region. Core Ti and target heat flux are observed to scale linearly with injected ICH power. Measurements of plasma loading and target heat flux as a function of the magnetic field strength at the antenna, together with modeling of the wave propagation from the antenna to the ion cyclotron resonance using the ANTENA and COMSOL codes with a warm plasma dielectric tensor, indicate that power is coupled to the core plasma via fast wave excitation of a kinetic Alfvén wave.

Runaway electron dynamics in Experimental Advanced Superconducting Tokamak helium plasmas

The generation of runaway electrons (REs) is observed during the low-density helium ohmic plasma discharge in the Experimental Advanced Superconducting Tokamak (EAST). The growth rate of hard x-ray (HXR) is inversely proportional to the line-average density. Besides, the RE generation in helium plasma is higher than that in deuterium plasma at the same density, which is obtained by comparing the growth rate of HXR with the same discharge conditions. The potential reason is the higher electron temperature of helium plasma in the same current and electron density plateau. Furthermore, two Alfvén eigenmodes driven by REs have been observed. The frequency evolution of the mode is not fully satisfied with the Alfvén scaling and when extension of the Alfvén frequency is towards 0, the high frequency branch is ∼50 kHz. The different spatial position of the two modes and the evolution of the helium concentration could be used to understand deviation between theoretical and experimental observation.

Peripheral temperature gradient screening of high-Z impurities in optimised ‘hybrid’ scenario H-mode plasmas in JET-ILW

Screening of high-Z (W) impurities from the confined plasma by the temperature gradient at the plasma periphery of fusion-grade H-mode plasmas has been demonstrated in the JET-ILW (ITER-like wall) tokamak. Through careful optimisation of the hybrid-scenario, deuterium plasmas with sufficient heating power ( 32 MW), high enough ion temperature gradients at the H-mode pedestal top can be achieved for the collisional, neo-classical convection of the W impurities to be directed outwards, expelling them from the confined plasma. Measurements of the W impurity fluxes between and during edge-localised modes (ELMs) based on fast bolometry measurements show that in such plasmas there is a net efflux (loss) between ELMs but that ELMs often allow some W back into the confined plasma. Provided steady, high-power heating is maintained, this mechanism allows such plasmas to sustain high performance, with an average D–D neutron rate of s−1 over a period of ∼3 s, after an initial overshoot (equivalent to a D–T fusion power of ∼9.4 MW), without an uncontrolled rise in W impurity radiation, giving added confidence that impurity screening by the pedestal may also occur in ITER, as has previously been predicted (Dux et al 2017 Nucl. Mater. Energy 12 28–35).

Surface-limited deuterium uptake of Ru films under plasma exposure

Blister formation has been an emerging research topic for extreme ultraviolet (EUV) mirrors exposed to hydrogen plasmas. Similar to plasma-facing materials in nuclear fusion reactors, it has been reported that blister formation in EUV mirrors is initiated by hydrogen uptake due to hydrogen ion or atom bombardment. However, the research so far has focused on Mo/Si multilayers exposed to only hydrogen ions or atoms, while the EUV mirror typically has a Ru capping layer facing hydrogen plasmas. We present experimental work to measure plasma-induced hydrogen uptake of Ru films. We bombarded our designed Ru-capped target with a low-temperature deuterium plasma and measured the deuterium retention using elastic recoil detection. Contrary to ion-driven deuterium uptake, the deuterium uptake rate of the Ru film had no dependence on the deuterium ion flux or energy after a period of plasma exposure. A reaction–diffusion model has been built to calculate the time evolution of deuterium retention, which well fits the experimental data. Based on this model, we conclude that the surface composition of the Ru film is the limiting factor for the deuterium uptake, which is seriously weakened when the surface is covered by Ru oxide. After the Ru oxide is reduced by the plasma, the uptake rate is predominantly driven by the deuterium surface coverage on metallic Ru. Our model also indicates that at the deuterium-populated Ru surface, deuterium has a low absorption barrier to penetrate the surface, which is supported by previously reported computational work.

