Pulsed Tokamak Research Articles

Design, construction and test of a prototype coil with demountable CuBe joints for pulsed table-top tokamaks

In a classic small tokamak, the magnetic confinement is achieved by toroidal coils that fit into a donut-shape mechanical structure. To facilitate their handling and maintenance as well as the access to various auxiliaries components, their design has evolved toward a modular configuration. For such demountable design, it is important to guarantee the electrical continuity of the winding of the coil across the modular parts through electrical contacts. The resulting joints rely on pressure and specific materials to transfer high current densities at a good mechanical stability and manageable losses. In the present work, a prototype of a circular demountable toroidal field coil (CDTFC) was designed and built to test some technical choices to be used in the final coils of a table-top tokamak referred to as TPM-1U. For this small tokamak, the magnet wires to wind the toroidal coils should be able to handle large pulsed currents in the tens of thousands of Amperes whereas the coil joints should handle a peak current density of at least 75 kA/cm2. The case study here is the TPM-U1 table-top tokamak. Its conceptual design is recalled and mechanical and magnetic details of the demountable coil prototype are provided. For the experimental test, the contacts were slightly pressed using a custom-made contact assembly and current pulses between 4 kA and 8 kA were fedto the coil prototype. The degradation of the joint was visually estimated by assessing the amount of micro-melting, plastic deformation and oxidation appearing at the interface between the materials in contact. It is shown that the selected AWG-04 magnet wire is appropriate for carrying kiloAmpere range pulsed currents and that the CuBe contacts are able to withstand current densities larger than the specification, up to 263 kA/cm2, under low contact pressures of a few mega-Pascal simplifying their handling. Some technical choices have been probed on a prototype coil yielding a generic sequence of tasks to build and test mechanically and electromagnetically demountable coils and their joints.

Postmortem assessment of the WEST Fiber Bragg grating and thermocouple diagnostic performances with HADES

Plasma facing components temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. During the WEST phase 1, a set of 20 thermocouples (TC) and 4 fibers Bragg grating (FBG) have been exposed to five experimental campaigns with 3889 plasmas for a total plasma duration of about 7h15 and 21.2 GJ of cumulated injected energy. This paper presents the performances in terms of response time and high temperature sustaining assess with high heat flux tests performed in the HADES facility. The response time of the TC remains at the same level as before their installation in the divertor. However, the FBG have shown clear increase of their response time depending on the location due to a degradation of the thermal contact of the graphite adhesive and the FBG or component. Finally, one regenerated FBG has been exposed to high temperature, showing that absolute temperature measurement is found reliable up to 1100 °C with the start of some spectrum degradations. Full erasure of the grating is obtained after 10 s at 1200 °C well above the stable temperature of 900 °C of the regenerated FBG.

Real-time plasma horizontal displacement estimator based on the neural network

Plasma displacement diagnostics is one of the basic diagnostics for evaluating the plasma operating state in tokamak devices. It is related to the plasma current centroid and generally obtained by integrating and calculating the induced voltage in the displacement coils. This induced voltage is derived from the magnetic field generated by the plasma. If the induced voltage is not solely due to the plasma magnetic field, there will be an unreal deviation in the plasma displacement. In J-TEXT, this phenomenon usually occurs when the Resonant Magnetic Perturbation (RMP) coils close to the displacement coils are energized. For long pulse tokamak devices, the source of the unreal deviation may be drifting integrators or radiation-induced voltages in the magnetic pickup coils, not only the RMP coils. To solve this problem, this paper proposes a real-time non-magnetic plasma horizontal displacement measurement method using a deep fully connected neural network. The soft-X ray (SXR) diagnostic data and the plasma displacement diagnostic data are used as the training data for the network. The training and testing results show that the network obtains a satisfactory fitting effect. On the other hand, to deploy this network for real-time diagnostics, this paper has designed and developed a plasma horizontal displacement estimator based on J-TEXT real-time framework (JRTF) in J-TEXT. The experimental results show that the real-time estimator is sufficient for real-time plasma displacement diagnostics. The feasibility of establishing real-time plasma displacement diagnostics based on the SXR diagnostic system has been verified.

