Vol:.(1234567890)

Tungsten (2020) 2:72–82https://doi.org/10.1007/s42864-020-00036-8

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REVIEW PAPER

A brief summary of tungsten technology development in Korea

Suk‑Ho Hong1

Received: 25 April 2019 / Revised: 4 June 2019 / Accepted: 4 June 2019 / Published online: 9 March 2020 © The Nonferrous Metals Society of China 2020

AbstractA review of tungsten technology development in Korea is briefly given. According to the upgrade plan of Korea super-conducting tokamak advanced research (KSTAR) tokamak associated with Korean fusion energy R&D program, graphite plasma-facing components will be replaced with tungsten-based ones to handle the high-peak heat load caused by an increase of heating power up to 26 MW. Brazing technique to bond tungsten was developed and tungsten blocks were manufactured. Blocks were installed at the central divertor in KSTAR and exposed to high heat flux. Under high heat flux and long-pulse discharge, tungsten blocks were severely damaged. Molten tungsten materials show movements towards the high field side, which is j×B direction. The COMSOL® modeling described the melting event quantitatively well. The failure of a water cooling system with a metal wall environment during a long-pulse plasma operation is very critical.

Keywords Demonstration reactor · Divertor · Tungsten

1 Introduction

The use of fusion energy in the sun can be realized and uti-lized on earth by constructing a device called “tokamak”. In the fusion reaction, a so-called “hydrogen burning process”, six protons fuse together resulting in a helium (He) atom and two protons, is the most effective energy production process in nature. Furthermore, fuels—deuterium (D) and tritium (T)—are easily obtained from the ocean (D) and lithium (T). Tokamaks create plasmas confined inside very strong magnetic fields. The plasma is hot enough to fuse D atoms into T, 3He, and 4He through the sequential fusion reaction.

The components inside the vacuum vessel, the so-called plasma-facing components (PFCs), are the final destination of particles, heat flux, and radiation: PFCs are exposed to harsh environments, and consequently, the lifetime of them is directly connected to the interactions between plasma and the surrounding surfaces of PFCs, which is called “plasma–surface interaction” (PSI). Recently, as the research interests on PSI move from low atomic number (Z) to high Z metal PFCs, issues such as (1) erosion/recrystallization/

melting of PFCs due to high heat flux; (2) high Z impurity accumulation in core plasma will be critical for the long-pulse tokamak operation.

Tungsten (W) is considered as one of the favorite PFC materials because of its excellent physical and chemical properties such as the high thermal conductivity, high melt-ing point, high sputtering resistance, and low deuterium/tritium retention. As international thermonuclear experi-mental reactor (ITER) has chosen full tungsten divertors from the beginning of its first campaign [1], Joint European Torus (JET) transformed the machine configuration for ITER-relevant experiments, which is the so-called ITER-like wall (ILW) consisting of beryllium and a full tungsten diver-tor. The axially symmetric divertor experiment (ASDEX) upgrade (AUG) has changed the inner wall to tungsten coated graphite tiles [2, 3]. The experimental advanced superconducting tokamak (EAST) has a full tungsten diver-tor since 2018 and tungsten environment in steady-state tokamak (WEST) started with several tungsten monoblock fingers and tungsten coated graphite blocks [4, 5].

Initial design limits of PFCs in Korea superconducting tokamak advanced research (KSTAR) were 20 s of opera-tion time with a 16 MW input power (KBSI internal report: KSTAR physics validation review documents, 1997). Since carbon-based graphite tiles cannot tolerate heat flux more than 3.5 MW·m−2 in such a condition, KSTAR has a PFC upgrade plan to employ the tungsten divertor, and thus,

Tungstenwww.springer.com/42864

* Suk-Ho Hong sukhhong@nfri.re.kr

1 DEMO Technology Division Advanced Technology Research Center, National Fusion Research Institute, Daejeon 34133, Korea

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National Fusion Research Institute started the development activities of tungsten first wall components [6, 7]. For long-term technology development, the pre-conceptual study of the Korean demonstration reactor (K-DEMO) divertor was also performed [8].

This paper reviews the tungsten technology development associated with Korean fusion energy R&D program per-formed since 2012. In Sect. 2, the status of the tungsten tech-nology development in Korea is summarized. In Sect. 3, the research activities related to the tungsten PFCs, especially the damage and melting of tungsten PFCs are reviewed. In Sect. 4, a summary is given.

