Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m² anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m² unscathed.     

Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m^?2, advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material.     

Fatigue lifetime of repaired high heat flux components for ITER divertor

Thermal fatigue behaviour of repaired monoblocks was assessed from High Heat Flux (HHF) tests up to 20 MW m^?2 on 11 components. Among these components, 8 monoblocks were repaired (2 CFC and 6 tungsten). These components were manufactured by two EU industries: ANSALDO Ricerche and PLANSEE. Non destructive examination was performed on SATIR thermography test bed before and after HHF tests. SATIR results show that repaired monoblocks have a good thermal exhaust capability before HHF tests. For all monoblocks, no degradation of thermal properties was noticed during cycles at 10 MW m^?2. After hundreds of cycles at 20 MW m^?2, two W repaired monoblock melted. Post-HHF SATIR examination revealed a degradation of thermal properties which is systematic for W melted monoblocks and non-systematic for W repaired ones. For CFC repaired monoblocks, no damage was observed up to 20 MW m^?2. For the first ITER divertor set, specifications for the pre-qualification are that CFC (Resp. W) components have to sustain in steady state 1000 cycles at 10 MW m^?2 (Resp. 3 MW m^?2) followed by 1000 cycles at 20 MW m^?2 (Resp. 5 MW m^?2). For the first ITER divertor set, the repair process is validated for CFC and W monoblocks.     

Powder Injection Molding – An innovative manufacturing method for He-cooled DEMO divertor components

At Karlsruhe Institute of Technology (KIT), a He-cooled divertor design for future fusion power plants has been developed. This concept is based on the use of modular cooling fingers made from tungsten and tungsten alloy, which are presently considered the most promising divertor materials to withstand the specific heat load of 10 MW/m². Since a large number of the finger modules (n > 250,000) are needed for the whole reactor, developing a mass-oriented manufacturing method is indispensable. In this regard, an innovative manufacturing technology, Powder Injection Molding (PIM), has been adapted to W processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining.The complete technological PIM process for tungsten materials and its application on manufacturing of real divertor components, including the design of a new PIM tool is outlined and, results of the examination of the finished product after heat-treatment are discussed. A binary tungsten powder feedstock with a solid load of 50 vol.% was developed and successfully tested in molding experiments. After design, simulation and manufacturing of a new PIM tool, real divertor parts are produced. After heat-treatment (pre-sintering and HIP) the successful finished samples showed a sintered density of approximately 99%, a hardness of 457 HV0.1, a grain size of approximately 5 μm and a microstructure without cracks and porosity.     

Research and development status on fusion DEMO reactor design under the Broader Approach

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with “pulsed” DEMO having a major radius (R_p) of 9 m, and “steady state” DEMO of R_p = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress.     

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested.A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process.The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France).The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m², 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m², 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m² Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign.This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up.These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

Evaluation of energy and particle impact on the plasma facing components in DEMO

We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.     

Current status of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO is being investigated at the Forschungszentrum Karlsruhe within the framework of the EU power plant conceptual study. The design goal is to resist a heat flux of 10 MW/m² at least. The major R&D areas are design, analyses, fabrication technology, and experimental design verification. A modular design is preferred for thermal stress reduction. The HEMJ (He-cooled modular divertor with multiple-jet cooling) was chosen as reference concept. It employs small tiles made of tungsten, which are brazed to a thimble made of tungsten alloy W-1%La₂O₃. The W finger units are connected to the main structure of ODS Eurofer steel by means of a copper casting with mechanical interlock. The divertor modules are cooled by helium jets (10 MPa, 600 °C) impinging onto the heated inner surface of the thimble.In cooperation with the Efremov Institute a combined helium loop & electron beam facility (60 kW, 27 keV) was built in St. Petersburg, Russia, for experimental verification of the design. It enables mock-up testing at a nominal helium inlet temperature of 600 °C, an internal pressure of 10 MPa, and a pressure difference in the mock-up of up to 0.5 MPa. Technological studies were performed on manufacturing of the W finger mock-ups. Several high heat flux tests were successfully performed till now. Post-examination and characterisation of the mock-ups subjected to the high heat flux tests were performed in collaboration with Forschungszentrum Jülich. Altogether, the test results confirm the divertor performance required. The helium-cooled divertor concept was demonstrated to be feasible. The knowledge gained from these experiments and some aspects on the design improvement are discussed in this contribution.     

Design and concept validation of the new solid tungsten divertor for ASDEX Upgrade

Div-III, a divertor with solid tungsten target tiles for ASDEX Upgrade is designed and tested and will be installed in 2013. It is a further step in exploring tungsten as material for plasma facing components. It avoids the restrictions of tungsten coatings on graphite and realizes an operation range up to 50 MJ energy removing capability in the outer divertor. In addition, it allows physics investigation such as erosion and deuterium retention as well as effects of castellation and target tilting. The design of the target itself and the attachment was optimized with FE-analysis and was intensively high heat tested up to a double overload. Cyclic tests reveal that the target and the attachment can be operated with the design load of 50 MJ without any damage. Even a twofold overload results in local recrystallization and minor cracks but the targets did not fail during operation. The redesign of the divertor structure was used to increase the conductance between the cryo-pump and the divertor region. The impact of the changed pumping efficiency was investigated with SOLPS/Eirene modeling. The modeling results are an indication for an easier access to lower SOL densities as expected for a higher pumping efficiency in the main chamber.     

