Deuterium retention in molten salt electrodeposition tungsten coatings

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Deuterium retention in molten salt electrodeposition tungsten coatings Hai-Shan Zhou a , Yu-Ping Xu b , Ning-Bo Sun c , Ying-Chun Zhang c , Yasuhisa Oya d , Ming-Zhong Zhao a , Hong-Min Mao b , Fang Ding a , Feng Liu a , Guang-Nan Luo a,b,e,f,∗ , EAST contributorsa a

Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China Science Island Branch of Graduate School, University of Science and Technology of China, Hefei, China c School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing, China d Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, Japan e Hefei Center for Physical Science and Technology, Hefei, China f Hefei Science Center of Chinese Academy of Science, Hefei, China b

h i g h l i g h t s • • • •

We investigate D retention in electrodeposition W coatings. W coatings are exposed to D plasmas in the EAST tokamak. A cathodic current density dependence on D retention is found. Electrodeposition W exhibits lower D retention than VPS-W.

a r t i c l e

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Article history: Received 30 November 2015 Received in revised form 3 June 2016 Accepted 28 July 2016 Available online xxx Keywords: Hydrogen isotopes Retention Tungsten coating Graphite Molten salt electrodeposition EAST tokamak

a b s t r a c t Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W. © 2016 Elsevier B.V. All rights reserved.

1. Introduction Tungsten (W) has been selected as the plasma-facing material (PFM) for ITER (International Thermonuclear Experimental Reactor) divertor [1] due to its outstanding material property. To understand the impacts of W-PFM on ITER plasma operation, extensive research activities are being organized in the fusion

∗ Corresponding author at: Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, China. E-mail address: [email protected] (G.-N. Luo).

community. Recently, an ITER-like full W upper divertor has been successfully commissioned in EAST (Experimental Advanced Superconducting Tokamak) and a W first wall is proposed as well [2]. Although W plasma-facing components (PFC) can be fabricated by joining bulk W to the heat sink with existing technology, e.g. hot isostatic pressing (HIP) [3], the production of large size W-PFC is still challenging and costly, which may limit the prospect of using W as the PFM in fusion reactors. The typical of the first wall of fusion power reactors is as following: (1) the wall area will be as large as ∼1000 m2 ; (2) heat flux from plasma to the wall is significantly lower than that to the divertor target and (3) the armor

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Fig. 1. Plasma exposure experiments using the MAPES system in the EAST tokamak.

layer must be thin to reduce thermal stress and to relieve neutron capture effect (i.e. to increase the tritium breeding ratio in the blankets) [4]. Taking all these into account, coating is considered to be a promising technology to manufacture the PFC (except divertor target) for fusion reactors. Various W-coatings have been developed and tested for fusion purpose, e.g. vacuum plasma spraying W (VPS-W), physical vapor deposition W (PVD-W) and chemical vapor deposition W (CVD-W) [5–7]. Some coating materials exhibit relatively good compatibility with hydrogen isotopes plasmas (e.g. low retention) [5]. In several large- and medium-sized fusion confinement facilities, coating technology has already been utilized to prepare the first wall [7–9]. Molten salt electrodeposition is another attractive method to fabricate the W-PFCs because of its simple technics, capability to cover complex surfaces and relatively low cost. A high density (>97%) and thick (>1 mm) W coating has been successfully achieved with this technology by researchers from University of Science and Technology Beijing [10–12]. Our previous study indicated that higher electrodeposition current density will lead to larger grain size [11]. In the present work, deuterium (D) retention behavior in the coatings prepared with different current densities is investigated in the EAST tokamak using a material manipulator. To make a comparison between the coating and other W materials, D-ion implantation experiments are performed in a laboratory facility as well.

2. Experimental 2.1. Material preparation Anhydrous Na2 WO4 and WO3 (Na2 WO4 :WO3 = 3:1 by mole ratio) molten salts were mixed in the eutectic composition into an alumina crucible to make the W coating. The crucible was heated up to 1173 K with a ramping rate of 5 K/min. A 15 × 10 × 5 mm graphite (IG-430) substrate and a 15 × 10 × 5 mm W plate were used as the working electrode and the counter electrode, respectively. After the deposition process, samples were cooled in the air and then cleaned in ultrasonic bath to remove adherent molten salts. Further details of this material can be found in the reference [10]. In this work, W coating samples were prepared with four different cathodic current densities: 30, 40, 50 and 60 mA/cm2 . The substrates with coatings were cut into 10 × 10 × 1 mm pieces and the W surfaces were mechanically polished for the following plasma

Fig. 2. Thermal desorption spectroscopy of D for the W coating samples after plasma exposure.

and ion irradiation experiments. The thickness of the coatings was measured to be 160–230 ␮m. 2.2. Plasma and ion exposure experiments Deuterium-plasma exposure experiments were performed during the 2015 spring EAST campaign. Four W samples were fixed on the sample holder of the material and plasma evaluation system (MAPES) at the mid-plane of EAST [13] (Fig. 1). The sample surface was 5 mm behind the limiter and the local electron temperature and density were measured to be Te = 5–10 eV and ne = ∼1 × 1018 m−3 by a Langmuir probe. The materials were irradiated by 367 shots and the total plasma exposure time was ∼2000 s. Thermocouples were attached to the samples and the measured temperature varied from 323 to 623 K due to the heat from plasmas. Before and after the plasma experiments, surface morphology of the samples was examined by scanning electron microscope (SEM). Finally, D retention properties of the W coatings were analyzed using a thermal desorption spectroscopy (TDS) device which was calibrated with D2 and H2 standard leaks. The samples were heated up to 1273 K with a ramping rate of 10 K/min. To make a direct comparison with other coating materials (e.g. VPS-W) and bulk W, D-ion exposure was also conducted using the triple ion implantation system at Shizuoka University in Japan [14]. For this case, the ion species is D2 + and the incident energy was 3 keV (i.e. the sample was irradiated with 1.5 keV D ions). The ion flux is 1 × 1018 D/m2 s and the fluence is 1 × 1022 D/m2 . The sample was kept at near room temperature during the experiment. After