Power conversion from spherical tokamak test reactor with helium-cooled and water-cooled blanket

Possible configurations for Power Conversion System (PCS) for a futuristic, small, pulsed Spherical Tokamak (ST) fusion reactor are presented. For the typical blanket coolant choices namely, helium and water, an intermediate heat-exchanger (IHX) cum buffer energy storage system (ESS) with Molten Salt (MS) is introduced to examine Rankine and Brayton cycle’s effectiveness. Preliminary thermal sizing of the involved heat-exchangers and parameters of the optimized cycle are discussed. S-CO2 Brayton cycle seems to be a promising candidate for efficient power conversion for a pulsed plasma reactor with efficiency of 42.75%. Rankine cycle yields efficiency of 35–42% depending on water/helium as the choice of the blanket coolant.

First temperature database achieved with Fiber Bragg Grating sensors in uncooled plasma facing components of the WEST lower divertor

Plasma Facing Components (PFCs) temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. A set of twenty thermocouples (TCs) and four optical fiber temperature sensing probes, each of them including eleven wavelength-multiplexed fiber Bragg gratings (FBGs) written with UV radiation, have been integrated and deployed in the WEST lower divertor. To avoid any thermal erasure occurring above 400 °C, the gratings have been regenerated using a high temperature annealing process. The FBG probes are embedded in tungsten-coated graphite components with no active cooling. The diagnostic was working correctly from the first plasma breakdown achieved in WEST (in 2017) until the first dismantling of the lower divertor three years later (2020). Collected data are sorted by temperature range to investigate the ageing of both PFCs and diagnostics. Measurements obtained with FBGs and TCs are consistent, no long-term drift is reported during the overall period of use. The maximum temperature reported by FBGs in the divertor is 830 °C obtained with 90 MJ of cumulated plasma energy in one pulse. A short but reversible collapse of the Bragg peak (from 5 down to 2 dB) has been observed under high temperature due to the strong temperature gradients measured along the fiber (about 20 °C/mm, for an FBG size of 3 mm). This feedback demonstrates that regenerated FBGs are suitable for long-term temperature monitoring in tokamak environment with high heat load, high temperature and gradient, strong magnetic field and vacuum constraint.

Steady state versus pulsed tokamak reactors

We have carried out a detailed analysis that compares steady state versus pulsed tokamak reactors. The motivations are as follows. Steady state current drive has turned out to be more difficult than expected—it takes too many watts to drive an ampere, which has a negative effect on power balance and economics. This is partially compensated by the recent development of high temperature REBCO superconductors, which offers the promise of more compact, lower cost tokamak reactors, both steady state and pulsed. Of renewed interest is the reduction in size of pulsed reactors because of the possibility of higher field Ohmic transformers for a given required pulse length. Our main conclusion is that pulsed reactors may indeed be competitive with steady state reactors and this issue should be re-examined with more detailed engineering level studies.

Backwards extrapolation activation diagnostics and their dynamic range for pulsed neutron source measurements

Activation materials implanted within radiation detectors can be used to measure pulsed neutron fields. This work develops an instrument concept with the aim to maximise sensitivity to pulsed fusion neutron fields and, using a data-rejection algorithm combined with backwards extrapolation, enable neutron fluence estimates to be made over a large dynamic range. Through high-fidelity modelling of residual temporal emissions, and a parameterised approach, we study the sensitivity to neutrons of a plastic scintillator–Ag foil layer detector concept. For an optimal design we apply paralysable and non-paralysable deadtime models to the predicted response to D–D fusion neutron fields at various neutron field intensities. In high neutron fluence irradiation scenarios, where deadtime effects are strongly evident, we use our approach to make estimates of the fluence from instrument response data. We discuss the practical applications of such diagnostics used for plasma focus (PF) fusion experiments, such as at the PF-1000U facility in IPPLM, Poland, inertial confinement fusion and pulsed tokamak experiments, for example at MAST-U, where such diagnostics could complement fission counter-based neutron diagnostics in the future. Finally, we show that the calibration of such detection systems may be achieved using relatively low emission rate, steady state neutron sources, with calibration factors that are straightforward to apply to pulsed neutron field measurements.