2 Status of the tungsten technology development in Korea

2.1 Development of flat‑type tungsten bonding technology

To develop required tungsten bonding technology, we started with the bonding between tungsten plate and CuCrZr plate since 2013: In this section, we briefly review the main results reported in our previous publications [9].

A tungsten plate [ASTM B760-86 (1999)] was chosen as a plasma-facing material (PFM) and CuCrZr (ASTM C18150) was selected as a PFC material and cooling/heat sink mate-rial [10]. Direct bonding between tungsten and CuCrZr is not easy due to the large residual stress coming from the mismatch of their expansion coefficients. A thin interlayer of oxygen-free-copper (ASTM C10200) was employed for low yield strength and elastic modulus [11, 12]. Table 1 summa-rizes the properties of the materials used for the brazing pro-cess. The filler alloy with a thickness of 0.05 mm was used. The size of the brazing sample was 50 mm in length, 40 mm in width, and 27 mm in height [10 mm W, 2 mm oxygen-free high conductive (OFHC), and 15 mm CuCrZr], respectively.

The brazing process consists of five steps: (1) a nickel plate of a thickness of 5 µm was plated on tungsten prior to the brazing process for the improvement of wettability and bonding strength of the brazing filler, and a heat treatment

at a temperature of 600 °C for an hour was followed, which enhanced the quality of the nickel plating; (2) W was brazed onto CuCrZr at 980 °C for 30 min in vacuum; (3) fast cool-ing from 980 to 400 °C was followed; (4) the aging treat-ment for CuCrZr was carried out for ~ 180 min at 470 °C; (5) cool down to the room temperature [14]. Figure 1 shows the waveform of the W/Cu/CuCrZr brazing process.

The ultrasonic test (UT) was used for the detection of any defects between the brazed boundaries. The probe was a flat type with a size of 6.35 mm in diameter with the operating frequency of 10 MHz. Defects of a size larger than 2 mm in diameter could be detected. Figure 2 shows UT images of the samples brazed with different loadings, namely 5 kPa (sample A), 10 kPa (sample B), and 20 kPa (sample C). Only small defects in red circles were detected at the edge part of the W/Cu interface, while many defects were identified in the whole area of Cu/CuCrZr interface.

The joint conditions between W and CuCrZr were ana-lyzed by scanning electron microscopy (SEM). Figure 3 shows SEM images of samples with different loadings of 5, 10, and 20 kPa. Those images indicate clear micro defects at the boundary of the W/Cu joint, but the UT could not find the defects with the size smaller than 10 µm (Fig. 2). It is found that the bonding samples with a load higher than 100 g·cm−2 have very clean interface except for small num-bers of tiny voids.

A shear strength test at room temperature using the universal testing machine with a crosshead speed of 0.5 mm·min−1 was followed by a UT test [9]. It was found that the shear strength values of sample A, B, and C were 113, 112, and 119 MPa, respectively, and sample C brazed with a surface roughness of 6 µm Rs showed the highest value of 119 MPa. From the results, we chose the optimum brazing condition as 6 µm Rs with the loading parameter of 20 kPa at 980 °C for 30 min [9].

Table 1 Physical properties of materials (at room temperature) Reproduced with permission from Ref. [13] Copyright 2017 Institute of Electrical and Electronics Engineers (IEEE)

Density/(g·cm−3)

Coefficient of thermal expansion/(10−6 K−1)

Young’s modulus (E)/GPa

Thermal conductiv-ity/(W·K−1)

W 19.3 4.5 398 173CuCrZr 8.9 15.7 127.5 318OFHC 8.9 16.7 82.4 401

Fig. 1 Waveform of the W/Cu/CuCrZr brazing process: (1) heating up to 850 °C for 30 min, and further heating up to 980 °C; (2) braz-ing for 30 min; (3) quenching with a rate of 90 °C·min−1; (4) aging at 470 °C for 180 min; (5) cool down Reproduced with permission from Ref. [13] Copyright 2017 IEEE

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2.2 High heat flux test (HHFT) of flat‑type tungsten blocks

For the HHFT, six tungsten-brazed mock-ups with a cool-ing tube were fabricated, as shown in Fig. 4 (#1, #2, … #6). The size of the mock-up was slightly smaller than the previous test samples, 50 mm × 28 mm × 37 mm with a cooling tube (12 mm in inner and 15 mm in outer diam-eter). The UT test before HHFT shows that the bonding between each interface is very clean and solid.

