Search Results (23)

The development of nuclear fusion as a viable energy source will require the manufacturing of large-scale and complex components capable of withstanding extreme operational conditions. The RODAS project will address this challenge by developing and validating advanced manufacturing technologies, particularly additive manufacturing and Hot Isostatic Pressing, to fabricate demonstrator components representative of those needed in ITER (International Thermonuclear Experimental Reactor) and EU-DEMO (European Demonstration Power Plant) reactors. The project will focus on the production and qualification of fusion-relevant materials such as EUROFER97, CuCrZr, and AISI 316L (ITER-grade), using various additive manufacturing techniques adapted to large-scale applications. Through a coordinated effort among Spanish public research institutions and private technology centers and companies, RODAS will establish specifications, develop hybrid manufacturing routes, and implement non-destructive testing methods to ensure the integrity and performance of the produced components. Additionally, it will explore strategies for future in-service repair using remotely operated systems. The project will significantly strengthen the technological capabilities and international competitiveness of the Spanish industrial sector in the field of nuclear fusion. Full article

Temperature Swing Absorption (TSA) is the primary candidate for the Isotope Rebalancing and Protium Removal (IRPR) system within the envisioned EU-DEMO fusion reactor fuel cycle. TSA separates a mixed hydrogen isotope stream into two product streams using a semi-continuous process. One stream, enriched in heavy isotopes, is used to re-establish the required deuterium-to-tritium fuel ratio. The second, enriched in protium, is stripped off from the fuel cycle to counteract the protium build-up. Separation is achieved by cycling an isotope mixture between two columns filled with metallic absorption materials that have opposite isotope effects of metal hydride formation. The selection of these materials, the operation parameters and the column geometry allow for adjusting the resulting enrichments. To identify suitable operation parameters, a TSA process model is developed which depicts the process dynamics and interactions between the columns. A modified process operation mode is introduced, which enables higher system throughputs and non-cryogenic operation, i.e., operational temperatures between 0 to 130 °C, while reducing the tritium inventory due to shorter cycling times by reduced amplitudes of the temperature swings. Finally, simulations of a TSA system at relevant scale confirm the suitability of TSA technology for the separation task of the EU-DEMO IRPR system. Full article

The pursuit of compact and efficient fusion energy systems necessitates innovative structural concepts capable of withstanding extreme operational conditions. This study presents a preliminary structural evaluation and stress assessment of a bucked-and-wedged configuration for the EU DEMO tokamak, targeting a peak magnetic field of 16.5 T. The proposed concept leverages mutual wedging of the Toroidal Field (TF) coils and their interaction with the Central Solenoid (CS) to optimize stress distribution in the inner legs, a critical region in high-field fusion reactors. To address the significant tangential forces arising during plasma operation, the design integrates outer inter-coil structures and shear pins to enhance mechanical stability. A hybrid simulation approach—coupling 3D electromagnetic and structural finite element analyses—is employed to assess stress behavior and structural integrity under both in-plane and out-of-plane loading conditions. The results contribute to the optimization study of high-field fusion reactor components and offer insights into viable mechanical design strategies for next-generation nuclear energy systems. Full article

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations. Full article

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two BB concept candidates to be selected as the driver blanket for the EU DEMO fusion reactor. In this regard, the development of a sound architecture of the WCLL Central Outboard Blanket (COB) Segment, ensuring the fulfilment of the thermal and structural design requirements, is one of the main goals of the EUROfusion consortium. To this purpose, an exploratory research campaign has been launched to preliminarily investigate the thermo-mechanical performances of the Bottom Cap (BC) region of the WCLL COB segment because of its peculiarities making its design different from the other regions. The assessment has been carried out considering the nominal BB operating conditions, the Normal Operation (NO) scenario, as well as a steady-state scenario derived from the in-box LOCA accident, the Over-Pressurization (OP) scenario. Starting from the reference geometric layout of the WCLL COB BC region, a first set of analyses has been launched in order to evaluate its structural performances under a previously calculated thermal field and to select potential geometric improvements. Then, the analysis of a complete BC region was conducted from both the thermal and structural standpoints, evaluating its structural behaviour in compliance with the RCC-MRx code. Finally, after some iterations and geometric updates, a promising geometric layout of the BC region has been obtained even though some criticalities still persist in the internal Stiffening Plates and First Wall. However, the obtained results clearly showed that the proposed layout is worthy to be further assessed to achieve a robust enough configuration. The work has been performed following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted Ansys commercial FEM code. Full article