Progress in the Design of a Hybrid HTS-Nb3Sn-NbTi Central Solenoid for the EU DEMO

State-of-the-art high field solenoids make use of hybrid designs exploiting the superior high field performance of High Temperature Superconductors (HTS) in the innermost region. The benefits of a hybrid Central Solenoid in a pulsed tokamak like the EU DEMO can be two-fold: either to reduce its outer radius (which would result in a reduced overall size and cost of the tokamak), or to increase the generated magnetic flux (which could extend the plasma burn time and possibly increase the power plant efficiency). In the framework of the pre-conceptual design studies for DEMO coordinated by EUROfusion, a hybrid Central Solenoid is proposed based on ten layer-wound sub-coils using HTS, Nb3Sn, and Nb-Ti conductors respectively for the high, medium, and low field sections. The design exploits the flexibility of layer winding by grading both the superconductor and the stainless steel cross sections in each sub-coil, which has the potential for space and cost savings. Mechanical analyses have identified fatigue as the main design driver for the EU DEMO Central Solenoid. Possible alternatives to reduce the sensitivity of the proposed design to fatigue are currently under investigation.

Initial Exploration of High-Field Pulsed Stellarator Approach to Ignition Experiments

In the framework of fusion energy research based on magnetic confinement, pulsed high-field tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a high-field stellarator approach, and, (2) proposing a preliminary high-field stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finite-element calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wide-ranging parameters. Based on these studies, and on assumptions on the enhancement-factor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B ~ 10 T and 30 m3 plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable cross-section, detachable periods, and a lithium wall and divertor.

Integration of fiber Bragg grating temperature sensors in plasma facing components of the WEST tokamak.

Plasma Facing Components (PFC) temperature measurement is mandatory to ensure safe high power and long pulse tokamak operation. IR thermography systems which are widely used in magnetic fusions devices become challenged with the choice of tungsten as a PFC material in the ITER tokamak, mainly due to emissivity uncertainties and reflection issues in a hot environment. Embedded temperature measurements are foreseen to cross-check the IR thermography measurements. Fiber Bragg grating sensors are potentially of great interest for this application because they are immune to electromagnetic interference and allow the measurement of a large number of temperature spots on a single fiber. Four optical fiber temperature sensing probes, each of them including 11 regenerated fiber Bragg gratings equally spaced by 12.5 mm (equivalent to one ITER-like tungsten monoblock), have been specifically designed and manufactured for the WEST project (W-tungsten Environment and Steady State Tokamak). The four probes are embedded in W-coated graphite components at two different distances from the surface, 3.5 mm and 7 mm, to cover a wide range of temperatures up to 900 °C. This paper addresses the design and integration issues and the qualification and performance assessment performed in the laboratory. It also shows the first measurements of this new diagnostic achieved in a tokamak environment during baking of the machine and during early diverted plasma exposure.

Systems engineering perspective to the integration of the heating and current drive system in the EU DEMO: Analysis of requirements and functions

In the framework of the EUROfusion Consortium, the Work Package Heating and Current Drive (WPHCD) is undertaking the development and design of the DEMO Heating and Current Drive (HCD) system, supported by the Work Package Plant Level Systems Engineering, Design Integration and Physics Integration (WPPMI). As the DEMO project is in the Pre-Conceptual phase, its development is dominated by uncertainties on system requirements, and by different DEMO designs and scenarios. In this context, a number of outstanding technology and physics issues, e.g. related to safety, neutronics, performance and reliability, need to be resolved using an integrated approach.For coping with these boundary conditions, we present an initial analysis of HCD requirements and functions through the development of a System Architecture model, i.e. through describing the HCD system in terms of its functions and their interdependencies. Scope of the work is to address uncertainties i) associated with different reference scenarios (e.g. pulsed and steady-state tokamak) and operational concepts; ii) related to system performances. The assessment is part of integration activities, consisting of both systems engineering methodologies and design analysis, all aiming to ensure consistency in the overall EU DEMO plant design.

