Beryllium Armor Research Articles

High heat flux components with beryllium armour: From the small-scale mock-ups to the full-scale prototype of the ITER first wall panel

The Russian Federation shall supply 179 enhanced heat flux First Wall Panels (FWP) to the ITER construction site. These panels have about 30 types and consist of a steel supporting structure (FW Beam) with multilayer plasma-facing units (PFU) attached onto it. To start the production of the panels, the manufacturing readiness must be confirmed by the ITER Organization (IO). The JSC “NIIEFA” responsible for manufacturing of a part of the components, final assembly and testing of the panels has come a long way from the experimental work on small-scale mock-ups, through the development of the design and key manufacturing technologies to the production and testing of the full-scale prototype (FSP) which is the final step before the confirmation for the manufacture. The manufactured FSP has successfully passed most of the acceptance tests including hydraulic pressure and leak tightness tests. The high heat flux tests have been successfully carried out on the PFUs with beryllium armour. In parallel to the tests of the beryllium prototype, the tungsten armour has been analyzed and tested which is due to the IO's intention to change the armour material.

Microstructural analysis of the thermal fatigue tested Be/CuCrZr joint with artificial defects between Ti/Cu interlayers

The ITER enhanced-heat-flux (EHF) FW fingers undertook by China are made by hot isostatic pressing (HIP) method to bond beryllium armor tiles to CuCrZr/316L(N) heat sinks with 10 μm-thick Ti coating on beryllium tiles and 0.5 mm-thick oxygen free copper sheet as interlayer. To study its thermal fatigue behavior, small scale mock-up was manufactured with artificial defects (AD) on the beryllium tile surface with Ti coating. Thermal cycling of the mockup was carried out at 4.7 MW/m2 high heat flux (HHF) for 5000 cycles. The interface between Be/Ti/Cu and the fracture surface on beryllium tiles were observed by stereomicroscopy and scanning electron microscopy with energy spectrometer. The Be/Ti/Cu interfacial phase composition and the behaviors of defects after high heat flux test were analyzed. In addition, the generation and propagation mechanism of thermal fatigue defects were preliminarily studied in this paper. The results show that uniform bonding between Be and Cu was achieved and the interfacial phases are Be base/ Be-Ti intermediate/ α-Ti/ Ti2Cu/ TiCu /Cu base in the sequence. The formation of complex brittle phases reported in previous study, such as Ti3Cu4 or α-TiCu4, can be avoided by using Ti single interlayer. When the original defect dimension is only 3 × 3 mm2, the defect will not propagate after 5000 cycles 4.7 MW/m2 HHF testing (HHFT). Besides, decreasing beryllium tile size can reduce the possibility of thermal fatigue failure. In addition, from the research on fracture surface on beryllium tiles, the final fracture made for beryllium detachment mainly happened in Ti layer near the beryllium side. While thermal fatigue cracks preferentially originated at Be/Ti interface and the oxidation of beryllium surface under thermal cycling test would accelerate the generation and propagation of thermal fatigue defects.

Neutronic calculations of the China dual-functional lithium–lead test blanket module with the parallel discrete ordinates code Hydra

The China dual-functional lithium–lead test blanket module (DFLL-TBM) is a liquid LiPb blanket concept developed by the Institute of Nuclear Energy Safety Technology of the Chinese Academy of Sciences for testing in ITER to validate relevant tritium breeding and shielding technologies. In this study, neutronic calculations of DFLL-TBM were carried out using a massively parallel three-dimensional transport code, Hydra, with the Fusion Evaluated Nuclear Data Library/MG. Hydra was developed by the Nuclear Engineering Computational Physics Lab based on the discrete ordinates method and has been devoted to neutronic analysis and shielding evaluation for nuclear facilities. An in-house Monte Carlo code (MCX) was employed to verify the discretized calculation model used by Hydra for the DFLL-TBM calculations. The results showed two key aspects: (1) In most material zones, Hydra solutions are in good agreement with the reference MCX results within 1%, and the maximal relative difference of the neutron flux is merely 3%, demonstrating the correctness of the calculation model; (2) while the current DFLL-TBM design meets the operation shielding requirement of ITER for 4 years, it does not satisfy the tritium self-sufficiency requirement. Compared to the two-step approach, Hydra produces higher accuracies as it does not rely on the homogenization technique during the calculation process. The parallel efficiency tests of Hydra using the DFLL-TBM model also showed that this code maintains a high parallel efficiency on O(100) processors and, as a result, is able to significantly improve computing performance through parallelization. Parameter studies have been carried out by varying the thickness of the beryllium armor layer and the tritium breeding zone to understand the influence of the beryllium layer and breeding zone thickness on tritium breeding performance. This establishes a foundation for further improvement in the tritium production performance of DFLL-TBM.

