Fabrication and high heat flux test with the first wall mockups for developing the KO TBM

Korea has proposed and designed a helium cooled molten lithium (HCML) test blanket module (TBM) to be tested in the ITER, in which Ferrite Martensite steel (FMS) and Be were used as the structural and armor materials of the first wall (FW), respectively. In order to develop the fabrication technology for a TBM structure, joining of FMS to FMS was developed and proved through the high heat flux (HHF) test in the previous study. Then, joining of Be to FMS has been developed with several interlayers and mockups were fabricated with hot isostatic pressing (HIP) in the present study. Mockups with Be joined to the FMS were successfully fabricated with a HIP (580 °C, 100 MPa, 2 h) by trying the different interlayers (1 μm-Ti/0.5 μm-Cr/5 μm-Cu and 1 μm-Cr/5 μm-Cu). The HHF tests with a Korea Heat Load Test (KoHLT-1 facility) were performed with 1000 cycles under 0.5 MW/m² heat flux. The test conditions were determined with ANSYS-11 and the test results were compared with the preliminary analysis ones. During the test, there was no sudden increase of temperature but UT and DT results after the test showed a delamination in the case of using Ti/Cr/Cu interlayer. But the mockup with the Cr/Cu interlayer showed a sound joining even after HHF test. Moreover, large mockups considering the TBM FW were designed and started their fabrication for developing the fabrication methods and procedure.     

Fabrication of Be/CuCrZr/SS mock-ups for ITER first wall

Three types of the mock-ups consisting of Be, CuCrZr and stainless steel (SS) were fabricated by a HIP joining method to demonstrate the fabricability of the ITER first wall. The effects of the mock-up design and the interlayer type on the joining strength of the Be/CuCrZr joint were investigated on the basis of the shear test results for the Be/CuCrZr joint specimens. The joining strength of the Be/CuCrZr ranged from 78 MPa to 204 MPa depending on the types of mock-ups and interlayers. The highest shear strength was obtained from the Be/CuCrZr joint specimens which were taken from the Be/CuCrZr/SS mock-ups with three 80 mm × 80 mm × 10 mm Be tiles and a Cr/Cu interlayer. The effect of the size and the number of Be tiles used in the mock-ups was not as significant as the interlayer selection. Even though the shear test is considered to be very useful to elucidate which interlayer is the most applicable for the joining of Be and CuCrZr, the final interlayer selection should be based on the results obtained from the high heat flux test for the mock-ups fabricated in this work.     

Interfacial properties of HIP joints between beryllium and reduced activation ferritic/martensitic steel

ITER test blanket modules are the most important components to validate energy production and fuel breeding for future fusion demonstration reactors. Reduced activation ferritic/martensitic steel is recognized as one of the promising structural materials for the breeding blanket systems. Beryllium is a primary candidate plasma facing materials for ITER blanket. In this work, the interfacial properties of Be/reduced activation ferritic/martensitic steel (RAF/Ms) joints were investigated for the first wall of an ITER test blanket module (TBM). The joints were produced by the solid-state hot isostatic pressing (HIP) method. Chromium (Cr) was used as a diffusion barrier with a thickness of 1 μm or 10 μm, formed by plasma vapor deposition on the Be surface. The HIPping was conducted at 1023 K and 1233 K with 160 MPa of static pressure. The temperatures are standard normalizing and tempering temperatures of F82H. EPMA showed the Cr layer effectively worked as a diffusion barrier at 1023 K. However, for the F82H/Be interface which underwent HIP at 1233 K followed by tempering a Be rich layer was formed. Bend tests revealed that a thin Cr layer and low temperature HIP is preferable. The joint with a thick Cr layer suffer from brittleness of Cr itself.     

Fabrication of W/FMS joint mock-ups using a hot isostatic pressing

Blocks of tungsten and ferritic–martensitic steel (FMS) were joined without any interfacial defects or cracks. For the joining, two times of a hot isostatic pressing (HIP) were performed. The first HIP (900 °C, 100 MPa, 1.5 h) facilitates the diffusion bonding between W and FMS. The second HIP (750 °C, 70 MPa, 2 h) corresponds to a tempering process to retain the mechanical properties of the FMS. As an interlayer material, titanium foil that can mitigate the thermal expansion difference between W and FMS was used. In addition, a molybdenum foil was inserted to prevent an unwanted bonding of W to a canning material. The lateral cracks in W plates, which were usually observed in the case of a conventional HIP process, were not observed when the molybdenum separator was used. W/FMS joint mock-ups with a dimension of 50 mm × 50 mm × 32 mm (T) were successfully fabricated. The shear strength of the joints was 89 MPa on average.     