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021. Full article

Activated Corrosion Product (ACP) formation and deposition pose a critical safety issue for nuclear fusion reactors. The working fluid transports the ACPs towards regions accessible by worker personnel, i.e., the steam generator. The code OSCAR-Fusion has been developed by the CEA (France) to evaluate the ACP generation and transport in closed water-cooled loops for fusion application. This work preliminary assesses the impact of water chemistry on the transport, precipitation, and deposition of corrosion products for the EU-DEMO divertor Plasma Facing Unit Primary Heat Transfer System. Sensitivity analyses and uncertainty quantification are needed due to the multi-physics phenomena involved in ACP formation and transport. The OSCAR-Fusion/RAVEN code coupling developed by the Sapienza University of Rome and ENEA are used. This work presents the perturbation results of different parameters chosen for a closed water-cooled loop considering a continuous scenario of 1888 days. The aim of this work is to preliminarily assess the variation of build-up of ACPs, perturbing the alkalizing agent concentration into the coolant, and the corrosion and release rates of different materials. The assessment of ACP formation deposition and transport is fundamental for source term identification, reduction of radiation exposure assessment, maintenance plan definition, design optimization, and waste management. Full article

Recent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component. Full article

The in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case. Full article

In the EU, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept is one of the candidates for the design of the DEMO reactor. From the past campaign of analysis emerged that the thermal-induced stress led to the failure in the verification of the RCC-MRx structural criteria. Hence, in this paper the classic conceptual design approach, based on a pure FEM thermal and structural analysis, is compared to a coupled thermofluid-dynamic/structural one. Even though the coupled approach requires tremendous modelling effort and computational burden, it surely allows determining the thermal field with a higher level of detail than the FEM analysis. Therefore, in this work, the focus is put on the impact of a more detailed thermal field on the DEMO WCLL BB global structural performances, focusing on the Top Cap region of its Central Outboard Blanket segment. The obtained results have allowed confirming the soundness of the design solution of the Top Cap region, except for concerns arising on the mass flow rate distribution. Moreover, results have shown that, globally, the pure FEM approach allows for obtaining more conservative results than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible adopting the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalise the most promising. Full article

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO. Full article

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed. Full article

The Water-Cooled Lithium–Lead (WCLL) is one of the most promising technologies for power conversion and tritium production in future fusion-powered reactors; it will be implemented in one of the Test Breeding Modules (TBM) inside the ITER reactor and the DEMO EU reactor. However, the simultaneous presence in the system of high-temperature PbLi and high-pressure water poses significant safety issues in the event of an in-box LOCA (Loss Of Coolant Accident). For this reason, a complete understanding of the system response is crucial to avoid extensive damage in such a scenario. This paper describes the status and design features of the LIFUS5/Mod4 facility, an experimental plant that is currently being designed and constructed at ENEA CR Brasimone in the framework of the FP9 EUROfusion Horizon Europe to address these issues. This facility aims at being representative of the geometry and operational conditions of the Test Breeding System (TBS) to allow the precise reproduction of its behavior under simulated incidental scenarios. For this reason, peculiar design choices have been made, which will be extensively discussed throughout this work and which will allow the generation of high-quality data useful for the TBS development. Moreover, the facility is expected to become a test stand for the implementation of different safety functions, to identify the best accident-mitigation strategy. Possible upgrade plans for the facility are described as well, with the chance for it to become a fully functional test stand for any component of the TBS in their operative conditions. Full article

The international fusion materials irradiation facility-DEMO-oriented neutron source (IFMIF-DONES) is a facility that is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an energy spectrum like the one present in a fusion power reactor. The material specimens are irradiated in a containment structure named the test cell (TC), which is part of the test systems (TS). The TC also houses a part of the other major system (lithium system, LS), which provides the liquid lithium for the reaction through a piping system. At a point, the lithium piping needs to exit the TS, but the primary safety boundary must be continuous around these penetrations. Therefore, a special barrier, called the test systems–lithium systems interface cell (TLIC), has been developed around the piping system to provide a safety-approved and remotely maintainable vacuum boundary envelope. In this paper, the integrated design development of the TLIC is described, consisting of the design development according to the RCC-MRx code, the remote-handling (RH) needs, and the procedures and safety-related special needs of the design. Full article

The fusion program in the Karlsruhe Institute of Technology (KIT) leads the R&D of the DEMO helium-cooled pebble bed (HCPB) breeding blanket within the work package breeding blanket (WPBB) of the Eurofusion Consortium in the European Union (EU). A new design of the HCPB breeder zone, with a layout inspired by a nuclear reactor fuel rod arrangement, was developed recently and called the fuel-breeder pin concept. In addition, a mock-up (MU) of this fuel-breeder pin was designed and manufactured at KIT in order to test and validate its thermal–hydraulic performance. This paper reports on the results of the first experimental campaign dedicated to the fuel-breeder pin MU testing that was performed in the Helium Loop Karlsruhe (HELOKA) facility. The paper presents: (i) the integration of the fuel-breeder pin MU into the HELOKA loop including considerations of the experimental set-up, (ii) an overview of the plan for the experimental campaigns, and (iii) a discussion of the experimental results with a focus on aspects relevant for the validation of the thermal–hydraulic design of the HCPB breeder zone. Full article