In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration.In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel.In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed.

For the upcoming long pulse operation phase 2 "OP2" of Wendelstein 7-X (W7-X), new water cooled non planar stainless steel panels have been manufactured to protect the wall of the plasma vessel behind the divertor pumping gap. Such a panel is made of a machined ground plate with channels which are covered by likewise machined sheets. The latter are electron-beam welded to the ground plates. There are 60 panels of 7 different types which are designed to remove a stationary heat flux of up to 100 kW/m(2). The specified water cooling conditions are: 2.5 MPa inlet pressure, 30 degrees C inlet temperature, and 0.27 l/s flow rate. A panel prototype has been manufactured to validate the design and manufacturing process. In order to verify the thermo-hydraulic calculations performed with ANSYS (R) CFX, a thermal loading test on the prototype was carried out in the SIR HEX test facility at KIT. The loaded surface of the prototype (about 0.11 m(2)) was black-finished, and thermocouples were installed at different positions. The prototype was placed in the vacuum tank and heated by an array of six infrared lamps (400 V, max. 16 kW per lamp). The deposited load of up to about 100 kW/m(2) was measured by calorimetry. The paper describes the loading experiment of the panel, and its results. A good agreement was found between calculations and test results, and thus the thermo-hydraulic panel design was well validated.

In future nuclear fusion reactors, plasma facing components have to sustain specific neutron damage. While the majority of irradiation data provides a relatively clear picture of the displacement damage, the effect of helium transmutation is not yet explored in detail. Nevertheless, available results from simulation experiments indicate that 9%-chromium steels will reach their operating limit as soon as the growing helium bubbles extent a critical size. At that point, the material would most probably fail due to grain boundary embrittlement. In this contribution, we present a strategy for the mitigation of the before-mentioned problem using the following facts. (1) The neutron dose and related transmutation rate decreases quickly inside the first wall of the breeding blankets, that is, only a plasma-near area is extremely loaded. (2) Nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium, which would suppress the formation of large helium bubbles for a much longer period. (3) Compared to conventional steels, ODS steels also provide improved irradiation tensile ductility and creep strength. Therefore, a design, based on the fabrication of the plasma facing and highly neutron and heat loaded parts of blankets by an ODS steel, while using EUROFER97 for everything else, would extend the operating time and enable a higher heat flux. Consequently, we (i) developed and produced 14%Cr ferritic ODS steel plates and (ii) optimized and demonstrated a scalable industrial production route. (iii) We fabricated a mock-up with five cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. (iv) Finally, we performed high heat flux tests in the Helium Loop Karlsruhe, applying a few hundred short and a few 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e. 550 °C) was finally exceeded by 100 K. (v) Thereafter, microstructure and defect analyses did not reveal critical defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. However, a solution to prohibit the formation of such heat affected zones is given. These research contributions demonstrate that the use of ODS steel is not only feasible and affordable but could make a decisive difference in the future design and performance of breeding blankets.

A major part in the EUROfusion materials research program is dedicated to characterize and quantify nuclear fusion specific neutron damage in structural materials. While the majority of irradiation data gives a relatively clear view on the displacement damage, the effect of transmutation - i.e. especially hydrogen and helium production in steels - is not yet explored very well. However, few available results indicate that EUROFER-type steels will reach their operating limit as soon as the formation of helium bubbles reaches a critical amount or size. At that point, the material would fail due to embrittlement at the considered load. This paper presents a strategy for the mitigation of the before-mentioned problem using the following facts: the neutron dose and related transmutation rate decreases quickly inside the first wall, that is, only a plasma-near area is extremely loaded nanostructured oxide dispersion strengthened (ODS) steels may have an enormous trapping effect on helium and hydrogen, which would suppress the formation of large helium bubbles compared to conventional steels, ODS steels show improved irradiation tensile ductility and creep strength In summary, producing the plasma facing, highly neutron and heat loaded part of blankets by an ODS steel, while using EUROFER97 for everything else, would allow a higher heat flux as well as a longer operating period. Consequently, we (1) developed and produced 14 % Cr ferritic ODS steel plates. (2) We fabricated a mockup with 5 cooling channels and a plated first wall of ODS steel, using the same production processes as for a real component. And finally, (3) we performed high heat flux tests in the HELOKA facility (Helium Loop Karlsruhe at KIT) applying short and up to 2 h long pulses, in which the operating temperature limit for EUROFER97 (i.e., 550 degrees C) was finally exceeded by 100 K. Thereafter, microstructure and defect analyses did not reveal defects or recognizable damage. Only a heat affected zone in the EUROFER/ODS steel interface could be detected. This demonstrates that the use of ODS steel could make a decisive difference in the future design and performance of breeding blankets.

