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ITERではタングステン(W)ダイバータ板等で反射した光の混入によるプラズマ発光計測の精度悪化が問題視されており、この反射特性は壁面の表面状態に依存する。本研究では、様々な熱負荷を与えたITERダイバータ用Wモノブロックに対してゴニオフォトメーターを用いて双方向反射率分布関数(BRDF)を計測した。熱負荷を与える前のWモノブロックでは切削痕によって異方的な反射特性を示す一方、熱負荷を与えたWモノブロックでは再結晶化によって表面が粗くなり、拡散的な反射特性を示した。熱負荷による表面構造変化に応じたBRDFの変化について報告する。, 第38回 プラズマ・核融合学会 年会

ITERダイバータの外側垂直ターゲット（OVT）プロトタイプ及び・実機用のCuCrZr合金-ITERグレード管の量産が、プラズマ対向材料等の他の材料と同様に進行中である。OVT完成後も水漏れ防止や電磁力に対する耐久性の観点から、設計で要求される機械的特性を有し、かつ結晶の異常粒成長を抑えたCuCrZr管を製造することが求められている。そのため，OVTに使用する前に機械的特性の試験を行い，CuCrZr管の量産時の機械的特性と品質を確認している。本論文では、OVTのプロトタイプおよび実機に使用されるCuCrZr管に要求される重要な機械的特性の一つである引張特性について報告する。量産時の抜き取り検査の結果，CuCrZrチューブは，PFU製造時の溶体化や時効などの熱処理を経ても，要求される引張特性を満たすことができた。, The 30th International Toki Conference on Plasma and Fusion Research

ITERダイバータのプラズマ対向機器、特に日本が調達を担当する外側垂直ターゲット(OVT)の調達スケジュールや設計を紹介するとともに、それを実現するための技術開発状況として、アーマ材と冷却管の接合技術、組立、検査の各テーマにおいて、プラズマ対向機器受熱部の製造法や接合部検査、プラズマ対向ユニット製作において重要となるWモノブロック管のギャップ管理に関して計測方法やプロトタイプでの実績を紹介した。合わせて、QSTで整備を進めている高温ヘリウムリーク試験装置の概要を紹介した。, 核融合エネルギーフォーラム スクレイプ・オフ層及びダイバータ物理サブクラスター、炉材料サブクラスター、ブランケットサブクラスター、ダイバータサブクラスター（令和3年度第1回）合同会合

Power exhaust scenario and divertor design for a steady-state Japan (JA) DEMO and a pulse Europe (EU) DEMO1 have been investigated as one of the most important common issues in Broader Approach DEMO Design Activity. Radiative cooling is a common approach for the power exhaust scenario. For the JA DEMO, development of the divertor design appropriate for high Psep/Rp ~30 MWm-1 is required, while radiation fraction in the main plasma (fradmain = Pradmain/Pheat) is ITER-level (0.40-0.45) and the exhaust power above the L to H-mode power threshold (fLH = Psep/PthLH) is large margin (~2). For the EU DEMO1, larger fradmain (= 0.67) and smaller fLH (=1.2) plasma is required, using higher-Z impurity seeding, in order to apply ITER-level devertor (Psep/Rp = 17 MWm-1). ITER technology, i.e. water cooling with W-monoblock and Cu-alloy (CuCrZr) heat sink, is a baseline for JA and EU to handle the peak heat load of 10 MWm-2-level, and neutron flux and dose are comparable. For the JA DEMO, two different water-cooling pipes, i.e. CuCrZr and F82H steel, are proposed. For the EU DEMO1, heat sink consists of all Cu-alloy pipe, and several choices of the heat sink components have been developed appropriate to the high neutron irradiation condition. These JA and EU approaches of the power exhaust scenario will provide important case-studies for the future decision of the DEMO divertor design.

