Component Test Facility Research Articles

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities

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.

Bayesian Parameter Estimation of the k-ω Shear Stress Transport Model for Accurate Simulations of Impinging-Jet Heat Transfer

Given the lack of fusion-relevant component test facilities, current estimates of the thermo-fluid performance of plasma-facing components are based for the most part on numerical simulations. A major source of uncertainty in these simulations is the semiempirical turbulence (closure) models for the Reynolds stresses appearing in the governing Reynolds-averaged Navier-Stokes equations, which involve a set of constants that depend upon the flow. The objective of this study is to evaluate Bayesian parameter estimation of turbulence closure constants in ANSYS Fluent to model heat transfer in impinging jets. The Bayesian statistical calibration produces a probability distribution for these constants from experimental data; the maximum a posteriori estimates are then taken to be the calibrated constants, or parameters. The turbulence model constants are calibrated using an experimental study of a submerged jet of air impinging on a flat heated surface at Reynolds numbers Re = O(104) and impingement distance in jet diameters H/d = 2. Numerical predictions using the calibrated model parameters are then compared with those generated using the default constants. Predictions obtained with model parameters calibrated on datasets of two different sizes are compared to evaluate the effect of the number of calibration samples. Finally, the extrapolative ability of the calibrated model is examined by predictions at a Re beyond the calibration values.

Lean-Burn Combustor Simulator for an Engine-Component Test Facility: An Experimental and Computational Study

Abstract The computational and experimental assessment of a lean-burn low-NOx combustor simulator for an engine component test facility is presented. The Engine Component Aero-Thermal (ECAT) facility is a full-scale engine-parts facility, designed for the study of the aero-thermal performance of fully cooled high-pressure nozzle guide vanes (NGVs). The facility operates with non-dimensionally matched engine conditions in terms of Reynolds number, Mach number, and coolant-to-mainstream pressure ratio. The combustor simulator is designed to replicate lean-burn conditions of swirl and temperature distortion upstream of the nozzle guide vanes. The purpose is to allow the study of flow capacity, aerodynamic performance (with film cooling), and thermal performance (overall effectiveness) in the presence of engine-realistic inlet distortions. The ECAT combustor simulator design builds on the work of Hall and Povey, who developed a full-scale low-speed atmospheric-pressure combustor simulator for the preliminary design of similar simulators later implemented in the Oxford Turbine Research Facility and ECAT. This pilot facility produced non-dimensional lean-burn combustor-exit conditions closely matched to target profiles representative of modern aero engines. This design was modified and scaled for compatibility with the annulus-line and higher-pressure operating conditions of the ECAT facility. In the present study, detailed experimental measurements with multi-hole probes and thermocouples (pressure profile and temperature field) are presented and compared to results from Reynolds-Averaged Navier–Stokes Simulations. Additional simulations were performed to understand how the elevated back-pressure and vane potential field affect the non-dimensional profiles of pressure loss, residual swirl, and temperature at the combustor–turbine interface. This is perhaps the most comprehensive study to date of a combustor simulator in an engine-scale research facility, providing a unique insight into the known challenges of simulator design, scaling issues when moving from low to high Reynolds number, and limitations of computational fluid dynamics in this flow environment. The results, which will serve as boundary conditions to understand the impact of this flow on components in future studies in this facility, demonstrate the fidelity with which lean-burn target conditions can be replicated in a non-reacting environment.

A Transient 3-D CFD Model for the Simulation of Forced or Natural Convection of the EU DEMO In-Vessel Components

As the EU DEMO reactor will act as a Component Test Facility for the breeding blanket (BB), it is foreseen that the different BB concepts will be tested throughout the plant’s lifetime. The maintenance of all the in-vessel components (IVCs), as for all D-T fusion machines, must be carried out employing remote handling (RH) technology, as the structural materials will be activated by the neutrons. The maintained segment and possibly other nearby segments cannot be actively cooled and will heat up due to the decay heat. For these reasons, alternative cooling strategies need thus to be investigated to ensure that the BB segment will cool down within the limits required by the RH in a reasonable amount of time. In the present work, two possible cooling options are investigated for the case of the Water-Cooled Lithium-Lead BB concept. One is based on the passive cool-down by natural convection of the BB segments, whereas the second one relies on a forcing flow of cool air on the BB surfaces. A computational fluid dynamics (CFD) approach has been used to study the different options for performing transient analyses through the Star-CCM+ commercial code.