The Lead-Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) being developed by India for testing in ITER adopts various well developed design engineering and manufacturing technologies. The First Wall (FW), which is directly exposed to the incident heat flux, is designed for high pressure helium flow, high operating temperature (up to 100 bar and 550 degrees C) and considerable thermal stress and electromagnetic disruption loads. In order to check the thermal performance of the FW and ensure its structural integrity, a mock-up of the FW fabricated in India was tested in HELOKA test facility at KIT, Germany. Both normal and accidental operating conditions were investigated under ITER-like surface heat fluxes. Based on these results the thermal performance of the FW were validated.

The European Union Demonstration Fusion Power Reactor (EU DEMO) is facing its preconceptual design phase. In this phase, the research and development activities make extensive use of computational tools, to, e.g., verify the design calculations or to perform parametric analyses aimed at optimization. The design of the breeding blanket (BB), which will be a first-of-a-kind component in EU DEMO, is supported from the thermal-hydraulic point of view by local three-dimensional (3-D) computational fluid dynamics (CFD) analyses, mainly aimed at verifying the heat removal capabilities of the system, and by analyses at the system level using one-dimensional (1-D) codes. This work presents the development and application of a detailed 1-D model of the coolant manifolds for the helium-cooled pebble bed BB concept for EU DEMO. This model, implemented in the GEneral Tokamak THErmal-hydraulic Model (GETTHEM), allows fast analyses to be performed at the global level but still maintain a good level of detail concerning the coolant distribution. The first results obtained with the model prove that 3-D CFD analyses of the manifolds may provide misleading results due to nonrepresentative boundary conditions (BCs), which must be used to avoid having a domain that is too complex. The application of a global model, which is indeed characterized exploiting local analyses, can in turn provide better BCs to the detailed 3-D CFD analyses.

The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS, focusing in this phase on the design life cycle and on R&D activities in support of the design. The TBS ancillary systems are mainly circuits devoted to the removal of thermal power and to the extraction and recovery of the tritium generated in the Test Blanket Modules. They are: The Helium Cooling System (HCS); The Coolant Purification System (CPS); The Tritium Extraction System (HCLL-TRS, HCPB-TES); The Lead Lithium Loop. Their conceptual design was deeply analyzed during the ITER Conceptual Design Review (CDR) in 2015. The assessment of the CDR was overall positive and recommendations for improvements were made. The present design takes into account the main CDR recommendations as well as the implementation of the requirements related to ITER operation, safety principles application and physical space constrains.

The power exhaust from high temperature plasma at the level of divertor is one of the biggest technological challenges in the path to establish fusion power reactors. At the Karlsruhe Institute of Technology (KIT), a novel helium-cooled divertor using jet impingement cooling array has been investigated as one of the possible solutions for a DEMO tokamak. The concept, consisting of tungsten armour slabs fixed on a tungsten-laminate tube inside which a jet distribution cartridge is placed, has undergone an extensive optimization process. Among other things, the heat transfer coefficient (HTC) and pressure drop over a divertor unit have been thoroughly looked upon in order to improve the divertor thermohydraulic performance. The present paper is presenting such an optimization study in which the effect on the divertor performances of the distance between the cartridge and the heat loaded surface is investigated. Given the geometrical constraints, the nozzle-to-target surface distance is changed either by increasing the diameter of the inlet manifold, or, by varying the cartridge position, starting from a coaxial configuration, to various positions closer to the impingement surface.

