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Within the DEMO first wall 3D shape design activity, studying the effect of misalignment started in 2017. Such assessments have been conducted in the past for ITER and heat flux penalty factor maps have been created [ 1 ]; this route could be a feasible approach in the case of DEMO as well. This paper details the methodology that allows the effects of misalignments to be assessed for DEMO. The test cases focus on steady-state plasma operation (start of flat top). The aim is to understand the effect of basic misaligned cases, for example radial protrusion/recession or poloidal rotation of a single module. To do so, particle tracing software codes such as SMARDDA and PFCflux have been used to create heat flux maps that reach the first wall surfaces. These heat flux maps, combined with the specified radiative heat load, are used as input for simplified FE models of the blanket modules. As a result, not only the effect on heat flux, but also on the temperature (and later stress) distributions can be estimated. The paper describes how the obtained results can be implemented in ANSYS in the identified critical cases from the test matrix that has been studied. The results obtained from the nominal heat flux map are compared to the misaligned cases. The mitigating effect of the 3D nature of the heat conduction on the peak temperature is discussed. This work paves the way to assessing more realistic combined misaligned cases (such as misalignment from different thermal expansion, or due to electromagnetic loads etc. of neighbouring blankets) in the future. [ABSTRACT FROM AUTHOR]

Highlights • A CO 2 -cooled FW has been proposed in this paper. • A high heat flux with radiation heat flux and particle heat flux has been applied on FW. • Different wall roughness position of channel has been shown. • Different wall roughness value has been given. • Sensitivity particle heat flux has been studied in this paper. Abstract The Demonstration Fusion Reactor in Europe (EU-DEMO) is considered to be the last step before building a commercial fusion power plant. Up to now, He has been chosen as gas coolant for three of the four candidate breeding blankets proposed for the EU-DEMO, being the Helium Cooled Pebble Bed (HCPB) among them. The choice of He coolant is supported, among other features, by its superior heat transfer capabilities, its transparency to neutrons and chemical inertness. However, the use of this coolant in DEMO poses some feasibility questions, like the large coolant inventory, resulting in possible safety issues, as well as the large circulating power. On the other side, CO 2 has been the fluid coolant choice in the nuclear fission industry for gas-cooled reactors since the 1950s. The key advantage of this gas is its larger density, which results in more compact components, lower coolant inventories, lower circulating powers and thus better plant thermal efficiency. This, coupled also with a relatively high molecular stability at moderate temperatures (up to 600 °C), its larger availability and lower price make CO 2 an attractive coolant for a pebble bed breeding blanket, as an alternative to He. As a Plasma Facing Component, the First wall (FW) is one of the components under the most challenging environment. Currently, the FW region at the upper port inboard location is expected to suffer the most critical heat flux loads. Indeed, a peak charged particle heat flux 0.99 MW/m2 together with a radiation heat flux of 0.18 MW/m2 is foreseen at this location. This paper reports the thermal-hydraulic analyses carried out to find the practical heat flux limits by using CO 2 as coolant. The analyses have been performed by using the commercial computational fluid dynamics (CFD) code ANSYS CFX. Different augmented technical wall roughness areas have been considered, for a best heat transfer-to-pressure drop ratio. Sensitivity analyses have been performed as well for higher peak particle heat fluxes, evidencing that a CO 2 -cooled FW can dissipate peak heat fluxes up to 1.8 MW/m2 keeping the circulator power in a reasonable range. [ABSTRACT FROM AUTHOR]

In the framework of the European "HORIZON 2020" program, EURO fusion develops a fusion power demonstrator (DEMO). According to a recent study about plasma heat loads, it appeared that the First Wall, which is the first part of the Breeding Blanket in front of the plasma, will face some high heat fluxes on some poloidal locations of the Breeding Blanket. In order to cope such high heat fluxes, this paper presents the investigation on different Helium cooled First Wall design integrated to the Breeding Blanket on the basis of standard square and circular smooth channels and with different options for the Tungsten armor surrounding the channels, taking advantages to the different material properties. The performances of the different concepts have been assessed with thermal and mechanical Finite Element Method numerical simulation based on a slice of the Helium Cooled Lithium Lead concept. Results are compared with the RCC-MRx design rules to prevent failure during normal and accidental condition. The results show that the options with channels surrounded by Tungsten could meet some plasma heat loads requirements from design point of view. However, the concept is still at an early stage of development and open issues are discussed. [ABSTRACT FROM AUTHOR]