Overview of manufacturing technologies under development in Russia for ITER plasma facing components

Manufacturing of the plasma-facing components (PFC) for the ITER is under qualification now (Merola et al., 2014) [1]. To start production of 179 first wall (FW) panels and 58 divertor domes (DD), for which the RF DA is responsible, it is necessary to manufacture their full-scale prototypes. In preparation for their manufacturing a wide range of technologies has been developed, which are continuously progressing (Mazul et al., 2011, 2012, 2016) [2–4]. Considering that changes were made in the DD design in view of the change-over to an all-tungsten divertor, as well as on the basis of the obtained results of manufacturing of the FW panel semi- prototype, the necessity arose to optimize some technological manufacturing stages. Thus, for the DD the composition of structural materials has been changed, the new assembly structure has been developed and analyzed; some technological operations have been optimized. The technological operations for FW panel have been also optimized, namely, the brazing temperature of the beryllium armor has been decreased, preparation works are under way on replacement of explosion welding by diffusion bonding for the steel-bronze joint. Most of the technological stages should be accompanied by non- destructive testing (ultrasonic, X-ray), the methods of which have been developed as well. The RF DA is also responsible for thermal testing of divertor elements. Due to application of the X-ray matrix the methods have been improved for recording of density of the applied heat flux produced by the electron beam. The ways for and the results of optimization of the technological and testing procedures are also presented below.

Thermal shock induced oxidation of beryllium

Beryllium has been chosen as a plasma facing material for the first wall of the experimental fusion reactor ITER, mainly because of its low atomic number and oxygen getter capabilities, which are favorable for a high plasma performance. While the steady state operational temperature of 250 °C has no deteriorating effect on the beryllium surface, transient plasma events can deposit power densities of up to 1 GW m−2 on the beryllium armor tiles. Previous research has shown that the oxidation of beryllium can occur under these thermal shock events. In the present study, S-65 grade beryllium specimens were exposed to 100 thermal shocks with an absorbed power density of 0.6 GW m−2 and a pulse duration of 1 ms, leading to a peak surface temperature of ∼800 °C. The induced surface morphology changes were compared to a steady state heated specimen at the same surface temperature with a holding time of 150 s. As a result, a pitting structure with an average pit diameter of ∼0.45 μm was observed on the thermal shock loaded surface, which was caused by beryllium oxide grain nucleation and subsequent erosion of the weakly bound beryllium oxide particles. In contrast, the steady state heated surface exhibited a more homogeneous beryllium oxide layer featuring small pits with diameters of tens of nm and showed the beryllium oxide grain nucleation in a beginning stage. The experiment demonstrated that thermal shock loading conditions can significantly accelerate the beryllium oxide grain nucleation. The resulting surface morphology change can potentially alter the fusion application relevant erosion, absorption, and retention characteristics of beryllium.

Mechanical and microstructural characterization of HIP joints of a simplified prototype of the ITER NHF First Wall Panel

The blanket of ITER protects the vacuum vessel from neutrons and other energetic particles produced in the fusion plasma. Each of the 215 Normal Heat Flux (NHF) panels of the blanket consists of a shield block and a First Wall (FW) panel. The NHF FW panels are a complex bimetallic structure of AISI 316L(N) stainless steel (SS) backing plate and a copper alloy (CuCrZr) heat sink, covered with beryllium armor tiles. Joining of these materials is done by solid state diffusion bonding. Under the framework of a R&D roadmap parallel to the manufacturing of a full-scale prototype of a FW panel of ITER, this work describes studies on the microstructure and strength of CuCrZr/SS and CuCrZr/CuCrZr joints of a simplified 10-fingers prototype of a FW panel manufactured by Hot Isostatic Pressing (HIP). Results on mechanical tests performed following ITER recommendations are compared to F4E specifications. Microstructural characterization of the interface was performed. Thermal history of the component is correlated with the mechanical behavior of the interfaces. Results show that appropriate parameters of the solution annealing after HIP and of the CuCrZr ageing during final HIP diffusion bonding are essential to achieve the specified strength of the joints.