The development of nuclear fusion, which is an environmentally friendly and future-oriented power generation method, needs to be developed for the prevention of global warming caused by the excessive use of fossil fuels. However, there are many scientific and technical challenges that should be overcome for the use of nuclear fusion reactors. The development of a divertor is one such challenge, in that it should be able to endure a high heat flux (similar to 10 MW/m(2)) and a particle flux (similar to 10(24)/s). To solve this challenge, the Karlsruhe Institute Technology (KIT) has been developing a helium-cooled divertor with a multi-array impingement jet. The main objective of developing this divertor module is not only to sufficiently dissipate the high thermal load from plasma for a reliable operation, but also to maintain a moderate pressure drop. To achieve this objective, we should understand the heat transfer characteristics of the divertor module and enhance the thermal performance, namely, the thermal conductance and pressure drop. The present study is aimed at an enhancement of the cooling performance of a helium-cooled divertor through the addition of rib structures in the divertor module. The results show that the convective conductance with rib turbulators is increased through an increase in the interfacial area on a heated surface. In addition, the thermal performance of the divertor module is enhanced by 20% or more compared with a no rib case.

The First Wall (FW) of the EU Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) faces the fusion plasma and experiences high heat fluxes; therefore its cooling channels design is a key R&D task for qualifying the HCPB TBM for the fusion reactors ITER and DEMO. Within the manufacturing and qualification activities performed in KIT for the HCPB TBM, a First Wall Mock-up (FWM) was designed and manufactured. The objective of this study is to characterize the hydrodynamic behaviour of the FWM cooling channels by investigating the coolant pressure drop for each channel. The FWM has a shape of rectangular prism (710 mm x 405 mm x 45 mm) with 10 U-shaped cooling channels which have a square cross section (15 mm x 15 mm) with rounded corners of 4 mm radius. The FWM was integrated into a gas loop that has the relevant instruments and piping system for measuring and controlling the air (6 bar abs.) flow parameters. This paper presents experimental measurements of the pressure drop across the FWM cooling channels. The present results support the qualification of the HCPB TBM mock-ups which will be tested in the Helium Loop Karlsruhe (HELOKA) facility with fusion-relevant heat flux and helium cooling. (C) 2017 Elsevier B.V. All rights reserved.

An experimental program based on a First Wall mock-up is presented as preparation for the qualification of breeding blanket mock-ups at high heat flux in the Helium Loop Karlsruhe (HELOKA) facility. Two objectives of the experimental program have been defined: testing of the experimental setup and a first validation of FE models. The design and manufacturing of mock-up representing about 1/3 of the heated zone of an ITER Test Blanket Module (TBM) First Wall is discussed. A modular attachment system concept has been developed for the fixation of the mock-up in order to be able to generate different stress distributions and levels on the plate, which is confirmed by thermo-mechanical analyses. The HELOKA facility is able to provide a TBM relevant helium cooling system and to generate the required surface heat flux by an electron beam gun. An installed IR camera can be used to measure the temperature distribution on the surface. (C) 2016 Published by Elsevier B.V.

In view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES–TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase.

The first 3D thermal-fluid-dynamic and structural analyses done for the design and pre-test assessment of the so-called Thermo-Cycle Mock-up (TCM), reproducing about 0.3 m(2) of a flat first wall (FW) with relevant helium cooling channels, are presented, based also on previous computational and experimental activities conducted at KIT but limited so far to a single cooling channel with straight heated length. The TCM is the first of a series of FW mock-ups presently under construction, to be tested starting from 2015 in the large HELOKA facility at KIT. Here, the fluid dynamics in the 180 turns of the TCM cooling channels is investigated together with the effects of heat transfer between neighboring channels, when the plate is subject to steady-state heat fluxes in the range 0.3-0.5 MW/m(2). Based on the computed temperature maps, the stresses in the TCM and the related damage figures for the main failure modes (Le., ratcheting and creep/fatigue) are assessed. These are compared with allowable limits in code and standards for the qualification of the TCM design and related to the prediction of the behavior of the component in the actual fusion environment.

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase. (C) 2015 Published by Elsevier B.V.

In this article a new helium cooled test facility is presented. The loop, KATHELO, is designed to operate at pressures up to 10MPa and temperatures up to 800°C in order to be able to provide suitable testing conditions for the helium cooled divertor concept developed at KIT. The general layout of the loop is introduced and the some of the technological solutions adopted for the loop components are discussed. The thermal-dynamic behavior of the loop during start-up and steady state operation for two different scenarios is analyzed using a RELAP-3D model of the circuit. Based on these simulations the heating power needed for high temperature operation is estimated. A particular attention is given to the thermal coupling between the cold and hot leg of the loop with emphasis on overall loop efficiency.