Welding techniques development of CLAM steel for Test Blanket Module

Fabrication techniques for Test Blanket Module (TBM) with CLAM are being under development. Effect of surface preparation on the HIP diffusion bonding joints was studied and good joints with Charpy impact absorbed energy close to that of base metal have been obtained. The mechanical properties test showed that effect of HIP process on the mechanical properties of base metal was little. Uniaxial diffusion bonding experiments were carried out to study the effect of temperature on microstructure and mechanical properties. And preliminary experiments on Electron Beam Welding (EBW), Tungsten Inert Gas (TIG) Welding and Laser Beam Welding (LBW) were performed to find proper welding techniques to assemble the TBM. In addition, the thermal processes assessed with a Gleeble thermal-mechanical machine were carried out as well to assist the fusion welding research.     

R&D activities of the liquid breeder blanket in Korea

A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m². Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high heat flux test facility. For an electromagnetic (EM) pump development for circulating the liquid breeder, magnetohydrodynamic (MHD) experiment, and flow corrosion test, a PbLi breeder loop was constructed. From the performance test, the EM pump and magnet show their capability, and flow and static corrosion tests including oxide coating for corrosion protection were performed. For tritium extraction from the liquid breeder, a gas–liquid contact method was adopted and a tritium extraction chamber was constructed. For measurement of the tritium amount in the liquid breeder, permeation sensors have been developed.     

Fabrication of small mock-ups reflecting the design features of the ITER semi-prototype

The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. However, its fabrication is expected to face great difficulty due to a design complexity. Even though joining technology for different materials such as beryllium, CuCrZr, and stainless steel (SS) was developed during the first stage of qualification, the joining is still a key issue for the fabrication of the semi-prototype. In this study, small mock-ups (SMU) were fabricated to realize and verify the manufacturing of the semi-prototype reflecting the described design features. The joining of multiple beryllium tiles on the angled CuCrZr surface was confirmed with SMU#1. Six beryllium tiles were joined using hot isostatic pressing (HIP), and slitting was then performed to form multiple tiles. In SMU#2, HIP was performed two times in order to facilitate the cooling channels at the CuCrZr/SS interface, and to join the beryllium tiles on CuCrZr/SS. The method used to form a pressure boundary for the complex cooling channels was also developed by fabricating the SMU#3. The SMUs confirmed the applicability of the HIP for the manufacturing of the semi-prototype.     

Optimisation of hot isostatic pressing bonded SS/SS joints conditions for ITER blanket shield

In the engineering design activity of international thermonuclear experimental reactor (ITER), stainless steels are being considered as candidates materials for several module type structures. Hot isostatic pressing (HIP) technique is expected for the fabrication of these modules. Stainless steel powders are simultaneously consolidated as mono-material block or/and joined in bi-material module.This paper reviews the manufacturing stages, non-destructive examination and the developments of the HIP bonded joints of 316L SS (powder and solid) for application to the ITER shield blanket.It is well known that the powder surface oxidation negatively influences the impact toughness of raw material and joints consolidated by this way. In order to get acceptable mechanical properties of materials, a study on the effect of reducing the powder oxygen content has been launched. To evaluate susceptibility to the oxygen content of HIPed joint specimens, tensile and toughness tests have been performed.From this study, optimal conditions of HIP were fitted and the influence of oxygen was mastered to obtain good mechanical properties of the consolidated powder material as well as for HIPed junction.     