Highlights • FEM analysis is performed to evaluate stresses at the geometries of FW cooling channels which can handle two different types of heat fluxes using water as a coolant. • For handling normal heat fluxes round tubes with countercurrent flow are chosen while for handling enhanced heat fluxes channel with the hypervapotron configuration is chosen. • The countercurrent flow cooling channels were able to withstand heat flux 1.7 MW/m2 and 15.0 MPa pressure inside the tubes. • The hypervapotron geometry needs further optimization to withstand required heat flux and pressure. Abstract The European DEMOnstration (DEMO) reactor precedes the development of the fully commercial fusion power plant. One of the key components of the DEMO is the first wall (FW). The FW is acted upon by extreme particle and radiation loads, which results in high heat fluxes varying across the reactor vessel. This paper attempts to evaluate two DEMO FW water cooling concepts from the stress point of view. The heat sink material chosen for the FW is Eurofer97 with thermal conductivity of ∼28 W/mK at 20 °C; the cooling conditions opted are those of pressurized water reactor (PWR), i.e. 285 °C and 15.5 MPa. In the present work it is attempted to perform the stress analysis of the two different heat sink channel configurations, these are circular pipes with countercurrent flow and channel with hypervapotron configuration. The thermo-hydraulic optimization of these components has been performed in one of the author's previous work. The stress in the components is caused by the pressure of the fluid and increase in the temperature of the material. To accomplish the work, commercial finite element method software ANSYS is used. In a case of unsatisfactory results, further recommendations are given. Simulations are performed with P91, which is material being commonly used in power engineering and will be also used for manufacturing of the discussed configuration of DEMO FW mock-ups which will be afterwards tested in high heat flux facility HELCZA. [ABSTRACT FROM AUTHOR]

Abstract The geometrical freedom that additive manufacturing (AM) provides enables realisation of formerly impossible design concepts for high heat flux components aimed at use in a fusion reactor. This paper demonstrates a number of these advantages using a tile-based divertor target with two examples of novel cooling geometry. To showcase the potential benefits of high temperature operation using materials newly available for AM processing, tantalum was used as the structural material. Rather than present optimised designs, these two concepts highlight features enabled by this new technology. Simple quantitative analysis shows heat transfer improvements of 25% for a multiple small pipe concept compared to single pipe designs. Finite element analysis for both designs, including a tungsten armour, also demonstrates a more uniform temperature distribution, reducing thermal stresses below elastic limits for 5 MWm−2 heat flux cases, even with 600 °C coolant. At 10 MW m−2 yield is exceeded but the expectation is further design optimisation should enhance structural integrity. Overall, there are modest gains in heat flux handling, lifetime, and thermal efficiency compared to existing concepts, but there are significant improvements in waste reduction and remote handling mass, using typically 80% less material. Moreover, integrated manifolding and careful location of coolant connection is suggested to facilitate repair and replaceability, which should also reduce risk associated with qualifying joints. [ABSTRACT FROM AUTHOR]

Highlights • Preliminary design of DEMO first wall 3D surface is presented. • Steady state and controllable perturbations are preented,evaluating the radiated HF density on the wall and the charged particle power. • Different plasma-wall contact phases have been analysed for a list of foresseable and unforeseeable transients. • 3D field line tracing codes were employed to analyse and design discrete high heat flux components surfaces. Abstract The present DEMO breeding blanked design heat load capability is limited to ≈1 MW/m2 for steady state plasma loading, due to the specific requirements on high neutron irradiation capable materials, and high coolant temperature for efficient energy conversion. While this limit is achievable in nominal conditions in the present DEMO blanket concept designs, the greatest challenges arise from the occurrence of plasma transients. The results of simulations of a number of plasma transients are presented in this paper. 3D field-line tracing codes have been used to analyses the maximum heat flux and energy density for a specific first wall shape design, and optimize it. A scoping study has been performed with the thermal analysis code RACLETTE, using a broad range of transient input heat fluxes, on a series of high heat flux (HF) components concepts with tungsten armor, Eurofer steel or copper alloy as heat sink materials, and helium or water as coolant. The results permit the identification of the operational space of the peak HF density that can be tolerated by different plasma facing components, for the different transient models. [ABSTRACT FROM AUTHOR]