In the frame of the activities of the EU Breeder Blanket Programme, the Karlsruhe Institute of Technology (KIT) is involved in several activities in support of the design and qualification of the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM). During operation the TBM will be subjected to complex thermo-mechanical loads, especially the plasma facing parts, the so-called First Wall (FW). In the recent years significant effort was dedicated to the FE modeling of the TBM box as a complete assembly: detailed FE models (I quarter) were developed and thermo-mechanical calculations were performed for the better description of the thermo-mechanical behavior of the HCPB-TBM box under nominal and accidental conditions. In parallel with the numerical studies the manufacturing process for the TBM-FW has progressed. Currently KIT has developed a manufacturing path that will allow the manufacturing of an FW plate, the so-called TCM (Thermo-mechanical Cycle Mock up), with an overall size of 710 mm x 484 mm x 30 mm. The present paper presents the results of the numerical investigations of a mock up designed to qualify the manufacturing procedure of the TCM. For this the TCM will be subject to the same heat loads as the central part of a TBM-FW operating under TBM characteristic conditions (nominal surface heat flux of 500 kW/m(2) in pulsed regime). Based on the comparison with the numerical simulations of a full TBM box, a design of the test mock up including the support structure behind the TCM plate will be proposed with the aim of reproducing an equivalent level of stresses in the test object as in a TBM-FW. (C) 2013 Elsevier B.V. All rights reserved.

The design of the auxiliary systems for the two European Test Blanket Modules is an important engineering task that will allow the successful integration of the two DEMO Blanket mock-ups into the ITER machine, and, through these experiments, demonstrate the technological feasibility of a blanket module for a future fusion power plant.These auxiliary systems are mainly circuits devoted to the removal of thermal power and tritium recovery from the blanket modules, as follows:The Helium Cooling System (HCS):The Coolant Purification System (CPS);The Tritium Extraction System, divided in two subsystems in case of Helium Cooled Lithium Lead (HCLL), the Tritium Extraction Unit and the Tritium Removal System;The Lead Lithium Loop (HCLL).Starting from the preliminary design developed in the past, the optimization and the engineering design of these subsystems was performed in the first year of activity of the TBM Consortium of Associates. The input parameters were reviewed and, when possible, DEMO relevant solutions were proposed. The present design takes in account both the TBM operational requirements as well the ITER operation, safety, and spatial constrains.In this paper the current loops design with accent on the integration into the ITER machine are presented. (C) 2011 Elsevier B.V. All rights reserved.

This paper presents the recent results of the biggest removable component of the European ITER test blanket system (TBS), the ancillary equipment unit (AEU) development. The subsystem components are located inside the AEU, which ensures quick and reliable operation, maintenance and transport of these port cell components.The initiative concept of AEU frame structure was very similar to the preliminary design of the transfer cask but the self weight was already too high. Hence a significant structure optimization work, based on FEM analyses, has been implemented. The concept of support structure has been changed, and the weight appreciably has been decreased. In the next step the layout of subsystem components and routing of pipes have been developed taking into account the maintainability requirements of components and an additional function of AEU. It will be also the basic support of remote handling equipment, which will be deployed to the port interspace for pipes connection and disconnection operations. (C) 2011 Elsevier B.V. All rights reserved.

HELOKA experimental facility, aimed to test the test blanket modules (TBM) mock-ups and prototypes, is under construction at the Research Centre Karlsruhe. HELOKA is an “8-shape” helium loop (HL), which feeds the TBM test section with helium (1.4kg/s, 300°C, 8MPa). The maximum temperature at the test section outlet is 500°C. HELOKA data acquisition and control system (DACS) comprises the facility control system and instrumentation. In addition to DACS and working independently from it, the central interlock and safety system (CISS) ensures the facility machine protection and personnel safety, with specific interlock logic, acting upon off-normal events or conditions. The paper briefly describes HELOKA DACS and CISS and reports in detail on the current development status: the installation and commissioning of DACS stage 1 consisting of the supervisory control system (SCS) and the control and monitoring for the water cooling system (WCS) and the related power supply.