Small Mock-ups Research Articles

Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China

ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.

Development of fabrication technologies for the first wall of the HCCR blanket

The helium-cooled ceramic reflector (HCCR) concept is a primary candidate among demonstration fusion power reactor (DEMO) blanket options in Korea. The HCCR blanket is composed of a first wall (FW), a breeding zone, side cover plates and a back manifold. This study focuses on fabrication technologies for the FW using laser-beam welding (LBW) and hot isostatic pressure (HIP) bonding. Permanent backing strips in the pressure boundaries of the HCCR blanket are not allowed according to regulations related to pressure vessels and nuclear pressure vessels because high-pressure helium gas at 8 MPa is used as the coolant. LBW conditions and procedures for the fabrication of a FW without backing strips were investigated using small mock-ups with an advanced reduced-activation alloy called ARAA, which was developed in Korea. The laser beam power, thickness of the cover plate, and width of the backing strips were considered as parameters. The LBW conditions and procedures were optimized to remove the backing strip and achieve uniform and continuous back beads. After LBW, the fabrication of the FW with small mock-ups was completed by joining the back plate with HIP bonding. The small mock-ups were subject to very high temperatures exceeding 1000°C during the LBW and HIP bonding processes, which led to changes in the microstructures and hardness levels. The microstructures and hardness levels were recovered through post-heat treatments performed after HIP bonding. The microstructures of the HIP bonding interface after the heat treatments were observed. A partial HIP bonding line was confirmed and transverse tensile tests were carried out to investigate the impact of the partial HIP bonding line on the mechanical properties of the HIP bonding interface. The tensile strength of the HIP bonding interface was approx. 2% lower than that of the base metal, indicating that the partial HIP bonding line caused a slight decrease in the tensile strength, but the impact on the HCCR blanket design would be negligible.

Gas pressure casting technique for manufacturing of W/OFHC-Cu monoblock

The important engineering part of the divertor vertical target is the quality of joint between armor materials and heat sink. To reduce the joint interface stresses, a pure copper (Cu) interlayer was provided between the tungsten(W) armor and the CuCrZr cooling tube. Gas pressure casting method was developed to reduce the porosity and improve the mechanical properties of the Cu interlayer of W/Cu monoblock. In the vacuum casting sample without pressure, porosities were found in Cu. These porosities become defects at the interface between W and Cu and decrease the mechanical properties of the Cu interlayer. In the cooling process of casting, argon gas was pressurized into a porous tube inserted in the Cu and tungsten monoblock applying perpendicular force on the joint surface. This method was effective to remove the porosity in Cu generated during the cooling process and to increase the bonding strength between Cu and tungsten. The quality of bonding interface, depth of the Cu diffusion, and hardness of Cu were compared by SEM, EDS, and Vickers hardness measurement. In the samples made by gas pressure casting method, the porosity was decreased in Cu, no defects on the interface between tungsten and Cu, Cu was diffused into tungsten with a depth of 2∼3 μm and the hardness of Cu was increased about 15%.Small mock-ups were manufactured by hot radial pressing (HRP) with gas pressure casted W/Cu monoblock and CuCrZr cooling tube and underwent high heat flux (HHF) of 10 MW/m2 up to 5,000 cycles. The ultrasonic test (UT) results before and after the HHF test showed no defects on interfaces and was confirmed by observation of SEM on the cutting surface of the monoblock mock-up after the HHF test.

Hygrothermal performance of a massive natural stone masonry wall insulated from the internal side with hemp concrete – field measurements in cold climate

Improving of exterior walls in historic and traditional buildings is often only possible with interior thermal insulation. The actual structure and material properties of the existing exterior wall are usually the main unknown factors. Therefore, field measurements with small mock-ups are helpful before large-scale renovation. The current study analyses by field measurements the hygrothermal performance of internally insulated massive stone wall. Two different hemp concrete mixes were developed for the insulation. Temperature and humidity conditions were measured periodically over one year period. Results showed a very low drying rate of hemp concrete interior insulation. The external side of hemp concrete insulation will stay moist for a very long period. Temperature of coarse hemp concrete was slightly higher during the cold period. Wooden studs used to install hemp concrete will stay in moist areas for a long period. Temperature drop below 0 °C shows that interior insulation should be durable for freezing-thawing cycles. Drying out of constructional moisture is absolutely necessary for hygrothermal design. Before considering large-scale renovations, it is necessary to further assess the long term durability and performance of hemp concrete in a moist environment. The temperature increase on the interior surface could slightly improve indoor thermal comfort.