ITER divertor outer vertical target (OVT) consists of 22 plasma facing units (PFUs) and 2 stainless support structures (SSSs) including a complex internal structure for coolant water. Pressure drop and flow rate in each PFU are identified as important parameters for acceptable range in terms of the pumping power of cooling water system and performance of heat removal. In this study, computational fluid dynamics (CFD) and a transient reactor analysis code (TRAC-PF1) which is simpler simulation than the CFD have been performed to estimate the pressure drop and flow distribution in 11 PFUs mounted on an OVT SSS (half of the OVT). Results of both prediction methods showed the proper pressure drop and the homogeneous flow distribution within acceptable range of design requirement. To verify these numerical simulations, the results were compared to experimentally measured values in the flow test. This study demonstrated that the TRAC-PF1 as well as the CFD are applicable tools for further optimization of the divertor OVT cooling structure in view of manufacturing.

ITER divertor outer vertical target (OVT) consists of 22 plasma facing units (PFUs) and 2 stainless support structures (SSSs) including a complex internal structure for coolant water. Pressure drop and flow rate in each PFU are identified as important parameters for acceptable range in terms of the pumping power of cooling water system and performance of heat removal. In this study, computational fluid dynamics (CFD) and a transient reactor analysis code (TRAC-PF1) which is simpler simulation than the CFD have been performed to estimate the pressure drop and flow distribution in 11 PFUs mounted on an OVT SSS (half of the OVT). Results of both prediction methods showed the proper pressure drop and the homogeneous flow distribution within acceptable range of design requirement. To verify these numerical simulations, the results were compared to experimentally measured values in the flow test. This study demonstrated that the TRAC-PF1 as well as the CFD are applicable tools for further optimization of the divertor OVT cooling structure in view of manufacturing.

The refined design of full-W ITER divertor outer vertical targets (OVT) has the combination of plasma-facing units (PFUs) variants with bevelled W monoblocks to prevent leading edge overheating. Full-length PFU prototypes have withstood against the repetitive heat load of 20MW/m2 and they met the requirements for W technologies qualification programed by the IO. Upon the successful results, JADA has launched series of qualification programs toward the mass production of components. Furthermore, JADA successfully verified the water flow design inside the OVT is fulfilling the ITER requirements. The flow test was done using a OVT prototype with dummy cooling tubes.

Tungsten coating with a thickness of 1 mm on reduced-activation ferritic/martensitic steel (RAF/M) F82H (Fe-8Cr-2W), which is a leading structural material candidate for DEMO, have been produced by Atmospheric Plasma Spraying (APS) and Vacuum Plasma Spraying (VPS). Heat loading experiments on thermal response and fatigue using an electron beam have been carried out on W-coated F82H brazed on oxygen free high purity copper (OFHC) block with a cooling tube to evaluate their possibility as a plasma-facing armor in the fusion device. In addition, quantitative analyses about temperature profiles and thermal stress have been carried out using FEA. Thermal response experiments show that the temperatures increased monotonically with increasing heat flux. Surface temperature of the VPS-W/F82H/OFHC is always lower than that of the APS-W/F82H/OFHC. Surface modification, exfoliation and crack are not formed by thermal fatigue experiments up to 200 cycles at a heat flux of 3.2 MW/m2. Thermal response experiments under steady state condition have been successfully modeled by FEA. Interfacial strength of VPS-W and F82H is also discussed using the FEA results.

For the series production of the divertor outer vertical target (OVT) of the ITER, approximately 200,000 tungsten (W) monoblocks as plasma-facing material are planned to be manufactured. However, thus far, the variation in the properties of the W monoblocks has not been investigated. In this study, the chemical impurity level, material properties, and response to cyclic heat loading (i.e., durability and surface deformation) of W monoblocks manufactured using W plates supplied by A.L.M.T. Corp., Japan (ALMT) and Nidec Sankyo CMI Corp., Japan (NSCM) were investigated. On examining the manufactured lots (four from ALMT and two from NSCM), it was concluded that the variation in the material properties might have been caused by variations in the impurity level of the raw powder, contamination during manufacturing, and changes in the processing parameters. The results of the high heat flux test indicated that all the examined lots met the acceptance criteria. Hence, the variations observed in the chemical impurity level and material properties (e.g., density, Vickers hardness, and grain size) of the W plates supplied by ALMT and NSCM were acceptable. Therefore, W from the two suppliers will be employed in the full-scale plasma-facing unit prototype, which will be used for the final qualification of the material and OVT manufacturing technology.