High pulse number transient heat loads on beryllium

The experimental fusion reactor ITER will apply beryllium as first wall armor material. In present fusion experiments, e.g. ASDEX Upgrade, it has been detected that up to 25% of the plasma energy loss is deposited in non divertor regions during edge localized mode (ELM) events. Therefore, the impact of transient heating events on beryllium needs to be investigated to reliably predict the performance of the beryllium armor tiles under ITER operational conditions. In the present experiments, the electron beam facility JUDITH 2 was used to exert transient heat pulses with power densities of 0.14–1.0GWm−2, as they can be expected for mitigated Type 1 ELMs in ITER, pulse durations in the range of 0.08–1.0ms, and a number of pulses in the range of 103–107 on S-65 beryllium specimens that were brazed on an actively cooled copper structure. Thereby, a strong drop of the melting threshold was discovered from a heat flux factor FHF = 22–25MWm−2s0.5 for 102 pulses to FHF < 12MWm−2s0.5 for 103 pulses. However, a saturation of the thermally induced damage was observed for FHF ≤ 9MWm−2s0.5 after 105 pulses. This result indicated a promising performance of beryllium under a high number of transient heat pulses in ITER. Nevertheless, the synergistic effects between thermal loads, particle loads, and neutron irradiation might affect the saturation threshold and need to be investigated in future studies.

Activities of HIP joining of plasma-facing armors in the blanket first-wall in Korea

Joining technology for dissimilar materials was developed for the fabrication of an ITER blanket first-wall, which consisted of Be, CuCrZr, and stainless steel (SS). The Be/CuCrZr/SS joint was fabricated using a hot isostatic pressing (HIP) method. Beryllium armor was joined to the CuCrZr/SS block at 580°C under 100MPa. The optimal interlayer coatings of Cr/Cu and Ti/Cr/Cu were developed using an ion-beam assisted physical vapor deposition. Beryllium is also a candidate armor material for the TBM first-wall. Successful joining of Be to ferritic-martensitic steel (FMS) was accomplished using an HIP method by introducing the thick diffusion barrier. A thick diffusion barrier of a Cu foil(10μm) limited the excessive diffusion and prevented the formation of brittle phases at the Be/FMS interface. Be and FMS were bonded at 650–850°C; however, a temperature of lower than 750°C was recommended to avoid material degradation of FMS. In addition, the joining of W to FMS has been developed. Tungsten is another armor material applicable to more severe plasma conditions. The large difference in the thermal expansion between W and FMS was resolved by introducing the Ti interlayer and Mo separator. Moreover, the double-staged HIP (the first stage at 900°C and 100MPa and the second stage at 750°C and 70MPa) was applied to suppress the edge delamination of W/FMS joints during thermal history.

Impact of the surface quality on the thermal shock performance of beryllium armor tiles for first wall applications

Beryllium will be applied as first wall armor material in ITER. The armor has to sustain high steady state and transient power fluxes. For transient events like edge localized modes, these transient power fluxes rise up to 1.0GWm−2 with a duration of 0.5–0.75ms in the divertor region and a significant fraction of this power flux is deposited on the first wall as well. In the present work, the reference beryllium grade for the ITER first wall application S-65 was prepared with various surface conditions and subjected to transient power fluxes (thermal shocks) with ITER relevant loading parameters. After 1000 thermal shocks, a crucial destruction of the entire loaded area was observed and linked to the stress accelerated grain boundary oxidation (SAGBO)/dynamic embrittlement (DE) effect. Furthermore, the study revealed that the majority of the thermally induced cracks formed between 1 and 10 pulses and then grew wider and deeper with increasing pulse number. The surface quality did not influence the cracking behavior of beryllium in any detectable way. However, the polished surface demonstrated the highest resistance against the observed crucial destruction mechanism.

Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme.This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

Status of the beryllium tile bonding qualification activities for the manufacturing of the ITER first wall

The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

Final case for a stainless steel diagnostic first wall on ITER

In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline.

Manufacturing and Examination for ITER Blanket First Wall Small-Scale Mockups With KoHLT-EB in Korea

