Fusion Power Plant Research Articles

Neutron availability in the Complementary Experiments Hall of the IFMIF-DONES facility

The IFMIF-DONES facility will be dedicated to the irradiation of structural materials planned for the use in future fusion reactors such as DEMO (Demonstration Fusion Power Plant). The potentialities of the IFMIF-DONES facility to complement its principal purpose by other experiments that would open the facility to other communities is addressed in this work. It concerns a study based on simulations to evaluate the neutronic performances of IFMIF-DONES in an hall dedicated to complementary experiments where neutrons can be transported. With the simple beam tube geometry of 4.5 cm entrance diameter studied in this work we have shown that a collimated fast-neutron beam of about 2 1010 n/cm2/s is available in the hall. Adding a moderator in the hall with neutron extraction lines would allow to get thermal neutron beams of about 106-107 n/cm2/s for dedicated experiments. The results show that IFMIF-DONES has the potentialities to be a medium-flux neutron facility for most of the neutron applications.

Benefits and Challenges of Advanced Divertor Configurations in DEMO

The divertor design and configuration define the power exhaust capabilities of DEMO, and act as a major design driver. They set a number of requirements on the tokamak layout, including port sizes, poloidal field coil positions, and size of toroidal field coils. It also requires a corresponding configuration of plasma-facing components (PFCs) and a remote handling scheme to be able to handle the cassettes and associated in-vessel components the configuration requires.There is a risk that the baseline ITER-like single-null (SN) divertor configuration cannot meet the PFC technology limits regarding power exhaust while achieving the target plasma performance requirements of DEMO or a future fusion power plant. Alternative magnetic configurations (AMCs) – for example, double-null, snowflake, and super-X – exist and potentially offer mitigation solutions to these risks and a route to achievable power handling in DEMO. However, these options impose significant changes on machine architecture, increase the machine complexity and affect remote handling and plasma physics and so an integrated approach must be taken to assessing the feasibility of these options.This paper describes work carried out to define a set of design limitations that any alternative divertor configuration for DEMO must consider, and assesses the feasibility of integration of a set of potential AMCs for DEMO.

Plasma parameters and tungsten sputter rates in a high-frequency CCP

Monitoring and investigating the fusion plasma in ITER will be crucial to pave the way to a fusion power plant. However, the harsh conditions in the vacuum vessel are detrimental for the optical diagnostics systems. Replacing the element with a direct line of sight to the fusion plasma by a metallic mirror shifts the problem to this component. The flux of impurities onto these mirrors accumulate to deposits, which degrade their optical properties over time. It has been proposed to address this issue by igniting discharges in front of the mirrors during the maintenance phases allowing the deposited material to be sputtered away and recover the mirror properties. To further the knowledge for such an option, in this work, plasma parameters and sputter rates in a high-frequency (60 MHz) capacitive discharge in argon at pressures below 10 Pa are studied. The powered electrode consists of tungsten as a cheap rhodium proxy—the material of the metallic mirrors in ITER—and to simulate tungsten deposition. Its size is equivalent to a mirror for charge-exchange recombination spectroscopy at ITER (8.5 cm × 18 cm). The discharge is studied using and interpreting voltage measurements, microwave interferometry, electrical probe measurements, and optical emission spectroscopy. These investigations provide the opportunity to identify the optimal conditions for the process based on various requirements, such as damage threshold of the mirror crystal and severity and type of contamination.

Evaluation of the thermal-hydraulic performances of a once-through steam generator in nuclear fusion applications

After decades of operation in nuclear power plants, Once-Through Steam Generators (OTSGs) were recently proposed for nuclear fusion applications. In particular, they are supposed to be installed in the primary cooling systems of the European Union Demonstration fusion power plant (EU-DEMO). One of the key reactor components is the Breeding blanket (BB). Among the BB concepts that are currently under study, Water-Cooled Lithium-Lead (WCLL) option was considered for this work. The WCLL blanket is divided in two main subsystems, the breeder zone (BZ) and the first wall (FW), each one provided with an independent cooling circuit, named Primary Heat Transfer System (PHTS). Thermal power removed from BB by BZ and FW PHTS is driven to the Power Conversion System (PCS) to be converted into electricity. The thermal coupling is ensured by two OTSGs per system. At the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome, a simulation activity was carried out to understand the component thermal-hydraulic behavior during DEMO normal operations. For calculation purposes, a full model of the steam generator was prepared by using a modified version of RELAP5/MOD3.3 system code. The computational activity performed allows to preliminary characterize the OTSG thermal-hydraulic performances during both pulse and dwell phases.

Termomeccanica Pompe backs clean fusion research

An experimental programme is under way to prepare the way for clean fusion power plants. Termomeccanica Pompe (TMP) is part of it.

Are tungsten-based nuclear fusion components truly invisible to x-ray inspection?