Systems code studies on the optimization of design parameters for a pulsed DEMO tokamak reactor

In the European strategy towards fusion power, a demonstration tokamak fusion reactor (DEMO) is foreseen as the next single step between ITER and a power plant. The current baseline concept is a tokamak reactor with net electrical output power of Pel∼500MW and plasma pulse duration of tpulse∼2h. Systems codes are commonly used in the design process as numerical tools for optimization studies. The key performance data of the reactor such as Pel and tpulse are depending on a variety of design and plasma parameters. In the application of systems codes within this multi-dimensional parameter space, a clear quantitative understanding of the most suitable optimization criteria has to be developed, and various physics and technology limits should be obeyed to obtain meaningful results.In this work we use a fusion reactor systems code to perform parameter variations for a pulsed DEMO tokamak reactor. Various output quantities are presented as a basis for the quantitative assessment of the numerical results, and options for a further development of the current DEMO baseline design are proposed and briefly discussed.

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.

Consequences of the technology survey and gap analysis on the EU DEMO R&D programme in tritium, matter injection and vacuum

In the framework of the EUROfusion Programme, EU is preparing the conceptual design of the inner fuel cycle of a pulsed tokamak DEMO. This paper illustrates a quantified process to shape a R&D programme that exploits as much as possible previous R&D. In an initial step, the high-level requirements are collected and a novel DEMO inner fuel cycle architecture with its three sub-systems vacuum pumping, matter injection (fuelling and injection of plasma enhancement gases) and tritium systems (tritium plant and breeder coolant purification) is delineated, driven by the DEMO key challenge to reduce tritium inventory. Then, a technology survey is carried out to review potential existing solutions for the required process functions and to assess their maturity and risks. Finally, a decision-making scheme is applied to select the most promising candidates. ITER technology is exploited where possible. As a primary result, a fuel cycle architecture is suggested with an advanced tritium plant that avoids full isotope separation in the main loop and with a Direct Internal Recycling path in the vacuum systems to shorten cycle times. For core fuelling, classical inboard pellet injection technology is selected, in principle similar to that proposed for ITER but aiming for higher launch speeds to achieve deep fuelling of the DEMO plasma. Based on these findings, a tailored R&D programme is shaped that tackles the key questions until 2020.

Diagnostics and control for the steady state and pulsed tokamak DEMO

The present paper is devoted to a first assessment of the DEMO diagnostics systems and controls in the context of pulsed and steady state reactor design under study in Europe. In particular, the main arguments treated are: (i) The quantities to be measured in DEMO and the requirements for the measurements; (ii) the present capability of the diagnostic and control technology, determining the most urgent gaps, and (iii) the program and strategy of the research and development (R&D) needed to fill the gaps. Burn control, magnetohydrodynamic stability, and basic machine protection require improvements to the ITER technology, and moderated efforts in R&D can be dedicated to infrared diagnostics (reflectometry, electron cyclotron emission, polarimetry) and neutron diagnostics. Metallic Hall sensors appear to be a promising candidate for magnetic measurements in the high neutron fluence and long/steady state discharges of DEMO.

Applications of Bayesian temperature profile reconstruction to automated comparison with heat transport models and uncertainty quantification of current diffusion

In the context of present and future long pulse tokamak experiments yielding a growing size of measured data per pulse, automating data consistency analysis and comparisons of measurements with models is a critical matter. To address these issues, the present work describes an expert system that carries out in an integrated and fully automated way (i) a reconstruction of plasma profiles from the measurements, using Bayesian analysis (ii) a prediction of the reconstructed quantities, according to some models and (iii) a comparison of the first two steps. The first application shown is devoted to the development of an automated comparison method between the experimental plasma profiles reconstructed using Bayesian methods and time dependent solutions of the transport equations. The method was applied to model validation of a simple heat transport model with three radial shape options. It has been tested on a database of 21 Tore Supra and 14 JET shots. The second application aims at quantifying uncertainties due to the electron temperature profile in current diffusion simulations. A systematic reconstruction of the Ne, Te, Ti profiles was first carried out for all time slices of the pulse. The Bayesian 95% highest probability intervals on the Te profile reconstruction were then used for (i) data consistency check of the flux consumption and (ii) defining a confidence interval for the current profile simulation. The method has been applied to one Tore Supra pulse and one JET pulse.