The KoHLT-EB (Korea heat load test facility-electron beam) in KAERI (Korea Atomic Energy Research Insti-tute) was used for the HHFT. An 800 kW electron gun (Von Ardenne, Germany) was equipped with a maximum beam power of 300 kW and maximum accelerating volt-age of 60  kV, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber [15]. The surface tem-perature of the mock-up was monitored by a pyrometer and the local temperature was measured using two K-type ther-mocouples, which were inserted into a position of 3 mm below tungsten and 2 mm below the Cu/CuCrZr interface.

Fig. 2 UT images of samples bonded by different loadings Reproduced with permission from Ref. [13] Copyright 2017 IEEE

Fig. 3 SEM images of samples bonded by different loadings Reproduced with permission from Ref. [13] Copyright 2017 IEEE

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There was a calorimeter to measure the temperature of water coolant (the flow rate of 0.35 kg·s−1, pressure of ~ 0.35 MPa, and temperature of ~ 18 °C).

Table 2 shows the results of the HHFT on the tungsten-brazed mock-ups under a heat flux of 5 MW·m−2 with a duty cycle of 50% (heating for 20 s and cooling for 20 s). The mock-up #2(D) was tested for 2000 cycles, while the mock-ups #3(E) and #4(F) were tested for 1000 cycles. The tem-perature waveform during the HHFT and the top/side views of mock-ups after HHFT are shown in Fig. 5. There was no delamination and failure at the bonding joints of all the tungsten-brazed mock-ups: the maximum surface tempera-ture of tungsten was ~ 520 °C, and the maximum temperature

of tungsten and CuCrZr blocks were ~ 415 °C and 243 °C, respectively. The high heat flux (HHF) test is performed up to a heat flux of 8 MW·m−2 and no failure was observed.

2.3 Fabrication of real size mock‑up

A real size mock-up with cooling lines is manufactured, as shown in Fig. 6. The height of the mock-up is the same as that of the divertor tile (3 cm), but half of the width (6 cm). The mock-up has two cooling lines at the top and bottom. The mock-up was tested under extreme HHF for a melting experiment applying up to a heat flux of 130 MW·m−2 for several seconds long (not shown here).

Fig. 4 Three tungsten-brazed mock-ups for the HHFT and results of the UT test of six mock-ups at W/Cu interface and Cu/CuCrZr interface

Fig. 5 Waveform of HHFT (left) and top/side views of mock-ups after the HHFT (right) Left: reproduced with permission from Ref. [16] Copy-right 2016 Elsevier. Right: reproduced with permission from Ref. [13] Copyright 2017 IEEE

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2.4 Fabrication of tungsten monoblocks

ITER tungsten monoblocks are manufactured using a two-step process: The first step is to cast OFC (oxygen-free cop-per) on the hole of tungsten monoblocks, and then, a CuCrZr tube is bonded on the tungsten/OFC units by the HRP (Hot Radial Pressing) or brazing. We developed a process of a two-step brazing method to bond all the joints between tung-sten and OFC, and OFC and CuCrZr. The brazing method is easier and simpler than HRP. And, the strength of the braz-ing joint is stronger than one of the casting joints between tungsten and OFC. Figure 7 shows the sequence of the two-step brazing process and fabricated tungsten monoblocks for the HHF test. The blocks are tested under ITER qualification

condition, i.e., a heat flux of 10 MW·m−2 up to 5000 cycles. The UT and shear strength test are performed before and after the heat flux test, and microstructures and the diffusion depth of all the brazed joints are analyzed by field-emission scanning electron microscope (FESEM).

3 Damage and melting test of tungsten blocks by power loading in tokamak

Castellated structure of monoblocks has intrinsic leading edges. They are heavily bombarded by energetic particles during plasma shots. Consequently, the temperature of the leading edge could reach the melting point of tungsten, resulting in the movement of molten tungsten in the direc-tion of j × B [17–19]. By tilting and shaping of the castel-lated structure, and by changing its angle with respect to the field line, leading edges can be hidden from the plasma particles. Nevertheless, tilting and shaping of a structure reduce projected areas leading to the increase of the heat flux on the surface [20]. On the other hand, the peak temperature of the block is significantly reduced.

For a systematic study on the power loading of the leading edge, multi-purpose castellated tungsten blocks were fabricated. A rectangular block with a dimension of 30 mm × 20 mm × 12 mm, made of 5 mm W on the top, 2 mm Cu as the intermediate layer, consisting of a 13 mm CuCrZr base, is designed. Blocks then were modified to have four different leading edge heights of 0.3, 0.6, 1.0, and 2.0 mm with one perfectly aligned case. Blocks were mounted on a stainless steel base. Both poloidal and toroidal gap distances between blocks are fixed at 0.5 mm. Blocks were installed in KSTAR as shown in Fig. 8 (left).