The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

Through R&D for a plasma facing units (PFUs) in an outer vertical target of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency succeeded in demonstrating the durability of the W divertor shaped by an electrical discharge machining (EDM). To prevent melting of W armors in the PFUs, an adequate technology to meet requirements of a geometrical shape and a tolerance is one of the most important key issues in a manufacturing process. From the necessity, the EDM has been evaluated to control the final shape of the W armor. Though the EDM was known to be advantages such as an easy workability, a potential disadvantage of presence of micro-cracks on the W surface appeared. In order to examine a potential effect of the micro-crack on a heat removal durability, a high heat flux testing was carried out for the W divertor mock-up with the polish and the EDM. As the result, all of the W armors endured the repetitive heat load of 1000 cycles at an absorbed heat flux of more than 20 MW/m 2 , which strongly encourages the realization of the PFUs of the ITER full-W divertor with the various geometrical shape and the high accuracy tolerance.

Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m 2 for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

The plasma facing unit (PFU) of ITER divertor is repeatedly exposed to high heat flux, and a reliable heat removal capability is crucial for its adequate performance. The cyclic high heat flux test (HHFT) that has been conducted to evaluate the reliability of PFU reveals that the surface modification and macro-crack formation in W monoblock, which is the plasma facing material, may hinder the operation of fusion reactors. Therefore, optimization of both manufacturing technology and material development is necessary for reliable operation. In this study, we investigate the dependence of material properties on the surface modification and macro-crack formation in W monoblocks. W monoblocks manufactured by two different suppliers were examined, and the results of cyclic HHFT revealed that the heat removal capability was not degraded in all the examined W monoblocks. Further, it was observed that the W monoblock exhibiting characteristic grain structure and tensile property suffered less surface modification without macro-crack formation due to cyclic heat loading.

Overview of plasma facing components (PFCs) in JT-60SA is introduced, including upper divertor, lower divertor, inboard first wall and outboard first wall. These PFCs are upgraded in 4 stages along plasma experiment of JT-60SA; In the Stage1, only partial inboard first wall and upper divertor are installed with graphite tiles and inertial cooling. In the Stage 2, remaining PFCs including outboard first wall and lower divertor are installed, and the inboard surface is fully covered. In the Stage 3, vertical targets of lower divertor are replaced by CFC monoblock targets with water cooling. In the Stage 4, whole carbon wall is replaced by tungsten wall, and remote handling for lower divertor is introduced. Three R&D on major issues concerning PFCs are presented. Mock-ups of CFC monoblock target were fabricated and it showed good heat removal of the expected maximum heat load of 15 MW/m2 without degradation of performance. Samples tungsten coating on CFC were fabricated and they showed no severe damage by surface temperature up to 2000 C. Laser welded samples showed good appearance and no defect was found by RT.

量研機構はITER国内機関(JADA)として、国内メーカと協力し、2012年からITER用タングステン(W)ダイバータの製作技術開発に着手し、実機長プラズ対向ユニット(PFU)のプロトタイプ製作及び加熱試験を行い、製作工程確立の目途を立てることができた。実機シリーズ製作では、1本あたり約150個Wモノブロックを接合したPFUを1300以上製作する必要がある。このようなPFUの大量生産に向けた技術確証試験として、1度にPFUの接合を行う熱処理試験を実施し、PFU試験体に対して超音波探傷による接合部の非破壊試験を実施し、本PFU熱処理工程が実機製作に適用出来る見通しを得た。, 第12回核融合エネルギー連合講演会

Japan Atomic Energy Agency (JAEA) is in progress for technology qualification toward full-tungsten (W) ITER divertor outer vertical target (OVT), especially, tungsten monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m2. To demonstrate the armor heat sink bonding technology and heat removal capability, 6 small-scale W monoblock mock-ups manufactured by different bonding technologies using different W materials in addition to 4 full-scale prototype plasma-facing units (PFUs). After non-destructive test, the W components were tested under high heat flux (HHF) in ITER Divertor Test Facility (IDTF) at NIIEFA. Consequently, all of the W monoblocks endured the repetitive heat load at 20 MW/m2 for 1000 cycles (requirements 20 MW/m2 for 300 cycles) without any failure. In addition to the armor to heat sink joints, the load carrying capability test on the W monoblock with a leg attachment was carried out. In uniaxial tensile test, all of the W monoblock attachments with different bonding technologies such as brazing and HIPping withstand the tensile load exceeding 20 kN that is the value more than twice the design value. The failures occurred at the leg attachments or the W monoblocks, rather than the bonding interface of the W monoblocks to the leg attachment.