The ability to detect undesired volumetric defects in reactor components could affect the safety and reliability of a fusion power plant and change the expected lifetime and performance of the reactor. This is even more true for critical reactor parts like plasma-facing components (PFCs) which have to withstand challenging in-vessel conditions due to a combination of plasma bombardment, radiation, and nuclear heating. The structural integrity of these components prior to their installation in a nuclear fusion reactor needs to be assessed non-destructively. Until now, industrial x-ray radiography and tomography have not been used to non-destructively inspect fusion components due to their lack of penetration power into dense material such as tungsten which is often used to manufacture PFCs. However, aiming to revert this consolidated belief, we have demonstrated for the first time the feasibility of assessing volumetric defects non-destructively on DEMO divertor mock-up by means of MeV energy range x-ray tomography. The authors believe that the application of this technology could be easily extended for inspecting large fusion components and positively impact procedures to be followed in the qualification of fusion components for current and future nuclear reactors.

Dimensional analysis on incipient flow instability in one-side high heat loaded rectangular flat heat sink under sub-cooled flow boiling conditions

The onset of flow instability (OFI) is a major threat to the operation of a fusion power plant. In particular, because the divertor is loaded with a high heat flux of up to 20 MW/m2, it may be vulnerable to OFI. Therefore, in this study, the OFI of a one-side heated flat heat sink was experimentally analyzed. When the effect of the system parameters on the OFI was analyzed, the faster the vapor inside the channel could be removed, the higher was the OFI that was experimentally recorded. As the flow rate and degree of subcooling increase, the OFI increases because it induces an enhancement of the forced convective heat transfer performance and a rapid coding rate, respectively. However, when the pressure is increased, the latent heat and liquid surface tension are reduced; thus, boiling occurs at a lower heat rate, which leads to a decrease in OFI. When the prediction performance of the existing OFI correlations developed under the subcooled flow boiling condition was evaluated, it showed a tendency to overpredict this experimental value significantly. This is because the evaluated correlations were developed based on relatively low heat flux conditions and narrow rectangular channels. Therefore, in this study, we developed a new OFI correlation optimized for experimental values using an artificial intelligence (AI) regression method. Because this correlation can be utilized under one-side high heat load conditions, it is expected that it will be especially useful when establishing the operation strategy of the cooling system of a fusion power plant.

Fusion DEMO sCO2 layout design with battery farm

Nuclear fusion is a promising low-carbon and low-emission source of energy. One of the first fusion power plants will be the European Union's demonstration fusion power plant DEMO. Among the key attributes that influence the whole DEMO design is the fusion reactor pulse operation. Due to the power fluctuations of the power source, there will be a significant impact on the power plant technology, turbine, etc. Therefore, a power conversion system based on a supercritical CO2 capable of operating in two nominal power levels is proposed. The system operates in nominal parameters during the entire power cycle without the need for thermal power pulses balancing. Unlike other designs, instead of connecting the energy storage system directly to the heat transfer system, this article proposes a layout with an energy storage system behind the generator. Power pulses are balanced using a battery farm, compensating the fluctuations in the gross power and power plant self-consumption. Power conversion system is based on a sCO2 Brayton simple cycle with regeneration and includes technology for nominal operation at two levels of thermal power. Optimization of the proposed layout shows thermodynamic net efficiency of 24 %.

Tokamak cooling systems and power conversion system options

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing.

EU-DEMO design space exploration and design drivers

Once the high-level requirements for a fusion power plant are set, the expected performance of plant systems, including the plasma, must be defined and then the available design space for the plant can be explored in order to converge on a final overall plant operating point, setting the major plant parameters such as number of toroidal field (TF) coils, tokamak major radius, plant power flows etc. Such design space explorations are conducted using systems codes which contain simplified models for plant systems, and attempt to capture the interactions between them in order to balance performance trade-offs and enforce overall consistency. This paper summarizes the work carried out to identify the EU-DEMO baseline operating point and the underpinning assumptions about technology and physics performance. The major design choices are described and the principle drivers for the direction of conceptual development, resulting in the operating space for EU-DEMO, are identified. The final output of the systems code forms the basis for more detailed engineering and physics evaluation and design work.

Prospective research and development for fusion commercialisation

As well as the mainstream activities aimed at delivering DEMO to meet the high-level targets and timetable outlined elsewhere, there is a parallel stream of research investigating and developing technology to provide alternative and risk-mitigating options for DEMO and/or a fusion power plant, targeting the delivery of commercially viable fusion energy. This programme is termed Prospective Research and Development (PRD). The focus is on research into promising alternative technologies that do not form part of the main DEMO programme due to their current readiness level or higher programmatic risk due to development uncertainty, but which offer the potential for improved reactor performance in the long-term, and/or risk mitigation in case the baseline options cannot be validated. These alternatives may naturally, if achieved in time, be re-adopted into the DEMO programme. As well as the technological challenges, fusion has to meet social and economic requirements. Therefore in complement to research on technology, there is research into the socio-economic factors which are anticipated to impact the wider adoption of fusion: future energy system modelling, public awareness and attitudes towards fusion, energy policies, etc. This work helps to indicate important factors which must be considered in the longer term if fusion is to become a widely-accepted and -deployed future energy source, and therefore provides a view of “market pull” for the requirements the technology programme must meet to be successful.