Novel features of the helical volumetric neutron source FFHR-b2

The compact helical fusion reactor FFHR-b2 is a multipurpose HElical Volumetric Neutron Source available as the component test facility to test the divertor and blanket systems. In the latest design, three new technologies of the high-temperature superconducting conductor, the ceramic pebble divertor, and the cartridge-type blanket are adopted. Tin-based functional liquid metals (FLMs) including lithium are considered as the working fluid for the blanket. The FLMs have multi functions of low vapor pressure, low density, low melting point, low corrosiveness, high tritium breeding ratio, and so on. The target construction cost of the FFHR-b2 has been set to 200 billion JPY. The main target of FFHR-b2 is to demonstrate the fusion gain larger than one. To achieve this in a relatively small device, the tangential neutral beam injection (NBI) heating with a moderate energy of 80 keV is adopted for the FFHR-b2. The beam-plasma fusion with 5 MW of NBI heating can produce a fusion power of 5 MW that is larger than the absorbed power of 3 MW.

Performance optimization of test facility for coaxial transmission line components based on traveling wave resonator.

As part of development program for a high power co-axial transmission line component test facility, an existing traveling wave resonator based test stand is modified to improve power gain and ring return loss. The 10 dB directional coupler in the earlier test stand is replaced with a 14 dB directional coupler to couple radio frequency power with the ring. To achieve an improved isolation and return loss, the 14 dB directional coupler design is equipped with two broadside strip-lines with a tunable gap between them. Detailed design and optimization of the 14 dB directional coupler with and without the traveling wave resonator setup is performed using a high frequency simulator Computer Simulation Technology Microwave Studio. The low power test of the fabricated directional coupler is performed at several tuning positions to achieve an optimum operating frequency for the traveling wave resonator. Furthermore, after optimization, the maximum power gain of around 18 dB and minimum return loss of about -22 dB inside the ring are obtained. Finally, a preliminary study of the future 3 MW test facility is discussed.

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.

Recent Developments in 3D Multiphysics Modelling of Whole Fuel Cell Systems for Assisting Commercialisation and Improved Reliability

3D multiphysics modelling of fuel cell components and fuel cell systems has been a core research field since many years. The simulation assisted research and development improves the understanding and overall success of the fuel cell technology, in particular the high temperature SOFC technology. The ultimate goal has been to support the commercialisation of the technology. This can be enhanced by improving the understanding and knowledge of the utilised fuel cell systems. Especially, gaining detailed information about their durability and life duration is of paramount importance. As a fuel cell system consists of many components like the fuel cell stack, reformer, afterburner, heat exchanger etc., it is important to account for the coupled processes, occurring within each component and their interaction with the neighbour components. This needs to be mastered, in order to improve the overall technology. However, this requires a large amount of experimental facilities, which have also limits. Even the presence of long-term testing results of individual solid oxide fuel cells are not sufficient yet to declare their autonomous performance as the same within the system. Moreover, many years are required that is constraining the scientists; thus, the assistance of 3D multiphysics modelling is attractive. If we consider that the technology has heating-up, operation and shut-down stages, broad knowledge is required for a safe and reliable SOFC development. For this purpose, the developed full scale multiphysics model reflecting a 5kW full integrated SOFC module is employed for the solution of the complex 3D problem and to gather invaluable information about various questions. The used numerical model of the complex 3D system depicts the true physical resolution of the components, as to account for the complex geometrical effects. Material non-linearity, including the temperature dependence of material properties, as well as the material models that describe the elastic, plastic and creep deformation of the materials are implemented within the model. Each component process and parameters are determined from component and system test facilities. To shed light to the long-term system behaviour, recent attention has been given to the steady operation and the durability response of the interacting components. The detailed 3D stress-strain field of the whole system could be analysed and invaluable practical information about the critical locations within the interacting whole system could be determined. To interpret the long-term reliability of the system, the thermomechanical fatigue behaviour of the critical system regions have been investigated in detail to predict and understand the durability under cyclic load. The results give a flair of the life expectation of a system subjected to current circumstances.

Analytical performance assessment of a novel active mooring system for load reduction in marine energy converters

Reliability and storm survival of Marine Energy Converters are critical to their commercial development and deployment. The Intelligent Active Mooring System (IAMS) is a novel device intended to minimise extreme and fatigue loading in mooring lines through a non-linear load–extension curve that is variable in operation to adjust to the prevailing metocean conditions. An analytical model of IAMS, validated by physical model tests at the Dynamic Marine Component test facility at the University of Exeter, is used in a numerical simulation of the performance of IAMS as part of the mooring system of the South West Mooring Test Facility buoy. A 10m length of IAMS can reduce the rms line tension in normal operating conditions by up to 21% and the peak line tension in storm conditions by up to 21% when compared to braided nylon mooring lines. Peak line tension reductions of over 50% can be achieved if a longer IAMS unit is used. The resulting mooring system can be optimised to give load reductions in a wide range of metocean conditions; while variable pre-tension could be used for tidal range compensation or to ease access for installation and maintenance.