Termination of graphene edges created by hydrogen and deuterium plasmas

Edge engineering is important for both fundamental research and applications as the device size decreases to nanometer scale. This is especially the case for graphene because a graphene edge shows totally different electronic properties depending on the atomic structure and the termination. It has recently been shown that an atomically precise zigzag edge can be obtained by etching graphene and graphite using hydrogen (H) plasma. However, edge termination had not been studied directly. In this study, termination of edges created by H-plasma is studied by high-resolution electron energy loss spectroscopy to show that the edge is sp2 bonded and the edge carbon atom is terminated by only one H atom. This suggests that an ideal zigzag edge, which is not only atomically precise but also sp2 bonding, can be obtained by H-plasma etching. Etching of the graphite surface with plasma of a different isotope, deuterium (D), is also studied by scanning tunneling microscopy to show that D-plasma anisotropically etches graphite less efficiently, although it can make defects more efficiently, than H-plasma.

Effect of anisotropic grain boundaries on the surface blistering of tungsten induced by deuterium plasma exposure

The effect of anisotropic grain boundaries on the surface microstructure changes and deuterium retention behavior in tungsten under deuterium plasma irradiation were studied. The samples named as NDW and RDW were cut from RD × TD plane and TD × ND plane of a rolled W plate (ND: normal direction, RD: rolling direction, TD: transverse direction), respectively. The results suggest that both surface morphology and deuterium retention in the RDW samples are suppressed compared to that of NDW samples for 1 h and 25 h deuterium irradiation. For 1 h exposure, the total deuterium retention in the NDW samples is about 1.6 times as much as that of RDW. However, as the exposure time increased to 25 h, the total deuterium retention in the NDW sample is two times the height of RDW sample. The experimental results show that the special grain structure of RDW can alleviate the surface morphology changes and decrease the deuterium retention in tungsten, which is interpreted by the lateral stress model of blistering.

Reducing the L-H transition power threshold in ITER-similar-shape DIII-D hydrogen plasmas

Recent dedicated DIII-D experiments in low-torque, ITER-similar-shape (ISS) hydrogen plasmas (at a plasma current I p ∼ 1.5 MA and ITER-similar edge safety factor q 95 ∼ 3.6) show that the L-H transition power threshold P LH can be reduced substantially (∼30%) with L-mode helium admixtures n He/n e ⩽ 25%. In the ensuing H-mode, helium ion fractions n He/n H remain below 25%. H-mode normalized pressure and confinement quality are only slightly affected by helium seeding, and Z eff ⩽ 2.15 (including helium and carbon content). The plasmas investigated here are electron-heat dominated, with temperatures T e(0)/T i(0) ⩾ 1 and edge heat flux ratio Q e/Q i(ρ = 0.95) ∼ 1.2–1.5. Without mitigation, P LH is higher by a factor of 2–3 in comparison to similar ISS deuterium plasmas. ISS hydrogen plasmas with lower plasma current I p ∼ 1 MA (increased edge safety factor q 95 ∼ 5.1) exhibit a substantially lower power threshold. This plasma current dependence, also observed previously on ASDEX-U and in JET, is not accounted for by the commonly used 2008 ITPA multi-machine threshold scaling, but could potentially allow H-mode access at marginal heating power during the initial plasma current ramp-up. Attempts to reduce P LH with low-field- and high-field-side hydrogen pellet injection, using 1.7 mm diameter pellets, have not demonstrated a robust threshold reduction, in contrast to successful earlier experiments with larger 2.7 mm pellets. Techniques for reducing P LH are very important for ITER, in particular for accessing H-mode in hydrogen plasmas during the Pre-Fusion Power Operation-1 (PFPO-1) campaign with marginal auxiliary heating power (20–30 MW of ECH).