A Combined Electromagnetic and Mechanical Approach for EU-DEMO Toroidal Field Coils

The roadmap to fusion electricity of the European scientific program aims at the realization of the future DEMOnstration (DEMO) fusion power plant. In 2020, the pre-concept design phase of DEMO was completed, defining the concept and characteristics of the main magnets and structures of the machine. Sixteen toroidal D-shaped magnets, six poloidal annular coils and a central solenoid constitute the functioning system core. The reactor is subjected to huge mechanical loads, mainly due to the Lorentz force produced by the combination of the high magnetic fields and operative currents. As a consequence, the loading conditions are extremely demanding for the structural components, and it is crucial to complete a comprehensive static and fatigue assessment before proceeding with the next design iteration. This work focuses on the electromagnetic and structural analyses performed on the toroidal field coil system and its support structures to present the methodological approach developed. Exploiting the finite element method, a three-dimensional model has been defined to obtain the electromagnetic loads on the main time points of the reference plasma scenario and then transfer them to a related 3D structural model, corresponding to the discretization of the electromagnetic one. The structural model was used to obtain the displacement and stress fields at the various time points to perform the mechanical evaluation as well as the fatigue assessment.

A generalised formulation of G-continuous Bezier elements applied to non-linear MHD simulations

The international tokamak ITER is progressing towards assembly completion and first-plasma operation, which will be a physics and engineering challenge for the fusion community. In the preparation for ITER experimental scenarios, non-linear MHD simulations are playing an essential role to actively understand and predict the behaviour and stability of tokamak plasmas in future fusion power plant. The development of MHD codes like JOREK is a key aspect of this research effort, and provides invaluable insight into the plasma stability and the control of global and localised plasma events, like Edge-Localised-Mode and disruptions. In this paper, we present an operational implementation of a new, generalised formulation of Bezier finite-elements applied to the JOREK code, a significant advancement from the previously G1-continuous bi-cubic Bezier elements. This new mathematical method enables any polynomial order of Bezier elements, with a guarantee of G-continuity at the level of (n−1)/2, for any odd n, where n is the order of the Bezier polynomials. The generalised method is defined, and a rigorous mathematical proof is provided for the G-continuity requirement. Key details on the code implementation are mentioned, together with a suite of tests to demonstrate the mathematical reliability of the finite-element method, as well as the practical usability for typical non-linear tokamak MHD simulations. A demonstration for a state-of-the-art simulation of an Edge-Localised-Mode instability in the future ITER tokamak, with realistic grid geometry, finalises the study.

Predictions of radiation pattern and in–out asymmetries in the DEMO scrape-off layer using fluid neutrals

Strongly radiating and detached high-power discharges in present-day full-metal tokamaks have a characteristic radiation pattern involving condensation of radiation near the X-point, with significant radiative losses above the X-point. In contrast, Demonstration Fusion Power Plant (DEMO) divertor exhaust scoping studies using reduced physics models, including a fluid description for the neutrals, place the strongest radiation fronts in the divertor legs, near the separatrix. The present contribution studies sensitivity of the radiation pattern corresponding to maximal divertor impurity radiation to those physics models that are typically neglected in the simulations due to their computational expense: cross-field drifts, complex impurity models and kinetic neutrals. Model benchmarking is carried out in comparison to L-mode discharges, which are shown to feature both divertor and X-point radiation. The simulated plasma conditions with maximal divertor radiation have in–out asymmetries in the divertor legs and at the divertor entrance, and the asymmetries and the radiation patterns are observed to be sensitive to both cross-field drift effects and the neutral model. DEMO simulations, carried out using SOLPS-ITER, show an impact of cross-field drifts on the divertor asymmetries, but the impact is not large enough to move the radiation front from the divertor legs to regions above the X-point.

Overview of the TCV tokamak experimental programme

The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Overview of the SPARC physics basis towards the exploration of burning-plasma regimes in high-field, compact tokamaks

The SPARC tokamak project, currently in engineering design, aims to achieve breakeven and burning plasma conditions in a compact device, thanks to new developments in high-temperature superconductor technology. With a magnetic field of 12.2 T on axis and 8.7 MA of plasma current, SPARC is predicted to produce 140 MW of fusion power with a plasma gain of Q ≈ 11, providing ample margin with respect to its mission of Q > 2. All tokamak systems are being designed to produce this landmark plasma discharge, thus enabling the study of burning plasma physics and tokamak operations in reactor relevant conditions to pave the way for the design and construction of a compact, high-field fusion power plant. Construction of SPARC is planned to begin by mid-2021.

