Demonstration Fusion Power Reactor Research Articles

Analysis of the Flow Distribution in the Back Supporting Structure Manifolds of the HCPB Breeding Blanket for the EU DEMO Fusion Reactor

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.

Interface and requirements analysis on the DEMO Heating and Current Drive system using Systems Engineering methodologies

In this paper we present the methodology implemented for analyzing System Requirements and Interfaces of the Heating and Current Drive (HCD) system of the European Demonstration Fusion Power Reactor DEMO.The work consisted in updating the preliminary framework of the Model-Based Systems Engineering model of the HCD System Architecture. This is now containing an ontology, a set of 6 perspectives and a defined set of viewpoints for each Perspective, for refining the HCD System Architecture.The scope of the work is to manage the interdependencies of HCD system elements and their integration into DEMO, for a given set of system functions. On the one hand, this means to address the identification and definition of the interfaces occurring, both internally in the HCD system, and between the HCD system and neighboring systems. On the other hand, this implies studying the impact of requirements coming from the ongoing physics studies.The rationale is to provide the technical foreground for supporting the decision-making processes related to the HCD system which is planned to be carried out during the forthcoming Conceptual Design Phase.The results we show in this paper are part of the design and integration activities consisting of both systems engineering methodologies and design analysis, all aiming at ensuring consistency in the overall EU DEMO plant design. In this framework the DEMO Heating and Current Drive system has been selected as pilot project for the application of Systems Engineering methodologies.

A multi-physics integrated approach to breeding blanket modelling and design

Often, for the design of a component, several kinds of analyses are needed. Even more frequently, the different fields of study, to be taken into account for the design verification, have to be examined minutely until the final results are satisfying. Furthermore, when geometry modifications are required, for instance to fulfill the component functions, the analyses cycle has to be restarted and an iterative process has to be carried out. This procedure may be time-consuming and herald of errors, in particular if it is demanded to the human activity. Therefore, it is more convenient for the scientific community to adopt a numerical tool that can combine various computational codes. On the base of these considerations, one of the greatest and important challenges for the new design tools is to demonstrate the capability for performing multi-physics analysis in an integrated way. This is a prerequisite, above all, when the component is part of a fusion utility like the Breeding Blanket (BB) in European Demonstration Fusion Power Reactor (DEMO). Indeed, for its design, several fields of analysis are involved such as the neutronics, thermal-hydraulics and the thermo-mechanics.The present work outlines a procedure for their coupling. The main characteristics of this new multi-physics integrated approach are (i) the use of the well-known commercial software, widely employed in the BB design, as well as (ii) the employment of the same geometry definition for all the phenomena studied. An effective application of the aforementioned approach to the pre-conceptual design of the Helium Cooled Pebble Bed (HCPB) and of the Water Cooled Lithium Lead (WCLL) is also provided in this paper. Finally, the achieved results are herewith reported and critically discussed.

Neutronics studies for the novel design of lower port in DEMO

The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) is in progress in the Power Plant Physics and Technology (PPPT) programme, within the EUROfusion Consortium. In this work neutronics studies, fundamental for the nuclear design of DEMO, are presented for a novel design of the lower port (LP). Two possible configurations of the LP have been investigated: the vacuum pumping port, with the pumping unit located inside the port, and an empty port designed for remote handling. For both configurations 3-dimensional Monte Carlo calculations have been performed with MCNP5 to assess the neutron flux inside and around the port and the nuclear heating in sensitive components, such as the toroidal field coil conductor, the vacuum pumps, the shielding elements and the port closure plate. Different shielding configurations have been considered, by adding shielding blocks at the lower port entrance. Single and double wall port walls and closure plates with different thickness have been studied to reduce nuclear loads and neutron flux. Nuclear quantities under analysis were found to be within the limit for all the components, with the exception of the nuclear heating on the toroidal field coils. The absence of any shield of the divertor cassette pumping duct, as in the DEMO 2017 baseline configuration, is responsible for a huge amount of radiation streaming inside the pumping duct that can cause the excess of heating on the coil conductor. The use of a liner, or of an equivalent shielding component, is proposed, but further improvements are needed to keep the nuclear loads on the coil conductor within the limit.

Systems Engineering approach in support to the breeding blanket design

Nowadays the Systems Engineering (SE) methodology is applied in several fields of engineering and it represents a powerful interdisciplinary means to enable the realisation of complex systems taking into account the customer and Stakeholder´s needs. Also in the fusion community, this theme is becoming increasingly pressing and the implementation of the SE approach, from the early stage of design, is now a must. Indeed, within the framework of EUROfusion activities, SE method has been selected for capturing the system and interface requirements and for their management and verification with particular focus to the Breeding Blanket (BB) System of the European Demonstration Fusion Power Reactor (DEMO). Specifically, various levels of functions and requirements have been elicited and a development of the BB SE model, set-up using the Systems Modelling Language (SySML), has been performed. This paper describes the advantages of applying a SE approach to the BB design considering, in particular, the BB requirement development and management and the definition of the interfaces between the BB and the major interconnected systems, including Remote Maintenance, Balance of Plant, Vacuum Vessel attachment, Heating and Current Drive and Fuelling Lines Systems. An effective application of SE technique to the pre-conceptual design phase of the BB is also provided in this paper.