Influence of impurity gas seeding into deuterium plasma on the surface modification, sputtering erosion and deuterium retention in W and W-La2O3 alloy

Surface modification, sputtering erosion and deuterium (D) retention were investigated of tungsten (W) and W-1 wt.% La2O3 (WL10) that were exposed to pure D plasma and D plasma with different impurities (He, N2, Ar). After the plasma exposure, the surface of WL10 is covered by blisters with various shapes, and their size and density increase with D fluence. When the fluence was higher than 8.6 × 1024 D m−2, the blister rupturing occurred causing a partially- and/or fully-opened lid features. The phase interface between La2O3 grains and W matrix was served as the nucleation sites for blistering and large-scale crack propagation near the surface. Small blisters were preferentially appeared on the grains with surface orientation of nearly (111) and that originated from the intra-granular crack at depths much closer to the sample surface. Micro-sized pits were formed on the WL10 surface after exposure to high fluence plasma, which was due to a higher sputtering rate of La2O3 grains during plasma exposure. For both materials, seeding He in D plasma can effectively inhibit the blister formation, and most serious blistering occurs with N seeding in D plasma. When exposed to Ar + D plasma, sparse small blisters were visible and most of blisters were ruptured for W, whereas there are some big blisters with long strip shape for WL10. The difference in blistering behavior between W and WL10 was attributed to the differences in microstructure, composition as well as ductility due to the presence of La2O3 phases. The sputtering yields increased slightly with He seeding in D plasma. N2 and Ar seeding in D plasma leaded to a significant increase in sputtering yields. As compared to the W, the WL10 exhibits a better sputtering erosion resistance in terms of the sputtering yield. The D retention in WL10 increases with D fluence and that was equal to or lower than that in W when exposed to the same conditions.

Neutron emission spectrometer to measure ion temperature on the Fusion Demonstration Plant.

General Fusion is building the Fusion Demonstration Plant to demonstrate a magnetized target fusion scheme in which a deuterium plasma is heated from 200eV to 10keV by piston-driven compression of a liquid-lithium liner. The multilayer coaxial time-of-flight neutron emission spectrometer is designed to measure the ion temperature near peak compression at which time the neutron yield will approach 1018 neutrons/s. The neutron energy distribution is expected to be Gaussian since the machine uses no neutral beam or radio-frequency heating. In this case, analysis shows that as few as 500 coincidence events should be sufficient to accurately measure the ion temperature. This enables a fast time resolution of 10 µs, which is required to track the rapid change in temperature approaching peak compression. We overcome the challenges of neutron pile-up and event ambiguity with a compact design having two layers of segmented scintillators. The error in the ion temperature measurement is computed as a function of the neutron spectrometer's geometric parameters and used to optimize the design for the case of reaching 10keV at peak compression.

Advances in Magnum-PSI probe diagnosis in support of plasma–surface interaction studies

Magnum-PSI is a linear plasma generator equipped with a superconducting magnet, assuring fusion devices relevant conditions at plasma–surface interface. The plasma column was diagnosed using 64 probes embedded in the target. The cross-sectional distributions of plasma parameters (floating potential, ion saturation current and electron temperature) were measured for hydrogen and deuterium plasmas under various discharge conditions. The radial profile of the floating potential across the plasma column can be described by a reversed Mexican hat-like wavelet, having the most negative potential at the center of the plasma column. The negativity of the floating potential diminishes when the discharge current increases or the magnetic field decreases. The axial gradient of the floating potential is reduced by increasing the magnetic field. The ion saturation current is maximum at the center of the plasma column, increasing with both the discharge current and magnetic field. The ion flux to the target, estimated from Thomson scattering (TS) data, was confirmed by probe measurements. The electron temperature estimated from the ion branch of the probe current–voltage characteristic is few times larger than that obtained from TS. By increasing the gas pressure in the target chamber, the time-dependent ion saturation current measured by probes changes from a constant average current (when the plasma column is attached to the target) to a fluctuating average current with scattered peaks (in a partially detached regime) which vanishes completely in the fully detached regime. With respect to hydrogen, the plasma column is wider in deuterium and is characterized by less negative floating potential distributions.