In order to develop and validate the high performance tungsten monoblock technology, the full-tungsten divertor qualification program was defined. As the first step, small-scale mock-ups were manufactured and successfully tested under the required high heat flux loads. The test results demonstrated that the technology is available in Japan and Europe. Post-tests observation of the loaded W monoblocks showed generation of self-castellation – a crack along coolant tube axis. The cause of the self-castellation was discussed and a tungsten material characterization program is being developed with the objective to understand mechanical properties that influence the occurrence of the self-castellation.

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

Hydrogen diffusion coefficients in a reduced activation ferritic/martensitic steel (F82H) and an oxide dispersion strengthened F82H (ODS-F82H) have been determined from depth profiles of plasma-loa...

High density W coatings on reduced activation ferritic martensitic steel (RAF/M) have been produced by Vacuum Plasma Spraying technique (VPS) and heat flux experiments on them have been carried out to evaluate their possibility as a plasma-facing armor in a fusion device. In addition, quantitative analyses of temperature profile and thermal stress have been carried out using the finite element analysis (FEA) to evaluate its thermal properties. No cracks or exfoliation has been formed by steady state and cyclic heat loading experiments under heat loading at 700 °C of surface temperature. In addition, stress distribution and maximum stress between interface of VPS-W and RAF/M have been obtained by FEA. On the other hand, exfoliation has occurred at interlayer of VPS-W coatings near the interface between VPS-W and RAF/M at 1300 °C of surface temperature by cyclic heat loading.

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

After the successful completion of the prequalification activity for ITER divertor procurement, Japanese Domestic Agency (JADA) and ITER Organization (IO) have entered into the procurement arrangement of divertor Outer Vertical Target (OVT) in June 2009. In accordance with the arrangement, JADA is going to fully procure the outer target components which correspond to 60 divertor cassettes. JADA has started to manufacture an OVT full-scale prototype in order to pick out/solve technical and quality issues and to establish a rational manufacturing process toward the start of the series of production of the OVT components to be installed in ITER. This paper presents the overview of JADA's manufacturing activity and the procurement schedule on the divertor outer target procurement.

Japan Domestic Agency (JADA) carried out RD (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu–W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20 MW/m2. All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3 kN and 8 kN, respectively.

Hydrogen including a trace amount of tritium was loaded on the edge surface of an F82H rod. After the loading, the rod was held at 298 or 323 K to allow hydrogen diffuse in and release out. Tritium tracer techniques have been applied to determine hydrogen depth profiles and hydrogen release rates by using an tritium imaging plate technique and a liquid scintillation counting technique, respectively. The depth profiles were composed of a surface localized component within 200 μm of the surface and a diffused component extending over 1 mm in depth. The apparent hydrogen diffusion coefficients obtained from the depth profile of the diffused component are near the extrapolated value of the literature data determined at higher temperatures. The surface localized component, which is attributed to trapping at surface oxides and/or defects, was released very slowly to give apparent diffusion coefficients much smaller than those determined from the diffused component.

Tritium retention of a carbon layer deposited in the gap of plasma-facing materials is a primary concern for next step fusion devices. In this study, deuterium concentration and carbon deposition profile in a gap were investigated for carbon layers prepared by using deuterium arc discharge with carbon electrodes. The deuterium retention was measured with thermal desorption spectroscopy (TDS). The discharge pressure was varied from 0.8 to 36 Pa. The amount of deposited carbon into the gap decreased exponentially with the increase of the depth. The atomic ratios of D/C of the carbon layers prepared at 0.8 and 36 Pa were approximately 0.1 and 1.0, respectively. For the carbon layer prepared at 0.8 Pa, most of retained deuterium was released in the form of D 2 and HD. On the other hand, at 36 Pa, approximately a half of retained deuterium was desorbed in the form of hydrocarbon, CD 4 and C 2 D 4 .