Modelling of pulsed and steady-state DEMO scenarios

Scenario modelling for the demonstration fusion reactor (DEMO) has been carried out using a variety of simulation codes. Two DEMO concepts have been analysed: a pulsed tokamak, characterized by rather conventional physics and technology assumptions (DEMO1) and a steady-state tokamak, with moderately advanced physics and technology assumptions (DEMO2). Sensitivity to impurity concentrations, radiation, and heat transport models has been investigated. For DEMO2, the impact of current driven non-inductively by neutral beams has been studied by full Monte Carlo simulations of the fast ion distribution. The results obtained are a part of a more extensive research and development (R&D) effort carried out in the EU in order to develop a viable option for a DEMO reactor, to be adopted after ITER for fusion energy research.

SST-1 Gas feed and Gas Exhaust system

SST-1 tokamak is a long pulse tokamak designed for the plasma operation up to 1000 sec duration. Gas feed system and gas exhaust management will play a very crucial role during plasma discharge. During the different type of operations of tokamak like wall conditioning, diverter operation and neutral beam injection, a large amount of gas will be fed into the vacuum chamber at different locations. Also during plasma operations, the gas will be fed both in continues and pulse mode. Gas feed will be carried out mainly using piezo-electric valves controlled by PXI based data acquisition and control system. Such operations will lead to a huge amount gas exhaust by the main system which requires good exhaust facility to searches, great care should be taken in constructing both. Also initial pumping of cryostat and vacuum vessel of SST-1 will release a large amount of gas. Exhausted gases from SST -1 will be Hydrogen, Nitrogen, Mixture gases or some toxic gases. Dedicated exhaust system controlling the different gases are installed. Special treatment of hazardous/explosive gases is done before releasing to the atmosphere. This paper describes design and implementations of the complete gas feed and exhaust system of SST-1.

0-D thermo-hydraulic approach for predicting pressure and temperature along HELIOS SHe closed loop under pulsed loads

Cryogenic systems for future large superconducting tokamaks (e.g. JT-60SA or ITER) are expected to cope with large pulsed heat loads due to cycling plasma operation. Their superconducting magnets are cooled down with forced flow supercritical helium.The aim of this paper is to verify to what extent a 0-D thermo-hydraulic model can well reproduce in space and time, the variations of pressure and temperature along a supercritical helium closed loop, subjected to pulsed heat loads.A 0-D model has been developed and the paper will present the corresponding equations and assumptions will also be justified. A pulsed heat load tokamak relevant scenario has been tested and the resulting variations of pressure and temperature have been compared with experimental data. The results of the 0-D model demonstrate the relevance of such approach for predicting transient behaviors in response of pulsed heat loads in a closed loop.This simple approach is also a justification to use process modeling codes where dynamics of the cryogenic circuits can be simulated with cryogenics components.

Stress and fatigue in the magnets in DEMO in pulsed and steady-state operation using geometric scaling factors

One of the most important lifetime-limiting effects for a pulsed tokamak reactor, the effect due to pulsing the magnets, is addressed here. Pulsed operation imposes stresses with significant alternating components on all the coils. This paper analyses the effects of fatigue on the coils. (Fatigue of the superconductor itself is not discussed.) Initial work has concentrated on the TF coils. The form of the finite element model is simple to permit parametrisation so that the results can be used in systems studies using codes such as PROCESS. The parameters used so far are geometric scaling factors describing the TF and CS coils only. Once sufficient parameters have been introduced, the resulting equations for stress can give realistic results in a systems code, which allows for the interdependence of all the variables. The magnetic field in the TF coil winding pack due to all the coils is calculated using the Biot–Savart law. This gives the Lorentz force in the winding pack, which is then used for the stress–strain calculation. The alternating stress is calculated by subtracting the stress tensors between two different load cases. In order to meet the lifetime fatigue goal the coil support structure needs to be made stronger than the structure intended for the small number of cycles of a steady-state reactor design, which will add to its radial build. We have quantified these requirements, allowing systems codes to estimate the additional costs.