FW and Blanket Technology Development Progress at SWIP

Blanket, including its first wall (FW), is one of the key components in the fusion reactor, which works in the severe condition with the huge high energy neutron irradiation, high heat flux and electromagnetic load. Facing these technical challenges for the blanket and its FW, Southwestern Institute of Physics is engaging in the development of the blanket technology following the proposed China magnetic confinement fusion development strategy. This paper presents the design and R&D progress of the ITER FW and blanket, Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) for ITER and HCCB tritium blanket for China Fusion Engineering Test Reactor (CFETR). The design of the ITER enhanced heat flux FW has been updated to improve manufacturability and to reduce water leak risk. Key technologies have been qualified by high heat flux tests of small mock-ups and full-scale plasma-facing fingers. New assembly technologies are under development. Following the completion of full-scale prototype qualification, the ITER shielding blanket modules are under manufacturing for series production. The preliminary design of HCCB TBM for ITER is under implementation to improve the manufacturability and inspectability based on the fabrication study. Two prototypes of TBM submodule have been fabricated and tested to validate the feasibility of manufacture procedure, which are finally assembling into one semi-prototype of TBM. The preliminary design of HCCB tritium blanket for CFETR has been completed and the R&D has been started, especially for the special technologies different with the ITER blanket and TBM. Meanwhile, the progress of reduced activation ferritic/martensitic steel and function materials for blanket are also presented.

Analysis of hardness and microstructural changes in Tungsten mono-blocks exposed to high heat flux at 10 MW/m2

In this study, we investigated the material properties of tungsten mono-blocks undergone high heat flux test. The ITER-like tungsten mono-block specimens were fabricated by two kinds of joining methods. The tungsten-Cu interlayer joint was made by gas pressure casting and Cu interlayer-CuCrZr cooling tube were joined by hot radial pressing (HRP) or brazing method. To evaluate the integrity of the fabricated mono-blocks, high heat flux tests were done on the small mockups under cyclic heat load of 10 MW/m2 up to 5,000 cycles. We measured the microstructure, hardness, and surface state of the tungsten armors before and after the heat flux exposure experiment. The Vickers hardness profile measured from the surface down to the Cu interlayer showed a noticeable reduction around 200 μm ~ 500 μm depth after high heat flux test compared to that before the test depending on the bonding quality of mockup specimens. At depth beyond 800 μm, the Vickers hardness did not show any notable difference before and after the high heat flux test, which indicates the material property of tungsten is not so much affected in this region at the above heat flux condition. The microstructure and grain size distribution was measured and analyzed by SEM and EBSD. The correlation between the microstructure and hardness profile, and the impacts of the mockup performance under HHF test to the integrity of tungsten armors are analyzed.

Fabrication and testing of small mock-ups for helium cooled ceramic reflector breeding blanket development

Korea has developed a helium cooled ceramic reflector test blanket module (TBM) for fusion reactors. An advanced reduced activation alloy has been developed as a structural material and various joining methods such as E-beam welding, tungsten inert gas (TIG) welding, laser welding, and hot isostatic pressing have been applied to establish its fabrication method and procedure. Based on the current design of the TBM and breeding blanket, TIG welding was selected as the main joining method, and the U-shaped first wall (FW) was adopted. For welding of the breeding blanket, a special joining technique without a permanent backing strip with simultaneous full penetration welding in accordance with the design rules of class N2RX of RCC-MRx is required for the pressure vessel. For the U-shaped FW, distortion of the drilled cooling hole during the bending should be limited. Three small mock-ups were fabricated in this study to investigate these issues. The joints and bending parts were visually observed after cutting, and the hydrostatic pressure test was performed for regions with low thickness. The results show that the proposed fabrication methods are applicable in the fabrication procedure development for the breeding blanket.

