Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components (original) (raw)
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Fusion Engineering and Design, 2020
The Divertor Plasma Facing Components (PFCs) of a fusion reactor are the most loaded components in terms of high heat fluxes, which, combined with high neutron irradiation, can severely compromise their thermo-mechanical and physical properties as well as their heat removal capacity. Therefore, neutronic assessment plays a key role in the design of these critical components. The aim of this work is to perform a dedicated nuclear analysis for the European DEMO divertor PFCs placed on the vertical targets, aimed to provide significant outcomes in the PFCs selection concept. In particular, the present assessment is devoted to the reference ITER-like configuration under study within the EUROfusion WPDIV-PPPT programme. Three-dimensional neutronics analyses have been performed with the MCNP5 Monte Carlo code. This work presents detailed neutronics results with heterogeneous materials constitution and actual geometry of the PFC concept. High resolution data on the nuclear heating density and neutron damage of the ITER-like PFCs placed on the divertor vertical targets, including helium production, assessed for first time for the latest DEMO design, are presented and discussed.
Pre-conceptual design of EU-DEMO divertor primary heat transfer systems
Fusion Engineering and Design, 2021
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.
R&D on divertor plasma facing components at the Institute for Plasma Research
Nukleonika, 2015
This paper is focused on various aspects of the development and testing of water cooled divertor PFCs. Divertor PFCs are mainly designed to absorb the heat and particle fluxes outflowing from the core plasma of fusion devices like ITER. The Divertor and First Wall Technology Development Division at the Institute for Plasma Research (IPR), India, is extensively working on development and testing of divertor plasma facing components (PFCs). Tungsten and graphite macro-brush type test mock-ups were produced using vacuum brazing furnace technique and tungsten monoblock type of test mock-ups were obtained by hot radial pressing (HRP) technique. Heat transfer performance of the developed test mock-ups was tested using high heat flux tests with different heat load conditions as well as the surface temperature monitoring using transient infrared thermography technique. Recently we have established the High Heat Flux Test Facility (HHFTF) at IPR with an electron gun EH300V (M/s Von Ardenne A...
Insulated fixation system of plasma facing components to the divertor cassette in Eurofusion-DEMO
Fusion Engineering and Design, 2020
The design activities of an insulated Plasma Facing Components-Cassette Body (PFCs-CB) support has been carried out under the pre-conceptual design phase for Eurofusion-DEMO Work Package DIV-1 "Divertor Cassette Design and Integration"-Eurofusion Power Plant Physics & Technology (PPPT) program. The Eurofusion-DEMO divertor is a key in-vessel component with PFCs which directly interact with the plasma scrape-off layer. The PFCs have to cope with high heat loads, neutron irradiation and electromagnetic loads. The mechanical integrity of the PFCs and water cooling pipes can be jeopardized by high heat loads and by electromagnetic loads generated in a disruption event. In European-DEMO the possibility to estimate the heat load by measuring the relative thermocurrents is under investigation. In order to allow thermocurrents measurements, a divertor design option provides that PFCs are electrically insulated from CB. In this work authors aim to analyze the opportunity that the PFC-CB fixing system incorporates an electrical insulation system, thus acquiring also an important diagnostic role in the measurement of the thermocurrents and in the management of the current flows. The possible use of ceramic material (e.g. alumina) as the insulating layer between the support components is investigated.
2021
The ITER tokamak, the experimental fusion reactor designed to be the first to produce net energy, has had a monoblock concept selected for use as a plasma facing component in the divertor region. This design currently consists of a CuCrZr cooling pipe surrounded by a copper interlayer and embedded in a tungsten armour plate. Additive manufacturing may facilitate a geometry capable of greater efficiency through the introduction of greater design freedom whilst maintaining compatibility with the monoblock concept. This is achieved through the addition of high conductivity material to the armour domain surrounding the coolant pipe. Finite element simulation of the heat transfer system combined with a topology optimisation methodology has been used to find the optimal distribution of high thermal conductivity material (such as Cu) for three thermal objectives: minimising temperature and thermal gradient, and maximising conductive heat flux. The topology optimisation relies on a density-...
A fusion reactor design with a liquid first wall and divertor
Fusion Engineering and Design, 2004
Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840 MW of fusion power of which 767 MW is in the form of energetic particles (alpha power) and 3073 MW is in the form of neutrons. The alpha plus auxiliary power total 909 MW of which 430 MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.
Conceptual design studies for the European DEMO divertor: Rationale and first results
Fusion Engineering and Design, 2016
In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package 'Divertor' consists of two project areas: 'Cassette design and integration' and 'Target development'. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.
IEEE Transactions on Plasma Science
Heating & Current Drive (H&CD) systems are being investigated for a demonstration fusion power plant DEMO to deliver net electricity for the grid around 2050 [1],[2]. Compared to ITER, which has to show the generation of 500 MW thermal power, the target of DEMO is the successful production of 300 to 500 MW electrical power to the grid and to aim for a self-sufficient Tritium fuel cycle [3]. Three H&CD systems are under development for DEMO in Europe, the Electron Cyclotron (EC) System, the Neutral Beam Injection (NBI) System and the Ion Cyclotron (IC) System. Based on present studies [4] for plasma ramp-up, rampdown and flat top phases, to be further validated in more detailed simulations, the assumed total launched power needed from the H&CD system in DEMO is in the range of 50-100 MW, to be provided for plasma heating and control. The paper describes the designs and R&D status of selected H&CD systems considered for their deployment in the EU DEMO. It was always considered that different H&CD Manuscript submitted on 28 th June 2016. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The reviews and opinions expressed herein do not necessarily reflect those of the European Commission.
Fusion Engineering and Design, 2001
In a next step D/T fusion device like ITER, an intense neutron flux will be produced as a consequence of the nuclear fusion reactions. The effects of the neutron induced damage in the microstructure of the plasma-facing material (PFM) may significantly change the thermal properties and the mechanical properties as well as the behaviour of the swelling and the tritium retention in such materials. In addition, a peak heat flux as high as 20 MW m − 2 and a plasma flux of 10 18 -10 20 cm − 2 s − 1 are expected in the divertor zone during the normal operation of the reactor. The divertor materials have to withstand the neutron damage, the high heat fluxes and the high erosion caused by the interaction with the high flux plasma. The sputtered particles are co-deposited with plasma, which may contribute significantly to the total tritium inventory in the PFM. Furthermore, the interaction of steam with the sputtered particles (with usually high specific surfaces) could produce large amounts of hydrogen. All of the above topics represent critical issues for plasma performance, safety and economy, as they could limit the use of some PFM materials in next generation fusion devices. Therefore, substantial R&D effort is needed to elucidate the effects of the neutron induced damage on microstructure, erosion/deposition, tritium retention and dust formation, as well as on hydrogen production. In the framework of the European Fusion R&D program, an extensive effort on neutron effects of the material properties: namely, thermal conductivity, mechanical properties, dimensional stability, tritium trapping, erosion/deposition, co-deposition, dust formation/removal, chemical reactivity with steam and oxygen, outgassing, baking and tritium removal from PFM have been undertaken during the past several years. In this paper, : S 0 9 2 0 -3 7 9 6 ( 0 1 ) 0 0 2 5 5 -1 the recent progress achieved within the European Fusion R&D program and contributions to the development of ITER PFMs are presented and critically discussed.