Development of data acquisition and control system for ICH & CD transmission line components test facility

India is developing a very high power test bed (∼MW level) based on concept of Travelling Wave Resonator (TWR) for testing of passive transmission line components [3]. Data acquisition and control system is required for safe and reliable operation of TWR. It includes online monitoring of input power and circulating ring power. Acquired RF signal provides information about forward and reflected RF power. For MW power level testing, system is protected against arc and thermal effect during CW operation. National Instrument make hardware and software have been used to acquire signals from RF detector, Arc detectors, IR camera and thermocouples etc. LabVIEW™ based software has been developed for calculation of RF parameters. Graphical user interface is developed for better visualization. Initial testing of the TWR setup with 10 dB coupler provides a power gain of 13.58 dB (22X). For future 3 MW TWR test bed the required power gain is ∼20 dB (100X) since the available source is ∼40 kW. This paper describes the detail design and integrated test results for data acquisition and control system for TWR test bed.

Design and installation of a ferromagnetic wall in tokamak geometry.

Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak-Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability.

Conceptual design study of the K-DEMO magnet system

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8m and 2.1m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb3Sn superconducting conductor. The peak magnetic field reaches to 16T with the magnetic field at the plasma center above 7T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

A preliminary conceptual design study for Korean fusion DEMO reactor

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

System analysis study for Korean fusion DEMO reactor

A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8T is expected to be achieved by using current state-of-the-art high performance Nb3Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300MWe can be achieved below the optimistic βN limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described.

MAST: Results and Upgrade Activities

MAST, alongside other spherical Tokamaks (STs), provides new perspectives on Tokamak physics for ITER and beyond and a platform to explore the potential of the ST as the core of a Component Test Facility (CTF) to test and develop components and technology for fusion power plants. MAST is one of the two largest STs in the world, the other being NSTX (PPPL, U.S.). Recent physics results include new measurements of the edge pedestal, the use of magnetic perturbations to influence edge-localized modes, pellet fueling, H-mode access, core transport, confinement scaling, fast-particle physics, and MHD instabilities. The program makes use of a set of unusually high resolution and wide-viewing diagnostics on MAST, since, in most experiments, the focus is on understanding the mechanisms, to support predictive modeling and comparison with other types of Tokamak. The first stage of a major upgrade is under way which will raise the heating power and toroidal field by 50% and almost double the inductive flux swing to allow long pulses and sustained noninductively driven plasmas. In particular, 17 poloidal field coils and an in-vessel cryopump will be added to allow an expanded divertor based on the “super-X” concept, as well as a conventional configuration. These upgrades will transform the operating space, pulselength, and flexibility of MAST for goal-oriented research and as a user facility. The heating and pulselength upgrades are aimed at testing off-axis current drive and fast-particle effects (important for sustained Tokamak fusion plasmas) and developing steady-state scenarios for a CTF. The divertor development is part of a growing international effort to find solutions for the power exhaust which ease the technology and material advances needed for the divertor plasma-facing components in a fusion power plant.

The effects of sheared toroidal rotation on stability limits in tokamak plasmas

Sheared toroidal rotation is found to increase the ideal external kink stability limit, thought to be the ultimate performance limit in fusion tokamaks. However, at rotation speeds approaching a significant fraction of the Alfvén speed, the toroidal rotation shear drives a Kelvin–Helmholtz-like global plasma instability. Optimizing the rotation profile to maximize the pressure before encountering external kink modes, but simultaneously avoiding flow-driven instabilities, can lead to a window of stability that might be attractive for operating future high-performance fusion devices such as a spherical tokamak component test facility.

Overview of physics results from MAST

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

Fusion Nuclear Science Facility (FNSF) before Upgrade to Component Test Facility (CTF)

The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

Ideal MHD stability of a spherical tokamak power plant and a component test facility

We have investigated ideal MHD stability of two advanced spherical tokamak confinement concepts: the spherical tokamak power plant (STPP), a 3 GW concept fusion power plasma producing 1 GW of electric power, and the component test facility (CTF), a concept designed for in situ materials testing for ITER and beyond. Detailed stability studies for toroidal mode number n = 1, 2, 3 displacements are presented as a function of conformal wall radius Rw and on-axis safety factor q0. For the STPP marginal stability scans held the current profile fixed, but varied the total plasma current. For the CTF we have extended and parallelized earlier marginal stability scans to scan over both the plasma beta and q0 by varying the current profile to preserve the total plasma current. These confirm that both concepts are stable provided that the wall is sufficiently close and q0 sufficiently large (q0 > 2.8 for the power plant and q0 > 2.1 for the CTF). Both power plant and CTF configurations are found to be ballooning stable.