Power balance analysis at the L-H transition in JET-ILW NBI-heated deuterium plasmas

The understanding of the physics underlying the L-H transition has strong implications for ITER experimental reactor and demonstration power plant (DEMO). In many tokamaks, including JET, it has been observed that, at a particular plasma density, n e,min, the power necessary to access H-mode PL-H is minimum. In the present work, L-H transitions of JET deuterium plasmas heated by neutral beam injection (NBI) are studied for the first time by means of a power balance analysis to characterize the main contributions in the transition, through integrated transport modelling. In the pulses analysed, we do observe a minimum of the L-H power threshold in density, indicating the presence of density branches and of n e,min. Electron and ion heat fluxes at the transition are estimated separately. The electron/ion equipartition power results in favour of the ions, as shown by QuaLiKiz quasilinear gyrokinetic simulations, which predict a larger ion transport that causes T e > T i. The resulting edge ion heat flux also shows a clear change of slope below n e,min, similarly to ASDEX-Upgrade (AUG) NBI pulses (Ryter et al 2014 Nucl. Fusion 54 083003). JET NBI data are compared to radio-frequency heated AUG and Alcator C-mod pulses (Schmidtmayr et al 2018 Nucl. Fusion 58 056003), showing a different trend of the power, coupled to ions at the L-H transition with respect to the linearity observed in the radio-frequency heated plasmas. The presence of n e,min and the role of the ion heat flux is discussed in the paper, although it seems it is not possible to explain the presence of a PL-H minimum in density by a critical ion heat flux and by the equipartition power for the JET NBI-heated plasmas analysed.

Effect of Fe11+ ion combined with helium and deuterium plasmas irradiation on the transient thermal shock behaviors of pure and potassium-doped tungsten

The large-scale warm-rolled pure tungsten (PW) and hot-rolled potassium-doped tungsten (W-K) plates were prepared by powder metallurgy technology to meet the demands of International Thermonuclear Experimental Reactor (ITER) in engineering application. In order to evaluate the performance of plasma-facing materials (PFM) in a more realistic service environment as much as possible, the influence of combined irradiation effects of high-energy iron (Fe) ion, deuterium (D) and helium (He) plasmas on the transient thermal shock behavior in the PW and W-K was investigated. The results indicate that the PW can generate more vacancy- and dislocation-type defects than that of W-K after irradiation. The cracking thresholds for un-irradiated PW and W-K are >0.66 and 0.44–0.55 GW/m2, respectively. However, the cracking threshold for both samples decreased to 0.22–0.33 GW/m2 after combined irradiation, indicating that the combined irradiation can significantly reduce the thermal shock resistance of PW and W-K. In addition, according to the statistical results of the crack density and width on the surface of irradiated samples, irradiated W-K owns better thermal shock resistance than irradiated PW under the same thermal shock condition. These results were also verified by the determination of residual stresses. Besides, the relationship between the evolution of irradiation-induced defects and K bubbles is also discussed in detail. These studies can provide an important reference for the identification of future PFM and numerical simulation under combined irradiation plus high thermal load.

Surface morphology and deuterium retention in W and W-HfC alloy exposed to high flux D plasma irradiation

In this work, pure W and W-0.5wt%HfC alloy (WHC05) were fabricated by sintering and hot-rolling following the same processing route. After exposing to a high flux deuterium plasma irradiation with the D+ flux to three fluences of 6.00 × 1024, 2.70 × 1025 and 7.02 × 1025 D/m2, the evolution of surface morphology, deuterium retention and hardening behaviors in pure W and WHC05 has been studied. The SEM results show the formation of D blisters on the irradiated area, and with the increase of D implantation, the size of these blisters increases from 200 ∼ 500 nm (2.70 × 1025 D/m2) to 1 ∼ 2 μm (7.02 × 1025 D/m2) in WHC05 and from 1 ∼ 2 μm (2.70 × 1025 D/m2) to > 3 μm (7.02 × 1025 D/m2) in pure W, respectively. A higher D retention and obvious hardening are observed in pure W than that of the WHC05 alloy, indicating an improve radiation resistance in WHC05 compared to pure W.