The ITER first wall (FW) includes beryllium armor tiles joined to a CuCrZr heat sink. The FWs are one of the critical components in an ITER machine with a surface heat flux of 4.7 MW/m2 or above. The small-scale mockup shall be a part of the qualification tests and used to validate the performance of the dominant manufacturing technologies before the production of larger scale components, and this mockup shall be equipped with a hypervapotron heat sink and manufacturing processes developed for a semiprototype design. The small-scale mockup includes 48 beryllium armor tiles (12 mm × 12 mm) capable of withstanding the specified heat flux values. The tile thickness shall be 6 mm to minimize the beryllium surface temperature and evaporation under high thermal loads. The detailed fabrication process of semiprototype small-scale mockups was developed for a qualification test in Korea. For the CuCrZr and stainless steel, the canned materials are processed into an hot isostatic pressing (HIP) device. In the case of beryllium-to-CuCrZr joining, the HIP was conducted at 580°C and 100 MPa. For nondestructive tests of the fabricated semiprototypes, visual and dimension inspections were performed whenever needed during the fabrication process, and ultrasonic tests were performed using ultrasonic probes. Destructive tests for the qualification semiprototype were performed on a small-scale mockup, which was fabricated together with semiprototypes. The Korea heat load test facility using an electron beam system was constructed with an electron gun (maximum electric power of 800 kW) for a high heat flux application with a 300-kW high-voltage power supply and maximum accelerating voltage of 60 kV. This facility was operated to evaluate the performance test of plasma facing components. A cyclic heat flux test will be performed to evaluate the ITER qualification program.

Preliminary Test and Evaluation of Nondestructive Examination for ITER First Wall Development in Korea

ITER first wall (FW) includes a beryllium armour joined to a Cu heat sink with a stainless steel back plate. These FW panels are one of the critical components in the ITER tokamak with a maximum surface heat flux of 5 MW/m2. Therefore, a qualification test needs to be performed with the goal to qualify the joining technologies required for the ITER FW. Various mock-ups were fabricated to develop the manufacturing procedure of FW components. For the nondestructive examination of the fabricated mock-ups, an ultrasonic test (UT) was performed with optimized probes. The UT was performed by using a three-axis digital ultrasonic C-scan system and software. The system comprised an ultrasonic pulser and receiver, model Panametrics 5800PR; a personal computer having an internal analog/digital converter board and four-axis motion control board; and a three-axis scanning tank. Two types of transducers were used for this experiment. One was Panametrics V312-SU, having a center frequency of 10 MHz (nominal) and a piezoelectric element diameter of 0.25 in with a flat protective layer for the Be/Cu. The other was Panametrics V309-SU with a center frequency of 5 MHz and an element diameter of 0.5 in for the Cu/SS interface. Winspect software controlled all aspects of data acquisition, motion control, data archiving, and image display. Based on the acceptance criteria, the average amplitude of the interface signals, which have about 50% of the reference echo amplitude, was recorded and analyzed on each beryllium tile. An image-analysis software analyzed the statistics of amplitude distribution and calculated the unacceptable area. Each mock-up that passed these UTs was concluded to qualify the joining technologies required for an ITER FW by using a high-heat flux test facility. As a result of these qualification tests based on the acceptance criteria of an ITER FW, the fabrication technologies will be utilized to develop the FW of plasma-facing components.

Effects of Hypervapotron Geometry on Thermalhydraulic Performance

Plasma disruptions and edge localized modes can result in transient heat fluxes as high as 5 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> on portions of a tokamak reactor first wall (FW). To accommodate these heat loads, the FW will likely use water-cooled hypervapotron heatsinks to enhance the heat transfer. In this article, we present the results of a computational fluid dynamics (CFD) study using 70 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">°</sup> C inlet water at 2.7 MPa to investigate the tooth height and backchannel depth of 50-mm-wide hypervapotrons with 6-mm-pitch and 3-mm side slots. We compare a popular design with 4-mm-high teeth and a 5-mm backchannel to a more optimal case with 2-mm-high teeth and a 3-mm backchannel under nominal heat loads (0.5 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> ) on a 100-mm-heated length and under single-phase flow conditions. Better heat transfer in the latter case and the smaller backchannel permit a factor of two reduction in the required mass flow while maintaining the same beryllium armor surface temperatures near 130 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">°</sup> C. The shallow teeth and smaller backchannel allow the 40 fingers in a typical panel to flow in parallel and simplify the water circuit. A comparison of the two hypervapotron designs during off-normal loading (5.0 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> ) and two-phase flow then follows. The design with 2-mm teeth has a 3.5% higher beryllium surface temperature of 648 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">°</sup> C and reduces the critical heat flux (CHF) by ~2%. Hypervapotron width also plays a role in heat transfer and CHF. CFD results for 36 and 70 mm wide hypervapotrons compared to the 50-mm case reveal similar thermal performance at low heat flux, but a reduction in CHF with increasing width. This study highlights the necessary compromise between design margin during transient events, effective heat transfer under nominal conditions, limitations on finger width, and the simplicity needed in the water circuit design.