Progress in developing ITER and DEMO first wall technologies at SWIP

The first wall (FW) is key component for ITER and the Chinese DEMO—the China Fusion Engineering Test Reactor (CFETR). It faces burning plasma and has a high heat flux (HHF) surface load. Critical issues and key technologies for manufacturing the ITER and CFETR FW are discussed with regards to thermal fatigue performance, Be/CuCrZr and W/reduced activation ferritic/martensitic (RAFM) steel bonding, material properties and failure mechanisms. Design improvement of the hypervapotron cooling channel and the joint interface structure was done, which increases the thermal fatigue lifetime of the enhanced heat flux (EHF) ITER FW by more than one order of magnitude as indicated by thermo-mechanical analysis. Small mock-ups and full-size EHF FW fingers were manufactured by qualified technologies in sequence. An HHF test of the small mock-ups showed that Be tile size and defects at the Be/CuCrZr interface have a great effect on the fatigue lifetime. Manufacturing tests showed a thick oxygen-free Cu interlayer could provide a good solution for the interface cracking issue. An ITER EHF FW semi-prototype was manufactured with additional two full-size finger pairs that successfully survived the demanding HHF test at 4.7 and 5.9 MW m−2. Various manufacturing technologies for joining W/RAFM steel for the CFETR FW have been studied, including a TiN coating at the interface as a tritium permeation barrier. Hot iso-static pressed W/RAFM steel joints showed a higher bonding strength than brazing joints but varied a lot with the interlayer metals. Further studies are required to optimize the technologies.

ENEA ultrasonic test on plasma facing units

ENEA has been working on improving the ultrasonic scanning technique of Plasma Facing Units (PFUs) components for many years, obtaining increasingly accurate and easy to interpret results. The first tests were performed on small mock-ups and, due to the geometry of the component, tests were performed from inside the tube. The first issue to solve was the reduced pipe thickness and the need to direct the ultrasound beam orthogonally to the tube surface. The choice of the correct probe characteristics and its particular shape solved this issue. Subsequently, the problem of a scan along a curved tube had to be addressed. Therefore, a special probe holder, able to follow the curvature placed at the end of a flexible shaft, was designed. The definition of more stringent acceptance criteria for the plasma facing components required to increase the scanning resolution. Defect detection and sizing were improved up to 0.5 mm by working on automatic system and data analysis software. The increasing of the sample sizes up to full-scale prototype for pre-production, led to the necessity to improve speed of the scanning without losing resolution. At this point, the scan along generatrix was discarded to pass to a helical path. In this work, the system created and patented in ENEA laboratories, for scanning full-scale Inner Vertical Target PFU prototypes of ITER, is illustrated. Test results in terms of resolution and defects sizing are shown.

AP&C opens second plasma atomized powder plant

ENEA and Ansaldo Nucleare S.p.A. are currently working on the fabrication of full scale prototypes of inner vertical target component for ITER (International Thermonuclear Experimental Reactor). The requirements is to manufacture a component by means of a qualified technology which uses armour and cooling pipe materials defined by ITER organization specification.A dedicated new Hot Radial Pressing facility was designed and finally installed in ENEA for the fabrication of the full scale Plasma Facing Units of the inner vertical target component prototype.HRP technology was widely tested with small mock-ups and medium qualification prototypes but its applicability to fabricate full scale units has to be anyway demonstrated and, therefore, their joint quality and also dimensions and shape has to be properly checked.The compliance of the joint quality with the acceptance criteria is then performed by ultrasonic water gap technique, using the qualified ENEA testing equipment. Dimensional check is performed as well by analyzing acquired data points compared to the reference drawings and 3D model.