Aiming at a design study of the first wall structure for next fusion devices and cost-down, the heat transfer characteristics of the first wall structure with CFC tiles bolted to a heat sink were investigated with a variety of graphite sheets and fixing-bolt torque conditions. The heat load tests of the mock-up were performed with the heat fluxes of 1 and 3 MW/m2 on JAEA electron beam irradiation system. The experimental results showed that the first wall structure with 0.1-mm thickness × 3 PGSs (Pyrolytic Graphite Sheet; Panasonic Co., Ltd.) had the highest heat transfer performance. Furthermore, the first wall structure with the PGSs withstood the heat fluxes of 1 MW/m2 × 100 s and 3 MW/m2 × 20 s. The maximum surface temperatures of the CFC tile were 500 °C, 800 °C, respectively. These results indicated that the bolted tile structure with the graphite sheets can be utilized as the first wall at the long pulse operation with the heat flux of several MW/m2.

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed wi...

Fluid flow and heating tests on the full-scale Test Blanket Module (TBM) First Wall (FW) mock-up under the TBM-relevant operating condition were carried out in an ion beam irradiation system with a pressurized high temperature water loop in JAEA to verify the fabrication process and design of the TBM First Wall and demonstrate its heat removal capability. The TBM FW mock-up is made of reduced activation ferritic/martensitic (RAFM) steel, F82H, and its parts such cooling channels and plates were assembled with Hot Isostatic Pressing (HIP) method. In the fluid flow test, the cooling water at room temperature is supplied to 15 parallel flow paths in the mock-up. The flow distribution in each path inside the First Wall mock-up is compared with a numerical simulation. This result shows that no severe cross-sectional deformation of the entire flow path in the mock-up takes place during the fabrication process. In the heating test, the mock-up was exposed to repetitive hydrogen ion beam irradiation with the maximum heat flux of 0.5 MW/m2. No indication of joint defect of the HIP joint like a hot spot was observed during 80 irradiation cycles.

In the first set of ITER divertor, the bottom segment of the inner and outer vertical target consists of the carbon-fiber-reinforced composite monoblock, the interlayer of the oxygen-free high-conductivity Cu and the cooling tube. In this part of the divertor, the capability to remove the high heat flux up to 20 MW/m 2 has to be provided. Therefore, the method of performance estimation for the outer vertical target is one of the key issues in procurement of the actual equipment. Non-destructive examination with the infrared thermography is required as one of the performance tests to detect the overall defect in the monoblock. In this study, the thermographic examinations for the outer vertical target of qualification prototype have been performed by the facility of infrared non-destructive examination for divertor (FIND) built in the Japan Atomic Energy Agency. The profile of the maximum temperature difference in FIND corresponds to that of the surface temperature of the monoblock in the high heat flux testing in Russian Federation. It indicates that FIND is able to predict the position of maximum temperature in the high heat flux testing.

Development of Water Cooled Solid Breeder (WCSB) TBM, the primary candidate of ITER Test Blanket Module (TBM), has been performed in Japan, according to the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. The TBM milestones consist of milestones on safety assessment, module qualification and design integration in ITER. For the safety milestones, essential source terms were evaluated, and failure modes and effect analysis (FMEA) was performed. Based on the results of FMEA, safety assessment was performed. For the milestones of the design integration, detailed structural design of the TBM and the interface structure with the ITER test port were performed. Based on the design, performance analysis such as thermo-mechanical analysis in over-pressurization was performed. For the milestones of the qualification of fabrication technology, essential fabrication technology was developed and near full size first wall of the TBM was successfully fabricated and demonstration of the integrity in heat flux equivalent to ITER. The development of the WCSB TBM is showing steady progress toward the installation in ITER.