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Fig. 3. (a) A polished surface of the molten salt electrodeposition W before plasma exposure; (b) a cross-section view of the W coating to check the coating thickness; (c) surface morphology of the W coating prepared with a cathodic current density of 30 mA/cm2 after plasma exposure in the EAST tokamak; (d) surface morphology after plasma exposure for the 40 mA/cm2 sample; (e) surface morphology after plasma exposure for the 50 mA/cm2 sample; (f) surface morphology after plasma exposure for the 60 mA/cm2 sample.

D-ion implantation, the sample was transferred into a TDS chamber without air exposure to evaluate D retention. 3. Results and discussion 3.1. Deuterium retention and surface damage from plasma exposure Deuterium desorption curves for the W coatings exposed to the EAST D plasma are plotted in Fig. 2. All the samples show two desorption peaks: a sharp one around 400 K and a broad one at higher temperatures. The D retention is found to depend on the cathodic current density applied to the sample under eletrodeposition of the W coating. For the 400 K desorption peaks, the D retention decreases as the current density increases. The 400 K desorption peak is usually considered to be the trapping effect of surface or dislocation loops [15,16]. Previous work on coating characterization indicates that at lower cathodic current densities, carbon impurity tends to accumulate at the coating surface [10], and the impurity may act as extra trapping sites for D. However, D release amount from the other desorption peak shows an opposite current density dependence. This retention behavior can be directly related to the surface morphology observed by SEM (Fig. 3). As the current density increases, the

plasma-exposed surface becomes rougher. Severe blistering can be observed on the 50 and 60 mA/cm2 sample surfaces (Fig. 3(e) and Fig. 3(f)), suggesting that D may be easier to accumulate in these samples to form high pressure bubbles. The data imply that coatings prepared with lower current density have better damage resistance to plasma bombardment. This result may be explained by the microstructure of the material. Coatings prepared with lower cathodic current density have smaller grain size [11] that can suppress blistering and D retention [17]. Fig. 4 shows the total retention for the four samples. Generally, D retention in the high current density samples is about two or three times larger than that in low density ones. This effect implies that lower cathodic current density should be more suitable if the molten salt electrodeposition technology is used to prepare the PFCs for fusion reactors.

3.2. Deuterium retention behavior after the ion implantation experiment Thermal release behavior of D from the D-ion irradiated sample is shown in Fig. 5. The TDS curve for polished VPS-W and polycrystalline W irradiated with D ions in the same facility [14] is also shown in Fig. 5 for comparison. The ion implantation parameters for the three samples are exactly the same.

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some candidate structural materials (e.g. copper alloys and reduced activation ferritic/martensitic steels), which will limit the application of this technology in manufacturing the PFCs of fusion reactors such as CFETR (Chinese Fusion Engineering Testing Reactor) [18]. Heat treatment after coating is a possible solution, which, however, may be engineering risky and costly. A more feasible way is to develop low temperature molten salt electrodeposition technics for fusion applications. Acknowledgments

Fig. 4. Cathodic current density dependence of total D retention.

The authors would like to thank Prof. Yuji Hatano from Toyama University for fruitful discussions on D retention mechanisms. This work is supported by National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB105001, 2015GB109001), the National Natural Science Foundation of China (Nos. 11505232, 11405201), Technological Development Grant of Hefei Science Center of CAS (No. 2014TDG-HSC003), Scientific Research Grant of Hefei Science Center of CAS (No. 2015HSCSRG054), the Korea Research Council of Fundamental Science and Technology (KRCF) under the international collaboration & research in Asian countries (No. PG1314), the Joint Sino-German research project GZ 763. References

Fig. 5. Comparison of D retention properties for polished molten salt electrodeposition W (MSE-W), vacuum plasma spraying W (VPS-W) and polycrystalline W samples [14].

Only one desorption peak has been found for the molten salt electrodeposition W sample, while the VPS-W and the polycrystalline W have several peaks. The total D retention in the electrodeposition W is about half of that in the other two materials, which is favorable from the view point of reducing fuel retention in fusion reactors. Unfortunately, examination of the surface morphology after ion implantation for the electrodeposition W sample is not possible because they are directly transferred to the TDS chamber without air exposure. From the single desorption peak of the electrodeposition W sample one can expect that the surface defect should be in a single form and the damage should be much milder than that of plasma irradiated samples. 4. Conclusion Deuterium retention properties of molten salt electrodeposition W are tested in the EAST tokamak and a laboratory ion implantation facility. Coatings prepared with low cathodic current density exhibit relatively good blistering resistance and low D retention. These results suggest that electrodeposition W can be an alternative first wall material for existing carbon wall machines (e.g. DIII-D). R&D work on this coating will be continued and the technics should be further optimized based on the testing results. Although molten salt electrodeposition W shows pretty good compatibility with hydrogen isotopes plasma/ion, it should be noted that the coating preparation temperature is still too high for

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Please cite this article in press as: H.-S. Zhou, et al., Deuterium retention in molten salt electrodeposition tungsten coatings, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.07.023