Solid state bonding of CuCrZr to 316L stainless steel for ITER applications

Dissimilar metal bonds between CuCrZr and 316L stainless steel were prepared using two different solid state joining techniques. In the first instance, hot isostatic pressing, a high temperature diffusion bonding process was used to join the copper alloy to the stainless steel substrate at temperatures near 1000 °C. In the second instance, explosion bonding at ambient temperature was employed. These two techniques both yielded mechanically robust joints, where the strength of the interface exceeded that of the copper alloy, the weaker of the two substrates. However, the two bonding techniques produced near-joint microstructures that were very different. The microstructure and mechanical performance of CuCrZr/316L stainless steel joints prepared via both techniques are compared. Microstructural analysis of the joints included scanning electron microscopy, electron microprobe analysis and Auger spectroscopy techniques. The bulk mechanical properties of the substrate alloys were very different as well and are described. Particular emphasis is placed on the residual mechanical properties of the CuCrZr after thermal processing that simulate beryllium tile bonding since once the Be tiles are in place, the copper alloy cannot be solutionized and age-hardened to return it to full strength.     

High heat flux test with HIP bonded Be/Cu/SS mock-ups for the ITER first wall

In order to verify the integrity of the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER), especially for preparing its qualification program by ITER-O, Be/Cu/SS mock-ups, which were the same size as the qualification mock-ups, were fabricated and tested at the TSEFEY, an e-beam facility, in Efremov, Russia. These mock-ups were joined with a 316 L austenitic stainless steel (SS316L) block for a structural material, CuCrZr for a heat sink material and SS316L tubes for a coolant and then, joined with three Be tiles for an armor material. A hot isostatic pressing (HIP) was used as manufacturing methods at a 1050 °C, 100 MPa for 2 h for a Cu/SS joining and at a 580 °C, 100 MPa for 2 h for a Be/(Cu/SS) joining. Two mock-ups were fabricated by using 1 μmCr/10 μmCu of an interlayer between the Be tile and Cu block. The high heat flux (HHF) tests were performed at 1.5 and 2.0 MW/m² heat fluxes for each mock-up. The given conditions and the expected fatigue lifetime were evaluated from a preliminary analysis with ANSYS. Both mock-ups survived for up to 1000 and 268 cycles at 1.5 and 2.0 MW/m² heat fluxes, respectively. They are higher than the expected numbers of cycles to a failure.     

Present status of beryllides for fusion and industrial applications in Japan

Beryllides have remarkable characteristics besides their low density, such as high radiation resistance, high chemical stability, low hydrogen isotope retention and good high temperature mechanical strength especially for fusion applications. In order to have a practical usage for industrial applications, it is important to compile a set of quantitative data on such properties. It is also important to develop fabrication and processing paths to assure the processing of inherently brittle beryllides.The characterization and manufacturing technologies development have been carried out mainly on the Be₁₂Ti interemetallic compounds at the Be-rich side of the Be-Ti binary system. In the present paper, up-dated results on several properties are described for the compound fabricated by hot isostatic pressing (HIP) and ingot metallurgy. Mechanical properties of the compounds having a duplex microstructure with neighboring phases are evaluated by compressive tests from room temperature to 1273 K. Radiation damage of the compound is preliminary studied by charged particle irradiation. Oxidation in air and the interaction with water vapor are evaluated. Thermal desorption of the deuterium is examined by using transmission electron microscopy (TEM) and thermal desorption spectrometry (TDS). Through these evaluations it seems that Be₁₂Ti is superior as neutron multiplier with respect to pure Be metal.     

Development of water-cooled solid breeder test blanket module in JAEA

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/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

Progress on design and R&D for helium-cooled ceramic breeder TBM in China

The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. Current progress on the design and R&D for Chinese helium-cooled ceramic breeder TBM (CN HCCB TBM) in China is presented. The main updated design and related R&D of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being carried out. Recent status of the components and fabrication technology development is also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CFL-1 are being prepared in the laboratory scale. The fabrication of 1/3 sized mock-up and construction of a He test loop are being carried out. The key technology development is proceeding to the large scale mock-up fabrication and demonstration tests toward on ITER testing.     

Design and performance analysis of structural components for a Korean He Cooled Ceramic Reflector TBM in ITER

Korea has developed a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) testing in ITER, which was considered one of the fusion DEMO-relevant blankets in Korea. The design and performance analysis of the TBM body have been carried out considering the uniqueness of the KO TBM and design requirements by the IO and KO design concept: (1) KO TBM has 4 sub-modules considering a post irradiation test (PIE) and its delivery. (2) A first wall (FW) design was changed into a 15 × 11 rectangular shape and its performance was confirmed by thermal-hydraulic and thermo-mechanical analyses using commercial ANSYS code. The results showed that the revised design model satisfied 1.5S_m and 3S_m of the allowable stress (S_m) in the RCC-MR code at the maximum stress region of the components for mechanical and thermo-mechanical analyses, respectively. (3) Considering the tritium breeding and cooling, a breeding zone (BZ) design was investigated. Three Li and Be layers, and one graphite layer, were proposed by the iteration, and the appropriate temperature distribution was obtained. The design for other components such as a side wall (SW) and back manifold (BM) is on-going considering 9 MPa of channel pressure and its functions of flow distribution as a manifold.     