High heat flux experiments using a hydrogen–helium mixture beam have been carried out on powder metallurgy tungsten (PM-W) and ultra fine grain W–TiC alloy (W–0.5 wt%TiC-H 2 ). The energy of is 18 keV. Beam flux and heat flux at the beam center is 2.0 × 10 21 atoms/m 2 s and 7.0 MW/m 2 , respectively. Typical ratio of He/D ion is 0.25. Beam duration is 1.5–3 s and interval of beam shot start is 30 s. The samples are irradiated up to a fluence of 10 22 –10 24 He/m 2 by the repeated irradiation pulses. After the irradiation, surface modification by the irradiation and hydrogen retention, surface composition have been investigated. Surface modification by hydrogen–helium mixture beams is completely different from results of single beam irradiation. In particular, mixture beam irradiation causes remarkably high hydrogen retention.

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m 2 , 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.

Two types of plasma spray tungsten coatings on ferritic/martensitic steel F82H made by vacuum plasma spray technique (VPS) and air plasma spray technique (APS) were examined in this study to evaluate the possibility as plasma-facing armor. The VPS-W/F82H showed superior properties. The porosity of the VPS-W coatings was about 0.6% and most of the pores were smaller than 1–2 μm and joining of W/F82H and W/W was fairly good. Thermal load tests indicated high potential of this coating as plasma-facing armor under thermal loading. In case of APS-W/F82H, however, porosity was 6% and thermal load properties were much worse than VPS-W/F82H. It is likely that surface oxidation during plasma spray process reduced joining properties. Remarkably, both coatings created soft ferrite interlayer after proper heat treatments probably due to high residual stress at the interfaces after the production. This indicates the potential function of the interlayer as stress relieve and possible high performance of such coating component under thermal loads.

As part of the development of plasma-facing components (PFCs) for fusion machines, JAEA has been developing high-performance cooling tubes with pressurized water flow. Along this line, a cooling tube with a helical triangular fin on its inner surface has been recently proposed for application in a fusion DEMO to enhance heat removal. Since the fin can be machined via simple mechanical threading, this tube is called as a “screw tube”. The divertor cooling conditions for the JAEA DEMO design will be a pressure of 4 MPa and an outlet temperature of 200 °C, in order to improve the thermal efficiency of power generation. In the this study, the effect of subcooling on the critical heat flux (CHF) by the screw tube has been investigated under DEMO divertor-relevant conditions, with the local pressure of 4 MPa and the inlet coolant temperature up to 180 °C. A test sample of the screw tube is made of pure Cu instead of F82H, a candidate structural material for the DEMO divertor. The results show that the ICHF values of the screw tube are more than double the values of the smooth tube at the inlet coolant temperature of 180 °C, although the temperature rise of the cooling water by 140 K leads to a reduction in the ICHF by almost half as compared with those values at room temperature.

A cyclic helium beam with energy 18.7 keV has been used to irradiate stress relieved and recrystallized powder metallurgy tungsten up to a fluence of ∼10 23 He/m 2 . After the helium irradiations, the samples have been exposed to pulsed high heat loading by an electron beam with a heat flux of 1 GW/m 2 and duration of 1, 1.5 or 2 ms. Surfaces of irradiated samples become smooth as a result of the electron beam irradiation of 1 GW/m 2 for 1 ms, due to atom migration on the surface at high temperature. On the other hands, the surface of irradiated samples is partially melted and resolidified. However, smoothing does not occur around the melted area. Stress relieved tungsten is melted by irradiation of a heat flux of 1 GW/m 2 and duration of 1.5 ms, but recrystallized tungsten is not melted by the same heat loading.

JAERI has successfully developed plasma-facing components (PFCs) for ITER application, and the R&D activities are now being shifted to the development of the PFCs for a fusion power plant. This paper presents the R&D activity on the PFCs, such as the first wall and divertor, for a fusion power plant. The PFCs of the power plant will be subjected to heavy neutron irradiation and high heat/particle fluxes from the plasma during the continuous operation. In the present design of the PFCs, the candidate structural material is a reduced activation ferritic-martensitic steel, F82H, selected for low activation and high robustness against neutron irradiation, and the candidate armor material is tungsten, selected for the low sputtering yield and low tritium retention. To achieve PFCs using these materials, there are several R&D issues to be solved: (1) development of a cooling structure with high heat removal capability, (2) development of a suitable form of tungsten, (3) selection of a bonding technique for F82H and tungsten and (4) evaluation of structural integrity. Recent JAERI R&D achievements and status of these issues are presented.