Lifetime analysis of the ITER first wall under steady-state and off-normal loads**The views and opinions expressed in this paper do not necessarily reflect those of the ITER Organization.

The lifetime of the beryllium armor of the ITER first wall is evaluated for normal and off-normal operation. For the individual events considered, the lifetime spans between 930 and 35×106 discharges. The discrepancy between low and high estimates is caused by uncertainties about the behavior of the melt layer during off-normal events, variable plasma operation parameters and variability of the sputtering yields. These large uncertainties in beryllium armor loss estimates are a good example of the experimental nature of the ITER project and will not be truly resolved until ITER begins burning plasma operation.

High Heat Flux Test of the KO Standard Mockups for ITER First Wall Semi-Prototype

The Korean standard mockups with beryllium tile were fabricated to perform the high heat flux test for the qualification test of ITER blanket first wall. These mockups include the 80 mm × 80 mm beryllium armor tiles joined to the CuCrZr heat sink with stainless steel cooling tubes by HIP (Hot Isostatic Pressing) technology. The high heat flux tests were performed in the Korea heat load test facility (KoHLT-1) with the averaged surface heat flux of 1.25 MW/m2 by using a graphite heater. Preliminary thermal and mechanical analyses were carried out to simulate the test conditions and to determine the number of cycles for the fatigue lifetime of the mockups. In our KoHLT-1 facility, the normal heat cycle was based on an expected heat flux of 1.25 MW/m2, and each mockup had to endure the 1,000 normal heat cycles in this heat flux in accordance with the mechanical simulation. In the cyclic heat flux tests, the maximum surface temperature of the beryllium tiles was controlled below 400 °C. As a result of these high heat flux tests with the acceptance criteria of the ITER blanket first wall, the manufacturing technologies of the Korean standard mockups will be utilized to develop the tokamak blanket for the international qualification procedure.

Self-consistent analysis of the effect of runaway electrons on plasma facing components in ITER

Physical and computational models are developed, used and benchmarked for studying the response of ITER tokamak plasma facing components to runaway electron impact following a plasma disruption. The energy deposition, temperature evolution and material melting thickness are calculated for a wide range of runaway electron parameters, namely, electron kinetic energy, magnetic field, energy partition ratio (along and across magnetic field direction) impact duration, and wall material composition. It is shown that the electron energy partition ratio will have a significant effect on the wall heat load with melting of the first wall with beryllium armor possible. If tungsten armor is used instead, the surface of the mockup is overheated and melted for all ranges of studied parameters of the runaway electrons. Using an insert of a thin layer of a high-Z material inside the beryllium armor can mitigate the heat load in the armor and heat sink structure.

Thermal-Hydraulic LOCA Analysis of Chinese Helium-Cooled Solid Breeder Test Blanket Module for ITER

The Chinese helium-cooled solid breeder (CH-HCSB) test blanket module (TBM) is designed to be tested in ITER, and its aim is to validate the feasibility of a DEMO fusion reactor. The thermal-hydraulic transient analysis has to testify that the TBM and its helium cooling system (HCS) will not impact the safe operation of ITER under both normal and accidental conditions. In order to simulate the transient accidents, the TBM and HCS are modeled using the RELAP5/MOD3 system code. The steady-state results indicate that the designed TBM inlet/outlet temperatures are obtained and the temperature of first-wall (FW) structural material is below the limit. An ex-vessel loss-of-coolant accident (LOCA) will induce the melting of FW beryllium armor after ~80 s of LOCA initiation, and some controlling measures have to be taken before melting. The pressurization of the vacuum vessel induced by an in-vessel LOCA is within the allowable value of the ITER design. Because of pressurization of the purge gas system, the tritium extraction system has to be isolated from the TBM quickly when an in-box LOCA happens. Based on the results, the design of the CH-HCSB TBM could be further modified in order to assure the safety of the TBM and ITER, from an engineering point of view.

Fast brazing development for the joining of the beryllium armour layer for the ITER first wall panels

This paper describes the development of a flat pancake inductor for brazing beryllium tiles on a primary first wall (PFW) panel heat sink as well as the development of the brazing parameters. Two main variables were tested: brazing filler metal on one hand and PVD coating material on beryllium tiles on the other hand. The nickel coating and silver base brazing filler metal was the best couple and obtained shear values higher than 100 MPa. The inductor design and the choice of materials were applied to the brazing of a mock-up for high heat flux testing. Some tiles sustained 1000 cycles at 1.5 MW.