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

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.     

New progress on design and R&D for solid breeder test blanket module in China

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced.The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.     

Development of fabrication procedure for Korean HCCR TBM

Korea has developed and plans to test a helium cooled ceramic reflector (HCCR) test blanket module (TBM) in the ITER. The HCCR TBM is composed of four sub-modules and a back manipulator (BM). Each sub-module is composed of a first wall (FW), breeding box, and side walls (SW). The fabrication procedure was developed to confirm the fabrication method for the HCCR TBM. The test specimens of the ARAA were prepared to test the weldability for tungsten inert gas (TIG) welding and electron beam (EB) welding. To establish and optimize the welding procedure in an EB weld from ARRA material, the variation in the bead width and penetration depth according to the welding current and welding speed were investigated. To verify the weldability and fabrication procedure for a complex structure such as the breeding zone, a small box with a cooling channel is being fabricated using the ARAA steel under development.     

Silicon doped carbon/Cu joints based on amorphous alloy brazing for first wall application

Amorphous ribbon-type filler-metals represent a promising selection for joining heterogeneous materials together. In this work, rapidly solidified ribbon-type Ti based amorphous filler with a melting temperature of 850 °C and a thickness up to 20 μm is used to join silicon doped carbon to pure copper. SEM examinations demonstrate that a high quality brazed joints could be acquired. The brazed seam has a uniform structure and pore free along its entire length. TiC and ZrC are formed near the interface of carbon and filler-metal when the brazing holds enough time. Using very thin Mo and Cu foil (0.2 mm in thickness) as multiple interlayer are very effective to mitigate the thermal stress occurred in the interface between carbon and copper. The shear strength of this carbon-multiple interlayer-copper joint is more than 20 MPa, and the rupture is mainly occurred on the carbon side.     

R&D progress of Korean HCSB TBM

Several R&Ds are being performed for Korean helium cooled solid breeder (HCSB) test blanket module (TBM) in the field of hydrogen isotopes permeation characteristics measurement in the helium purge line, joining technologies of structural materials, breeder pebble materials development, and the measurement of pebble bed characteristics. Electron beam welding for reduced activated ferritic–martensitic (RAFM) steel is evaluated to find optimal welding conditions. Also, a hydrogen permeation measurement apparatus is newly installed for the evaluation of the permeation barrier characteristics of stainless steel and RAFM steels. Two fabrication methods of lithium orthosilicate pebbles are investigated using slurry droplet methods. As methods of silicon carbide coating on the graphite pebble, microwave coating and chemical vapor deposition coating are evaluated. Two apparatuses are established to assess the thermo-mechanical properties of graphite and breeder pebble beds. The current status of R&D activities on these areas is introduced and the main progresses are addressed in this paper.     

Modeling and microstructural analysis of temperature effects in fused silica viewport

In ITER glass-metal joints are necessary to seal the optical windows, which are installed on the vacuum vessel, in order to perform diagnostics. These joints have to resist temperatures in a range 150-200 °C during normal operation and 250-300 °C during vessel out-gassing, while keeping a good mechanical strength.At the moment, there is a lack of data in literature about the behavior of glass-metal joints under ITER-relevant conditions. To obtain such data we studied the effect of temperature cycling on the microstructure of the glass-metal interface of a commercial brazed fused silica viewport. We have found that after thermal cycling already from RT to 100 °C cracks occurred in the interlayer (braze material) near the glass. With temperature increase the cracks became bigger and at 300 °C there was melting of one part of the interlayer, but the glass-metal joint sustained the treatment. To understand this behavior an elastic model of the viewport was developed using the Abaqus code. According to the model the temperature-induced maximum principal stresses in the interlayer exceeded the ultimate tensile strength of the braze alloy. The highest stresses in the interlayer were found near the glass, corresponding to the experimentally observed location of the cracks.