A solid breeder blanket was selected as the primary candidate blanket of the fusion power demonstration plant (DEMO plant) in Japan. Test blanket module (TBM) testing in ITER is the most important milestone for the development of the blanket of the DEMO plant. Therefore, the major development effort has been focused on the development of the TBMs in Japan. This paper presents a structural concept of solid breeder TBMs cooled by pressurized water and helium gas. In the sub-module concept which was developed, a slitted box structure was adopted for both TBMs to simulate the DEMO blanket structure and to endure the over-pressurization in case of the coolant ingress to box structure. The breeder and neutron multiplier formed by small pebbles are packed separately in a layered structure. Detailed internal structure was developed with consideration of fabrication procedure, coolant flow route and purge gas flow route. Coolant flow route was considered from the view points of thermal–hydraulic and heat transfer performances. As for He purge gas flow passages, separated passages for tritium breeder layers and neutron multiplier layers were adapted to avoid tritium migration to the neutron multiplier layers.

For development of high heat flux components for fusion machines, JAERI has developed high performance cooling tubes using pressurized water flow. A cooling tube with triangular fins on its inner surface has been tested. Because the fins can be machined by a simple mechanical threading, this tube is called a screw tube . The present study is intended to investigate heat transfer characteristics of the screw tube with M10 thread of 1.5-mm-pitch, which has the highest critical heat flux (CHF) in the previous experimental campaign compared with the tubes with various threads of different pitches. In addition, another CHF tests have been performed to examine the reduction of the heat removal performance of the screw tube caused with the increase of cooling water temperature up to 100 °C at the local pressure of 2 and 4 MPa. The results of CHF experiments are compared with an existing correlation for a smooth tube to evaluate its heat removal limitation at higher temperature condition compared with the previous ones.

This paper describes the R&D of plasma facing components, especially the joining technology of the tungsten/copper joint for the ITER divertor application. The major concern for this joint is detachment due to the large differences in the thermal expansion coefficient and the modulus of elasticity of each metal. One solution is to separate the joint area into small pieces to reduce interfacial stress at the joint. From this viewpoint, JAERI proposes the hot-press method using small tungsten rods pressed into cylindrical bores in the copper substrate at high temperature. A divertor mock-up was fabricated to investigate the durability of this joint. No failures of the mock-up occurred in a thermal cycle experiment involving 3,000 cycles of 5 MW/m2 heat flux. Using numerical analyses, the contact heat transfer coefficient between the tungsten and copper was determined to be 2,400 W/m2/K for this mock-up. Though this value is slightly inadequate to maintain the surface temperature below the recrystallization ...

To investigate the synergistic effects of heat load and hydrogen irradiation, cyclic heat load tests with a hydrogen beam and a comparable electron beam were performed for high purity CVD-tungsten coatings. Surface modification was examined as a function of the peak temperature by changing the heat flux. Scanning Electron Microscopy analysis showed that the surface damage caused by the hydrogen beam was more severe than that by the electron beam. In the hydrogen beam case, cracking at the surface occurred at all peak temperatures examined from 300 °C to 1600 °C. These results indicate that the injected hydrogen induces embrittlement for the CVD-tungsten coating.

The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

The first fabrication and heating test of a large-scale carbon-fiber-composite (CFC) monoblock divertor mock-up using an annular flow concept has been performed to demonstrate its manufacturability and thermomechanical performance. This mock-up is based on the design of the lower part of the vertical target of the International Thermonuclear Experimental Reactor (ITER) divertor adapted for the annular flow concept. The annular cooling tube consists of two concentric tubes: an outer tube made of CuCrZr and an inner stainless steel tube with a twisted external fin. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. A basic mechanical examination of CuCrZr undergoing the brazing heat treatment and finite element method analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1000 thermal cycles of 20 MW/m{sup 2} for 15 s and 3000 cycles of >10 MW/m{sup 2} for 15 s, which simulates the heat load condition ofmore » the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test although in the tile with exposure to the heat flux of 20 MW/m{sup 2}, the erosion depth has been measured as 5.8 and 8.8 mm at the 300th and 500th cycles.« less