The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. In DEMO, divertor target must survive an environment of high heat fluxes (∼up to 20 MW/m 2 during transient) and neutron irradiation. This paper presents the first part of the development for the realization of an advanced actively cooled concept aiming at improving the quality of material joining. It is based on the use of a W/Cu functionally graded material (FGM) interlayer instead of the copper interlayer. For this preliminary development, the interlayer (∼25 μm) is made with PVD (Physical Vapor deposition) and composed of continuous gradient of W from 100% at the interior part of the tile to 0% at the external part while Cu varies from 0% to 100%. Hot isotactic pressing is realized to join the CuCrZr tube to W tile equipped with graded material. With infrared thermography (SATIR facility) and US non-destructive examinations, manufactured mock-ups show good interface quality. [ABSTRACT FROM AUTHOR]

This paper analyzes a possible operational space for the DEMO device with the tin divertor. The simulation is performed with the COREDIV code which self-consistently solves core and scrape-off layer transport equations for plasma and impurity. A simple model of divertor surface evaporation was included in the code. The simulation was performed for the DEMO 1 April 2015 configuration. Influence of the sputtering, prompt redeposition and evaporation of the liquid tin divertor was taken into account. Simulation without impurity seeding shown that plasma in DEMO with the tin divertor is characterized with higher than 80% radiation fraction, and 55 MW of power to the plate. Power across the separatrix stays slightly below the L-H threshold. In spite of high radiation fraction seeding with additional impurity is necessary to reduce the power to the divertor targets to acceptable levels. In presence of the argon seeding power to the plate is reduced to 23 MW, power across the separatrix reaches 150 MW which is compatible with the H-mode operation. [ABSTRACT FROM AUTHOR]

The present European standard DEMO divertor target technology is based on a water-cooled tungsten mono-block with a copper alloy heat sink. This paper presents the assessment of the operational space of this technology under static and transient heat loads. A transient thermo-hydraulic analysis was performed using the code RACLETTE, which allowed a broad parametric scan of the target geometry and coolant conditions. The limiting factors considered were the coolant critical heat flux (CHF), and the temperature limits of the materials. The second part of the work is devoted to the study of the plasma strike point sweeping as a mitigation technique for the divertor power exhaust. The RACLETTE code was used to evaluate the impact of a large range of sweeping frequencies and amplitudes. A reduced subset of cases, which complied with the constraints, was benchmarked with a 3D FEM model. A reduction of the heat flux to the coolant, up to a factor ∼4, and lower material temperatures were found for an incident heat flux in the range (15–30) MW/m 2 . Finally, preliminary assessments were performed on the installed power required for the sweeping, the AC losses in the superconductors and thermal fatigue analysis. No evident show stoppers were found. [ABSTRACT FROM AUTHOR]

The first wall (FW) of DEMO is a component with high thermal loads. The cooling of the FW has to comply with the material's upper and lower temperature limits and requirements from stress assessment, like low temperature gradients. Also, the cooling has to be integrated into the balance-of-plant, in a sense to deliver exergy to the power cycle and require a limited pumping power for coolant circulation. This paper deals with the basics of FW cooling and proposes optimization approaches. The effectiveness of several heat transfer enhancement techniques is investigated for the use in helium cooled FW designs for DEMO. Among these are wall-mounted ribs, large scale mixing devices and modified hydraulic diameter. Their performance is assessed by computational fluid dynamics (CFD), and heat transfer coefficients and pressure drop are compared. Based on the results, an extrapolation to high heat fluxes is tried to estimate the higher limits of cooling capabilities. [ABSTRACT FROM AUTHOR]

Among the proposed First Wall (FW) cooling concepts for European Demonstration Fusion Power Plant (DEMO), water cooled FW is one of the options. The heat flux load distribution on the FW of the DEMO reactor is not yet precisely defined. But if the heat loads on the FW are extrapolated from ITER conditions, the numbers are quite high and have to be handled none the less. The design of the FW itself is challenging as the thermal conductivity ratio of heat sink materials in ITER (CuCrZr) and in DEMO (Eurofer 97) is ∼10–12 and the operating conditions are of Pressurized Water Reactor (PWR) in DEMO instead of 70 °C and 4 MPa as in ITER. This paper analyzes the performance of Hypervapotron (HV) heat sink for FW limiter application under DEMO conditions. Where different materials, temperatures, heat fluxes and velocities are considered to predict the performance of the HV, to establish its limits in handling the heat loads before reaching the upper limits from temperature point of view. In order to assess the performance, numerical simulations are performed using commercial CFD code, which was previously validated in predicting the thermal hydraulic performance of HV geometry. Based on the results the potential usage of HV heat sink for DEMO will be assessed. [ABSTRACT FROM AUTHOR]