Design and experimental demonstration of feedback adaptive RMP ELM controller toward complete long pulse ELM suppression on KSTAR

Operation of a fusion power plant requires robust edge localized mode (ELM) suppression simultaneously with high plasma performance. In this paper, we describe a novel feedback adaptive resonant magnetic perturbation (RMP) ELM controller designed to address this problem by achieving optimized ELM suppression through the advanced application of 3D RMPs. From real-time Dα data, the controller is able to achieve robust ELM suppression while simultaneously minimizing the applied RMP in order to enhance plasma performance. In real-time, the instantaneous ELM-frequency is analyzed with an adaptive feedback algorithm to determine amplitudes and phases of RMP coil currents that will maximize plasma performance while maintaining ELM suppression. When applied through the KSTAR plasma control system in several experiments using n = 1 RMPs, robust ELM suppression is achieved and sustained in feedback while reducing the RMP strength to ∼65% of its initial value. Minimization of the RMP strength in this manner not only allows for operation of longer discharges due to a decrease in flux consumption but also allows for a strong recovery of up to ∼60% of βN throughout the ELM-free period.

A Model to Simulate Gas Dissolution into/through Metals and Its Application to Deuterium in a 316L Steel Chamber with Pb-Li in a Quasi-2D Geometry

Liquid lead-lithium in eutectic proportions (Pb-Li) is a candidate material for Breeding Blankets (BBs) in future Fusion Power Plants (FPP). BB design depends on the diffusivity and Sieverts’ constant (solubility) of tritium in this alloy, but literature reports a large scattering of measurements for these values. A model was developed to address one possible source of this scattering in static experiments, i.e., non-negligible loss of hydrogen gas through steel walls of containers. This model simulates the dissolution of gases into, and their diffusion through, metallic barriers for diffusivity and Sieverts’ constant as inputs. When implemented, it can be used to compute the pressure decrease in a metallic chamber, and comparison of simulated curves with experimental ones allows for estimates of the diffusivity and Sieverts’ constant. This approach was used to estimate these coefficients for deuterium in stainless steel, using experiments performed with a 316L steel chamber from an existing facility (the Vacuum Sieve Tray setup) and simulations in a quasi-2D representation of this chamber. This validated the model, which was then used to simulate the chamber containing Pb-Li, as a means of planning for future experiments.

Integrated design of tokamak building concepts including ex-vessel maintenance

The EU-DEMO (referred to as DEMO) tokamak complex currently consists of three buildings, like in ITER: the tokamak building, the tritium building and the diagnostics building. The tokamak building houses the tokamak itself and the numerous plant systems that interface with the necessary systems to produce and control the plasma. It is designed to permit assembly, operation and maintenance of the DEMO tokamak. The vacuum vessel is organized in 8 sectors, with radial ports at the lower and equatorial levels and one vertical upper port. The general architectural structure is arranged around the tokamak with a cylindrical bioshield of 2m thickness around the cryostat and floor levels corresponding to the cryostat penetrations. Additional levels are used for the integration of auxiliary equipment for the various plant systems and for accident mitigation systems. Currently, the tokamak building also represents the final nuclear confinement barrier for the radioactive material towards the environment and the public. This safety function requires the plant and safety systems to limit the release of radioactive substances during normal operation and in accidental conditions well below the safety limits. The complexity of a fusion power plant, like DEMO, with regards to integration of plant systems is much higher than that in a fission power plant due to the larger number of plant systems. Three main criteria drive the integration work of the plant systems inside the tokamak building: (i) safety requirements, (ii) functional requirements of the plant systems themselves and (iii) the maintenance approach. Cost considerations are also taken into account together with the normal and accidental environmental conditions in the various areas of the tokamak building that might challenge the qualifications of structures, systems and components (SSC). The layout of the tokamak building has to be further developed in the Concept Design Phase to follow the plant design evolution providing feedback to the various designers in order to assure that DEMO meets the cited design criteria with an optimized and licensable layout of the most complex nuclear building.

Plasma Scenarios for the DTT Tokamak with Optimized Poloidal Field Coil Current Waveforms

In the field of nuclear fusion, the power exhaust problem is still an open issue and represents one of the biggest problems for the realization of a commercial fusion power plant. According to the “European Fusion Roadmap”, a dedicated facility able to investigate possible solutions to heat exhaust is mandatory. For this purpose, the mission of the Divertor Tokamak Test (DTT) tokamak is the study of different solutions for the divertor. This paper presents the plasma scenarios for standard and alternative configurations in DTT. The Single Null scenario is described in detail. The alternative configurations are also presented, showing the good flexibility of the machine.