The blocks were exposed to various plasmas during a campaign and observed by a high-resolution infrared (IR) camera (FLIR®, resolution: 640 × 320, 125 Hz or 2.7 kHz acquisition speed) [21]. Figure 8 (right) shows images of heat load patterns on the leading edge of a height of 0.3 mm under the edge localized mode (ELM) and inter-ELM phase (#16279), obtained by the divertor IR camera with 2.7 kHz acquisition speed. Prior to the ELM, the block was

Table 2 Results of the HHFT on tungsten-brazed mock-ups Reproduced with permission from Ref. [13] Copyright 2017 IEEE

Mock-up Coolant(inlet)

Absorbed heat flux/MW·m−2

Number of cycles

On/off time/s

Failure in joints

Surface temp./°C

Tungsten temp./°C

CuCrZr temp./°C

#2(D) WaterP = 0.35 MPaṁ = 0.35 kg·s−1

T = 18 °C

~ 5 2000 20/20 No 520 384 243

#3(E) ~ 5 1000 20/20 No – 415 243#4(F) ~ 5 1000 20/20 No 471 369 242

Fig. 6 Fabricated real size mock-up for divertor of KSTAR with cool-ing lines (the white box in the bottom indicates the carbon-based KSTAR divertor tile block)

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exposed to a heat flux of ~ q⊥ = 1.0 MW·m−2. Since the field line angle at the central divertor is in a range of 2°–3°, and q|| = 22.93 MW·m−2 (assuming 2.5°). Transient heat flux

during the ELM is ~ 5–50 MW·m−2 in ~ 400 μs (τIR, charac-teristic decay time of IR signal after an ELM crash) which gives much brighter edges in the IR image.

Fig. 7 Sequence of the two-step brazing process and fabricated tungsten monoblocks for the HHF test

Fig. 8 Experimental setup (W blocks) for the leading edge power loading experiment and divertor IR images Reproduced with permission from Ref. [21] Copyright 2018 Elsevier

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Figure 9 shows the visible camera image taken after the campaign, showing the damage of 2 mm leading edge. Cracks, droplets, and damaged structures were clearly iden-tified. Noting that the 2 mm height of the leading edge was extremely high, but there was no melting event even with ELMs. This is due to the position of the strike point, because the outer strike point was usually located in the middle or below the W block tile.

Figure 10 shows the side and top views of the W castel-lated blocks exposed to long-pulse H-mode plasmas in 2016 and 2017 KSTAR campaign. In both cases, the outer strike points were located at the center of the blocks, for at least 70 s long. It is noted that the graphite protection cover in front of the blocks was recessed ~ 2 mm making the blocks 2 mm leading edges with respect to the perfectly aligned ones. As a result, top W layers as well as the Cu interlayer below the W layer were severely damaged and melted. It was also found that the molten droplets of W were directed towards the high field side due to j × B force as reported in Refs. [17–19].

To understand the 2  mm leading edge melting, a COMSOL® analysis is performed: A W block was mod-eled, and time-dependent transient analysis was set with a heat load of 2 MW·m−2 on the block without cooling.

Because long-pulse shot was performed usually at the end of a run day, the initial temperature of the W block was set to 300°C. The spatial temperature distribution and time evolution of the temperature of W and Cu lay-ers were traced. Figure 11 shows the spatial temperature distribution and the time evolution of the W block. It is noted that the model only describes the steady-state heat load. After the exposure for 1 s, the temperature of the W layer reaches 1200°C, and 1400°C within 4 s, which is the W recrystallization temperature. This result shows that the surface of the W layer starts to be damaged within 4 s. After ~ 51 s, the W temperature reaches the melting point (3300°C). On the other hand, the melting tempera-ture of the copper layer is reached after ~ 20 s. The results depict time evolution of melting event of tungsten blocks in Fig. 12. At first, the tungsten leading edge was exposed and the temperature was increased. About 20 s later, the Cu layer started to melt, and thermal contact of tungsten and the Cu layer became poor. As the contact became poor, melting of the tungsten layer followed. The temperature of the block increases more and more, resulting in further melting of the Cu layer.

