High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance.The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper.During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis.The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural material in the component. Also the behavior for Cu-OFHC interlayer material based on the experimental fatigue curves was considered and the ultimate tensile strength for W, because their failure affects the heat removal capability of the component.The good correlation found between FEA results and testing campaign validated again the use of FEA itself for future design improved concepts.     

Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear.Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.     

Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

After an extensive R&D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype.     

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.     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

Manufacturing of small-scale mock-ups and of a semi-prototype of the ITER Normal Heat Flux First Wall

This paper describes the manufacturing development and fabrication of reduced scale ITER First Wall (FW) mock-ups of the Normal Heat Flux (NHF) design, including a “semi-prototype” with a dimension of 305 mm × 660 mm, corresponding to about 1/6 of a full-scale panel. The activity was carried out in the framework of the pre-qualification of the European Domestic Agency (EU-DA or F4E) for the supply of the European share of the ITER First Wall. The hardware consists of three Upgraded (2 MW/m²) Normal Heat Flux (U-NHF) small-scale mock-ups, bearing 3 beryllium tiles each, and of one Semi-Prototype, representing six full-scale fingers and bearing a total of 84 beryllium tiles.The manufacturing process makes extensive use of Hot Isostatic Pressing, which was developed over more than a decade during ITER Engineering Design Activity phase. The main manufacturing steps for the semi-prototype are described, with special reference to the lessons learned and the implications impacting the future fabrication of the full-scale prototype and the series which consists of 218 panels plus spares.In addition, a “tile-size” mock-up was manufactured in order to assess the performance of larger tiles. The use of larger tiles would be highly beneficial since it would allow a significant reduction of the panel assembly time.     

Manufacturing and testing of a HETS module for DEMO divertor

The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (～15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.     

Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R&D area in fusion research. Pre-qualification tests for brazed joints between W–CuCrZr and C–CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m² uniform heat flux. Details about experimental and computational work are presented here.     

Feasibility study of an actively cooled tungsten divertor in Tore Supra for ITER technology testing

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.     

The ITER divertor pumping system,design evolution,simplification and performance

The ITER tokamak will be fuelled at a time averaged rate of up to 200 Pam³ s^?1 requiring neutralised gas in the divertor to be pumped to balance the fuelling and remove the fusion helium and other impurities in the exhaust. This is achieved on ITER using large bespoke cryo-sorption pumps. In this paper design evolution of the ITER divertor pumping system is outlined from the 1998 configuration to the current design. Details of the new, 6 direct pump, system design which will be used in the build of ITER are given. The operating modes of the new system configuration for different plasma scenarios are described and the performance of the new system is analysed and compared with previous baselines.     

Fabrication and characterization of tungsten and graphite based PFC for divertor target elements of ITER like tokamak application

The development of the fabrication technology of macro-brush configuration of tungsten (W) and carbon (graphite and CFC) plasma facing components (PFCs) for ITER like tokamak application is presented. The fabrication of qualified joint of PFC is a requirement for fusion tokamak. Vacuum brazing method has been employed for joining of W/CuCrZr and C/CuCrZr. Oxygen free high conductivity (OFHC) copper casting on W tiles was performed followed by machining, polishing and ultrasonic cleaning of the samples prior to vacuum brazing. The W/CuCrZr and graphite/CuCrZr based test mockups were vacuum brazed using silver free alloys. The mechanical shear and tensile strengths were evaluated for the W/CuCrZr and graphite/CuCrZr brazed joint samples. The micro-structural examination of the joints showed smooth interface. The details of fabrication and characterization procedure for macro-brush tungsten and carbon based PFC test mockups are presented.     

Activation analysis of coolant water in ITER blanket and divertor

Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. ¹⁶N and ¹⁷N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The ¹⁶N and ¹⁷N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced ¹⁷N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 10¹⁰ nuclide/cm³ and 1.05 × 10¹⁰ nuclide/cm³ of ¹⁶N, 8.93 × 10⁶ nuclide/cm³ and 0.33 × 10⁵ nuclide/cm³ of ¹⁷N. The decay gamma-rays from ¹⁶N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from ¹⁷N decay neutrons was much lower than that from fusion neutrons.     

R&D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

Japan Domestic Agency (JADA) carried out R&Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (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.     

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m² plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.     

ITER Magnet Feeder:Design,Manufacturing and Integration

The International Thermonuclear Experimental Reactor(ITER) feeder procurement is now well underway.The feeder design has been improved by the feeder teams at the ITER Organization(IO) and the Institute of Plasma Physics,Chinese Academy of Sciences(ASIPP)in the last 2 years along with analyses and qualification activities.The feeder design is being progressively finalized.In addition,the preparation of qualification and manufacturing are well scheduled at ASIPP.This paper mainly presents the design,the overview of manufacturing and the status of integration on the ITER magnet feeders.     

Preparation and properties of CVD-W coated W/Cu FGM mock-ups

Tungsten was coated on a W/Cu functionally graded material (FGM) by chemical vapor deposition technique (CVD), and then the tungsten coated tile was brazed on the CuCrZr heat sink with a cooling channel. The thickness of CVD-W was 2 mm deposited by a fast rate of about 0.7 mm/h. The features of the CVD-W coating including morphology, element composition and thermal properties were characterized. A tungsten coating with high density, purity and thermal conductivity is achieved. The bonding strength between the CVD-W layer and FGM was measured using bonding tensile tests. Thermal screening and fatigue tests were performed on the CVD-W mock-ups under fusion relevant conditions using an electron beam device. Experimental results showed that the CVD-W mock-up failed by melting of Cu beneath the tungsten layer under a high heat load of 14.5 MW/m² and 30 s pulse duration. Thermal fatigue tests showed that the CVD-W mock-up could sustain 1000 cycles at a heat load of 11.7 MW/m² absorbed power density and 15 s pulse duration without visible failure.     

Thermal analysis of new ITER FW and divertor design during VDE energy deposition

Further developments and investigations in the area of fusion energy devices require extensive and comprehensive computer simulations with great precision to evaluate reactor components behavior during normal and transient events. In this work we performed detailed study of the first wall (FW) subjected to high heat flux during a vertical displacement event (VDE) with various initial steady-state conditions and heat flux histories for the transient plasma energy deposition. We calculated the spatial temperature profile through out the entire module and the maximum surface temperature, as well as melting and vaporization thickness of Be surface during VDE and just before thermal quench (TQ). We further studied possible potential damage to plasma facing components (PFC) and structural materials for VDEs with higher energy loads than currently estimated.     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.     

Infrared thermography inspection of the ITER vertical target qualification prototypes manufactured by European industry using SATIR

High heat flux loaded components which will be installed in the ITER Divertor require a heat flux removal capability in the range 5–10 MW/m² at steady-state and up to 20 MW/m² in transients. Within the ITER plasma facing components procurement context, each party should demonstrate its technical capability to carry out the manufacturing with the required quality. This is achieved through the successful manufacturing and testing of medium-size qualification prototypes. Each Qualification Prototype consists of three high heat flux units mounted onto an actively cooled supporting structure. Currently, the SATIR method has been identified by the ITER Organization as the basic test to decide upon the final acceptance of the ITER Divertor components. SATIR testing was performed on each CFC part of European HHF units prior to the insertion of the twisted tape and prior to assembling the units onto the steel support structure. The paper deals with SATIR results of all qualification prototypes manufactured by European industry.