Preliminary concept design of the divertor remote handling system for DEMO power plant

This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes.This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel.Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate.The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.     

Alternative electro-chemically based processing routes for joining of plasma facing components

Tungsten is considered in fusion technology as functional and structural material in the area of blanket and divertor for future application in DEMO. The KIT design of a He-cooled divertor includes joints between W and W-alloys as well as of W with Eurofer-steel. The main challenges range from expansion mismatch problem for tungsten/steel joints over metallurgical reactions with brittle phase formation to crack stopping ability and excellent surface wetting. These requirements were only met partly and insufficiently in the past e.g. by direct Cu-casting of tungsten onto steel.Both, the joining needs and the observed failure scenarios of conventionally joined components initiated the development of improved joining technologies based on electro-chemical processing routes. As electrolytes aqueous and aprotic, water free, system are integrated into this development line. In the first step principle requirements are presented to guarantee a reproducible and adherent deposition of scales based on Ni and Cu acting as inter layers and filler, respectively, to generate a real metallurgical bonding as demonstrate by 1100 °C joining tests. The development field aprotic systems based on ionic liquids is discussed with respect to enable development of refractory metal based fillers with focus high temperature W–W brazing.     

The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m² range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.     

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.     

Waste management strategy focused on maintenance,storage and recycling

Waste is generated at the moment when the operation of a fusion reactor is halted and maintenance is started for periodic replacement of blanket modules and divertor. Used blanket and divertor need to be replaced shortly after the shutdown for high plant availability, as long as high surface dose rate and decay heat of the blanket and divertor can be handled. In this sense, nuclear characteristics of the blanket and divertor need to be understood for a reasonable maintenance scheme. For the purpose, neutronic calculations were carried out on the blanket and divertor using a THIDA-2 code with FENDL-2.0. For a SlimCS DEMO reactor, the calculated decay heat for each 1/12-sector was as high as 5 MW just after the shutdown and 0.3 MW one month later. For the maintenance, a cooled shielding structure (CSS) was proposed to remove the decay heat and to shield gamma-rays from the sector. When maintenance is done one month after the shutdown, the sector temperature is maintained to be 550 °C or lower with the cooling by the CSS of 50 °C. In order to avoid tritium release from the sector during the maintenance, a cask should be used to transport the sector. For efficient use of resources, breeding and neutron multiplying materials should be reused or recycled. A possible strategy for reuse or recycle is also presented.     

Engineering evolution of the FAST machine

FAST (Fusion Advanced Studies Torus) is a proposal for a Satellite Facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploit innovative plasma facing component systems for DEMO. FAST is a compact (Ro = 1.82 m, a = 0.64 m, triangularity δ = 0.4) and cost effective machine able to investigate, with integration capability, non linear dynamics effects of alpha particle behaviour in burning plasmas. FAST operates in high performance H-mode (B_T up to 8.5 T; I_P up to 8 MA), as well as in advanced tokamak mode (I_P = 3 MA), and in full non inductive current mode (I_P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets. This allows for a pulse duration up to 170 s at 3 MA/3.5 T. The vacuum vessel (VV), segmented into 20-degree modules, is capable to accommodate a 40 MW RF power system. The machine has been designed to house a 10 MW Negative Neutral Beam Injection (NNBI) system. Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plate materials, and argon or neon as the impurities to be injected for mitigating the thermal loads.     

Simulations of the SOL plasma for FAST,a proposed ITER satellite tokamak

The first simulations with EDGE2D/EIRENE code of the SOL plasma in the FAST tokamak have been run for the basic H-mode scenario. Its similarity to ITER and relevance for DEMO bring interest to the study. Five different preliminary divertor designs have been examined by varying density at separatrix over the plausible range n_s,out = 0.7–1.0 × 10²⁰ m^?3. Margins exist for optimizing the design and minimizing the impurity injection rate even at the lowest density, with load below the safe limit of 18 MW/m² on the monoblock W targets, and to achieve a good degree of detachment at higher density. Both the plate tilting angle and the neutral dynamics are crucial factors. The detachment level can be significantly increased for the higher density scenario, while for the full non-inductive operation the injection of impurities will probably be necessary to reduce the heat load.     

Manufacturing,testing and post-test examination of ITER divertor vertical target W small scale mock-ups

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets.In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes.According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m² and finally 1000 cycles at 20 MW/m².After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing.A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area.The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks.This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.     

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.