Abstract: Helium-cooled reactors offer several important advantages as the helium is an inert and neutron transparent gas, compatible with high outlet temperatures and all structural and functional materials considered for plasma facing components. On the other hand, during the power plant studies performed up to now it became evident, that the helium pumping power could become very high if standard cooling methods based on hydraulically smooth cooling channels would be applied and that a very high heat flux in the divertor can only be achieved when using sophisticated gas cooling methods. In relation to this the main focus in this paper is on advanced gas cooling technology applications developed at Forschungszentrum Karlsruhe (FZK) for the main in-vessel components blanket/first wall and divertor of future fusion power reactors. Among a large number of considered cooling technologies the V-rib-based surface roughness approach for the first wall and the multiple jet impingement cooling technique for the divertor were identified as suitable and advantageous methods. The paper presents the main parameters to be considered and an evaluation of the achievable cooling performance in both cases. [Copyright &y& Elsevier]

This work presents an overview of the integrated strategy developed, as part of the DEMO Key Design Integration Issue 1 (KDII1), to protect the EU-DEMO first wall (FW) from planned and unplanned plasma transients by employing discrete limiters. The present Breeding Blanket (BB) FW design, which aims at minimizing the loss of neutrons while travelling to the breeding zone, is able to withstand steady state heat fluxes up to ≈1-1.5 MW/m² [1], which is not sufficient to guarantee its integrity for most plasma-FW direct contact. This is different from ITER, which has a FW designed for peak heat loads up to 4.6 MW/m2 [2], and it does not have the DEMO BB breeding related requirement. A series of documents was compiled in the DEMO Pre-Conceptual Design Phase, in support of the KDII1. The work presented here was presented at the 2020 DEMO Gate 1 (G1) review, and collects also the comments of the panel and the relative additional studies triggered by them. The design process, presented in this paper was adopted to systematically evaluate the impact of design changes, or new physics inputs, on the FW protection strategy and integration issues. It includes compiling the list of transients, and performing the relative plasma simulations, the design of discrete limiters and the evaluation of their capability to reduce the heat flux density on the FW, and finally a preliminary analysis of the heat loads effects on the Plasma Facing Components (PFC). All these aspects, together with preliminary limiter design, where considered since the beginning, in an integrated way. [ABSTRACT FROM AUTHOR]

• An overview about the engineering aspects of the European DEMO wall protection strategy is given. • The background of the wall protection issue and design approaches are explained. • Recent outcomes from the interim design studies and technology R&D efforts are presented. • Focus is placed on the innovative technology achieved for a novel armor material and joining. • Results from the extreme and medium heat flux tests are also presented. Off-normal macroscopic plasma instabilities which possibly occur in a fusion power plant represent the most critical operational accidents and pose a serious risk to economically viable exploitation of fusion power. During such an instability, a huge amount of plasma energy is rapidly deposited onto the plasma-facing first wall in form of extreme heat flux. Such heavy thermal shocks can cause severe damages or a fatal failure of the first wall even by a single event. This issue is a serious design concern raising the requirement of a wall protection strategy. To cope with this issue, the European DEMO project adopted the limiter concept. The limiters equipped with tungsten sacrificial armor shall be installed at those wall areas where plasma contact is expected to be probable. To implement the limiter concept, comprehensive engineering studies have been conducted including conceptual design of the plasma facing component and technology R&D for armor material and joining. In this overview paper, the background of the wall protection issue and its implication on design and materials are explained and the outcomes from the recent project period (2019-20) are reported. The focus is placed on the technology achievement related to the development of novel lattice-type tungsten armor material fabricated by an innovative additive manufacturing process. Interim results from the extreme and medium heat flux tests and the joining trials for tungsten-to-heat sink material (copper or steel) are also addressed. The design rationale and the interpretation of the test data are supported by computational simulations. [ABSTRACT FROM AUTHOR]

Abstract: A present topic of high interest in magnetic fusion is the “gap” between near-term and long-term concepts for high heat flux components (HHFC), and in particular for divertors. This paper focuses on this issue with the aim of characterizing the international status of current HHFC design concepts for ITER and describing the different technologies needed in the designs being developed for fusion power plants. Critical material and physics aspects are highlighted while evaluating the current readiness level of long-term concepts, identifying the design and R&D gaps, and discussing ways to bridge them. [Copyright &y& Elsevier]

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported. [ABSTRACT FROM AUTHOR]