Initial layout of DEMO buildings and configuration of the main plant systems

The realization of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle is viewed by Europe and many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. The development of suitable configurations of plant systems inside the tokamak building to allow for the various maintenance operations and to meet the numerous system operating functions and safety requirements is one of the initial priorities. A task to establish an initial layout of the DEMO buildings and the integration of the main plant systems was launched by EUROfusion with an industrial partner with experience in the construction of nuclear power plants.This paper provides an overview of the main plant systems to be integrated in the tokamak building including requirements due to their prescribed functions as well as from the overall safety and maintenance approach. These are transferred into layout requirements such as the need for separation of certain components, fire zones, access zoning, etc. In addition, the design requirements for the building itself are summarized. Some considerations of the lesson being learnt from ITER design and construction experience are provided too.

Study of a Plasma Boundary Reconstruction Method Based on Reflectometric Measurements for Control Purposes

A purely geometric approach has been investigated to reconstruct the Demonstration Fusion Power Reactor (DEMO) plasma boundary for control purposes. The whole plasma boundary is reconstructed by using a deformable template method based on B-splines. The final curve shape is achieved by minimizing the distance between a limited number of estimated and measured (at present provided by an equilibrium code) plasma boundary points along the reflectometer lines of sight. The resulting unconstrained optimization problem is solved by a simulated annealing algorithm. The method is complemented by including the available plasma and poloidal field coil current measurements to refine the boundary reconstruction in the X-point region. The robustness with respect to random measurement random errors and to a reduction in the number of measurements is discussed. The main equilibrium and shape geometric quantities (such as plasma cross-sectional area, plasma center position, elongation, and triangularity) were computed and compared to the corresponding quantities of a DEMO reference equilibrium.

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥1 km/s. An innovative approach, aimed at identifying suitable inboard “direct line” paths, to inject high-speed pellets (in the 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. A possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.

Development of helium coolant DEMO first wall model for SYCOMORE system code based on HCLL concept

The conceptual design of the demonstration fusion power reactor, known as DEMO, is ongoing and several reactor configurations have to be investigated by exploring different design parameters. For these reasons, within the European framework, systems codes like SYCOMORE (SYstem COde for MOdelling tokamak REactor) have been developed. SYCOMORE includes several specific modules, one of which is aimed to define a suitable design of the helium breeding blanket. The research activity has been devoted to improve the method to define automatically the First Wall design starting from thermal-hydraulic and thermo-mechanical considerations, using analytical design formulae and, also, taking into account the design criteria coming from Codes&Standards. Thanks to these considerations, it has been possible to derive the first dimensions of the First Wall channels from which all the other characteristics are deducted. Afterwards, it has been assessed the thermal and mechanical field using a theoretical approach. Therefore, in order to compare and validate the results, a 3D geometric model has been created and FEM analyses have been carried out finding out very satisfying results with maximum error of 6.06% for the thermal analysis while a maximum error of 20% for primary and secondary stresses.

Overview over DEMO design integration challenges and their impact on component design concepts

The EU fusion roadmap defines as one of its goals the development of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. This device shall be tritium self-sufficient, produce net electricity, and acts as a component test facility to demonstrate fusion power plant relevant technologies, e.g. those of the breeding blanket.This article identifies the main DEMO requirements, introduces the rationales for the selected architecture of the DEMO tokamak, and describes how the configuration of the main tokamak components has been derived. This includes (i) the DEMO shielding concept, (ii) the segmentation of the in-vessel components and their maintenance strategy, (iii) an overview of the vessel and in-vessel component technologies, and (iv) a description of the integrated design of the breeding blanket.

Initial definition of structural load conditions in DEMO

An essential goal of the EU fusion roadmap is the development of design and technology of a Demonstration Fusion Power Reactor (DEMO) to follow ITER. A pragmatic approach is advocated considering a pulsed tokamak based on mature technologies and reliable regimes of operation, extrapolated as far as possible from the ITER experience. The EUROfusion Power Plant Physics and Technology Department (PPPT) started the conceptual design of DEMO in 2014, see Federici et al. (2014) [1].This article defines, based on ASME III, the categories of loads to be considered in the design of the DEMO components, defines the categorization of load conditions based on their expected occurrence and provides the correlation of acceptable component damage levels. It furthermore defines the load combinations to be considered in the conceptual design phase of DEMO. Furthermore, with exception of heat loads from plasma particles and radiation to the plasma facing components, the most important load cases are described and quantified. These include (i) electromagnetic (EM) loads due to toroidal field coil fast discharge, (ii) EM loads in fast and slow plasma disruptions due to eddy and halo currents, (iii) seismic loads, and (vi) pressure loads in the dominant incident/accident events.

Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

Advances in the physics basis for the European DEMO design

In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

Considerations on the DEMO pellet fuelling system

The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

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 4m 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.

Overview of EU DEMO design and R&D activities

One important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a Demonstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several 100MW of net electricity to the grid and operating with a closed fuel-cycle by 2050. This is currently viewed by many of the nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power. This paper outlines the DEMO design and R&D approach that is being adopted in Europe and presents some of the preliminary design options that are under evaluation as well as the most urgent R&D work that is expected to be launched in the near-future. The R&D on materials for a near-term DEMO is discussed in detail elsewhere.

Oxidation behavior of plasma sintered beryllium–titanium intermetallic compounds as an advanced neutron multiplier

Beryllium intermetallic compounds (beryllides) such as Be12Ti are very promising candidates for advanced neutron multiplier materials in a demonstration fusion power reactor (DEMO). However, beryllides are too brittle to be fabricated either into pebble-type or rod-type shapes via conventional methods (i.e. arc melting and hot isostatic pressing). We have proposed a plasma sintering technique as a new method for beryllide fabrication, and our studies on the properties of plasma sintered beryllides are ongoing.In the present work, the oxidation properties of plasma sintered beryllides were investigated at 1273K for 24h in a dry air atmosphere to evaluate the high temperature properties of this material. Thermal gravimetry measurements indicate that specimens with larger fractions of Be12Ti phase corresponding to samples that have been sintered for longer time periods, exhibit superior oxidation properties.Our evaluation of the oxidation behavior of each phase in our beryllide samples is as follows: Be12Ti and Be17Ti2 both have good oxidation resistance, owing to the formation of dense and protective scales, while the Be and Be2Ti phases are mainly responsible for thermal-gravimetry (TG) weight gains, which is indicative of severe oxidation. We attribute the degradation in oxidation resistance specifically to Be2Ti that transforms into TiO2, and also find this phase to be the cause of deterioration in the mechanical properties of samples, owing to cracks near Be2Ti phase conglomerates.

Determination and prediction of axial/off-axial mechanical properties of SiC/SiC composites

A SiC/SiC composite is a candidate material for a demonstration fusion power reactor (DEMO). Identifying the inherent anisotropy of composites is required to predict axial/off-axial mechanical properties for various failure modes. This study evaluated crack propagation behavior by the various modes to provide strength anisotropy maps and we discussed a method to analytically predict this trend. The strength anisotropy maps identified for various fabric orientations clearly indicate that the composites failed in the mixed modes. Specifically, due to the axial anisotropy, five individual mode strengths such as tensile/compressive strengths in the axial/transverse directions, respectively, as well as the in-plane shear strength, are identified to become essential input parameters. With the analytical criterion based on the Tsai–Wu model, the strength anisotropy could satisfactorily be described.

Strength Anisotropy of NITE-SiC/SiC Composite by Various Failure Modes

SiC/SiC composite is a candidate material for demonstration fusion power reactor (DEMO). To identify the inherent anisotropy of composites, it requires predicting axial/off-axial mechanical properties by various failure modes. This study evaluated failure behavior of SiC/ SiC composites by various mode tests such as tensile, Inter-laminar shear and Trans-thickness tensile modes. To provide strength anisotropy maps and discussed a method to predict this trend analytically. The identification of various fabric orientations from strength anisotropy maps clearly indicate that the composites failed by the mixed modes. The strength anisotropy can be described satisfactorily. [doi:10.2320/matertrans.M2012182]

Russian concept for a DEMO-S demonstration fusion power reactor

Conceptual design studies for a tokamak-based demonstration fusion reactor have been carried out in Russia since 1991. The preferred concept was a steady-state operating tokamak with superconducting magnets, a single-null divertor configuration and a high contribution of bootstrap current into a plasma current drive. Two blanket concepts were analyzed: (1) a helium-cooled ceramic (Li 4SiO 4) design for tritium breeding, using ferritic steel as the structural material, and (2) a blanket using liquid lithium for the tritium breeding material and coolant and a vanadium–chromium–titanium alloy as the structural material. Conventional-type water/lithium-cooled divertor targets with a maximum heat load of ∼10 MW/m 2 were chosen. Blankets of both designs require beryllium as a neutron multiplier and have to be replaced after the integral fusion neutron load on the first wall reaches 10 MW a/m 2. The results of the analyses show the necessity of additional studies prior to choosing the most promising blanket concept for further development. Aspects of radioactive waste management and scarce material recycling were also considered.