Fabrication and acceptance of ITER vertical target divertor full scale plasma facing units fabricated by HRP

ENEA and Ansaldo Nucleare S.p.A. are currently working on the fabrication of full scale prototypes of inner vertical target component for ITER (International Thermonuclear Experimental Reactor). The requirements is to manufacture a component by means of a qualified technology which uses armour and cooling pipe materials defined by ITER organization specification.A dedicated new Hot Radial Pressing facility was designed and finally installed in ENEA for the fabrication of the full scale Plasma Facing Units of the inner vertical target component prototype.HRP technology was widely tested with small mock-ups and medium qualification prototypes but its applicability to fabricate full scale units has to be anyway demonstrated and, therefore, their joint quality and also dimensions and shape has to be properly checked.The compliance of the joint quality with the acceptance criteria is then performed by ultrasonic water gap technique, using the qualified ENEA testing equipment. Dimensional check is performed as well by analyzing acquired data points compared to the reference drawings and 3D model.

HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components.During normal operation the heat flux deposited on the bottom segment of divertor is 5–10MW/m2 but the capability to remove up to 20MW/m2 during transient events of 10s must also be demonstrated.In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP).The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances.The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

Fabrication of a 1/6-Scale Mock-Up for the Korea TBM First Wall in ITER

Korea has developed a liquid breeder blanket for the test blanket module (TBM) program in ITER with a helium-cooled molten lithium concept. Since ferritic martensitic steel is used as the structural material for the TBM first wall (FW), various joining methods have been developed with hot isostatic pressing in order to develop a TBM FW fabrication method. In this study, three small mock-ups were fabricated in order to develop and verify the manufacturing method of the TBM FW through the pressure and helium leak tests. They were successfully fabricated. After fabrication and checking the performance of the mock-ups, a 1/6-scale mock-up was fabricated with a 260-mm height, 444-mm width, and 435-mm depth, in which width and depth were preserved and the number of channels was reduced from 60 to 10. The mock-up has a U-type shape and ten channels with a size of 20-mm height and 10-mm width for cooling. A manifold for flow testing and high heat flux testing of the 1/6-scale mock-up was designed and fabricated to distribute fluid uniformly to the mock-up.

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.

Neutronics Experiments Using Small Partial Mockups of the ITER Test Blanket Module with a Solid Breeder

In order to evaluate the impacts of the incident neutron spectrum and the tungsten armor on the tritium production, integral experiments have been performed with small partial mockups relevant to the ITER test blanket module using DT neutrons at FNS of JAERI. The Monte Carlo calculation results for the integrated tritium productions agree well with the experimental data within 2 and 11 % for the mockups without the armor in the experiments without and with the neutron reflector, respectively. It is clarified that the tritium production can be very accurately predicted in the experiment without the reflector. In the mockups with the 12.6 and 25.2 mm thick tungsten armors, it is experimentally clarified that the integrated tritium productions are reduced by 3 and 6 % relative to the case without the armor, respectively.

Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated—selection of armour geometry and examination of tungsten–copper bonding techniques. Brazing and casting tungsten–copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20×5×10, 10×10×10, 20×20×10, 27×27×10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5–20) MW/m 2, the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing.

Analyses of divertor high heat-flux components on thermal and electromagnetic loads

Abstract In the International Thermonuclear Experimental Reactor (ITER), the divertor high heat-flux components are subjected not to only severe heat and particle loads, but also to large electromagnetic loads during reactor operation. A great deal of R&D has been carried out throughout the world with regard to the design of robust high heat-flux components. Based on R&D results, small and intermediate size mock-ups constructed from various armor tile materials have been successfully developed with respect to a thermomechanical point of view. However, little analysis has been carried out with regard to the elastic stresses induced with in the high heat-flux components via the electromagnetic loads during a plasma disruption. Furthermore, past research has only considered thermomechanical and electromagnetic loadings separately and uncoupled. Therefore, a systematic analysis of the combined effects of thermomechanical and electromagnetic loadings has been performed, with the analytical results assessed by ASME section 3 evaluation code.

Arterial blood pressure changes occur simultaneously with the EEG arousal response to an auditory stimulus

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.