Fig. 9 Visible camera image of the 2 mm leading edge after 2016 KSTAR campaign Reproduced with permission from Ref. [21] Copyright 2018 Elsevier

Fig. 10 Top and side views of W castellated blocks exposed to long-pulse H-mode plasmas in a 2016 and b 2017 KSTAR campaign a Reproduced with permission from Ref. [21] Copyright 2018 Elsevier

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4 A short remark on the KSTAR upgrade plan towards tungsten divertor within Korean fusion energy R&D program

According to the R&D roadmap of Korean fusion energy development, the policy goal of the phase 2 (2012–2026) is “development of core technology for DEMO”, and one of the key technologies that we have to urgently develop is tungsten technology, i.e., bonding technology as well as fabrication technology.

Along the development, KSTAR plays an important role in developing high-performance plasma opera-tion for preparations for the ITER Operation. KSTAR achieved a long-pulse H-mode discharge almost reach 90 s in 2018 [22]. Unfortunately, carbon-based graphite tiles used in KSTAR could not tolerate the heat flux more than 3.5 MW·m−2, parts of divertor tiles were severely damaged. On the other hand, KSTAR has an upgrade plan of heating and current drive up to 26 MW, as shown in Fig. 13.

Consequently, KSTAR decides to upgrade its divertor to tungsten monoblock-based one. There are some pre-requisites [24]: (1) upgrade for 10 keV, and long-pulse

(> 300 s) plasma; (2) accommodate various plasma scenar-ios (ITER-like shape, high-βp, ELM suppression, etc.); (3) enough space for advanced divertor diagnostics; (4) lower single-null plasma with a dome-type upper plate for top-launching Heating and Current Drive (H&CD). Figure 14 shows some examples of KSTAR’s new tungsten divertor designs [24] including ITER-like configuration (Fig. 14a, b), high delta configuration (Fig. 14c, d) and the mixture of ITER-like and KSTAR high-βp (Fig. 14e). The shape of new divertor will be decided soon, and scrape-off layer plasma simulation (SOLPS)-ITER modeling is ongoing to provide design support.

5 Summary

In this paper, tungsten technology development associ-ated with Korean fusion energy R&D program was briefly reviewed. In the KSTAR’s upgrade plan, PFCs will be upgraded to tungsten-based ones to handle the high heat flux caused by input power up to 26 MW. In preparation of the KSTAR upgrade as well as K-DEMO, tungsten bonding technology based on brazing was developed and

Fig. 11 Spatial temperature distribution and time evolution of temperature of block with 2 MW·m−2 heat load on the 2 mm leading edge Repro-duced with permission from Ref. [21] Copyright 2018 Elsevier

Fig. 12 A cartoon of the time evolution of melting event shown in Fig. 11

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tungsten blocks were manufactured. Fabricated tungsten blocks were shaped to have various edges and exposed to e-beam and long-pulse H-mode plasmas in KSTAR. It was clearly seen that tungsten blocks could be severely damaged during the plasma depending on the exposure time. Molten tungsten materials were moved towards j × B

direction (high filed side). The results of COMSOL® tran-sient modeling with heat loads could reproduce melting event of the blocks quantitatively. This indicates that fail-ure of the cooling system during long-pulse operation with metal walls will be very critical.

Fig. 13 KSTAR’s heating and current drive upgrade plan [23] (NBI neutral beam injection, ECH electron cyclotron heat-ing, ICRF ion cyclotron radio frequency, LHCD lower hybrid current drive, HCD helicon cur-rent drive)

Fig. 14 Examples of KSTAR’s new divertor design. a, b ITER-like configuration; c, d high delta configuration; e combination of ITER and KSTAR [24]

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Acknowledgements This research was partially supported by Min-istry of Science and Information and Communication Technology (ICT) under DEMO project (Grant No. CN1901-6), under Non-ITER procurement (Grant No. IN1910-2), and KSTAR project (Grant No. EN1901).

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Dr. Suk‑Ho Hong received his Ph.D. degree in plasma physics f r o m R u h r - U n i v e r s i t y Bochum(RUB, Germany) in 2014. Since then, Dr. Hong has experienced several European Tokamaks such as Tore–Supra, the Axially Symmetric Divertor Experiment (ASDEX) Upgrade, and Joint European Torus (JET) until 2008. Dr. Hong has worked at the National Fusion Research Institute (NFRI, Korea) since 2008, and studied

edge plasmas of fusion devices, plasma–surface interaction pro-cesses, dusts in plasmas, wall-conditioning techniques, and devel-oped diagnostic components for the KSTAR tokamak. As a senior scientist at NFRI, Dr. Hong is leading the DEMO technology divi-sion at NFRI, and developing the DEMO relevant advanced technology.