ITER is probably the most ambitious energy project in the world today, whose main objective is demonstrating the scientific and technical feasibility of nuclear fusion as an energy source. 35 nations are collaborating to build the world's largest tokamak fusion reactor, a magnetic fusion device that has been designed to prove the feasibility of fusion as a large-scale and carbon-free source of...
The Linear IFMIF (International Fusion Materials Irradiation Facility) Prototype Accelerator (LIPAc) is a key activity to demonstrate the validity of the low energy section of an IFMIF deuteron accelerator up to 9 MeV with a beam current of 125 mA in CW. For the successful acceleration of the high power beam, the input beam to the 5 MeV LIPAc RFQ should be fully characterized and controlled to...
The divertor which is used to discharge reaction energy is the core component of the EAST Tokamak. The divertor is cooled by cooling water when the EAST machine is in operation. In order to prevent the cooling water from corroding the components, and to avoid the uneven baking temperature caused by the cooling water residue during the next baking, the cooling water inside the divertor pipe...
A new Deuterium-Tritium campaign (DTE2) is planned at JET in 2019/20, with a proposed 14 MeV neutron budget nearly an order of magnitude higher than any previous DT campaigns. With this proposed budget, the achievable neutron fluence on the first wall of JET will be up to about 10E20 n/m2, comparable to that occurring in ITER at the end of life in the rear part of the port plug, where several...
An optimization approach that incorporates the predictive transport code TRANSP is proposed for tokamak scenario development. Optimization methods are often employed to develop open-loop control strategies to aid access to high performance tokamak scenarios [Fusion Eng. Des. 123 (2017) 513–517]. In general, the optimization approaches use control-oriented models, i.e. models that are...
The original DTT (Divertor Tokamak Test facility) proposal presented in 2015 [1] in the challenging area of plasma exhaust has been described in detail in [2] with a critical review of several aspects.
Afterwards, according to the conclusions of the DTT Workshop held at Frascati in June 2017, various points have been examined to improve the proposal in the design review phase.
One issue is the...
Using helium as a working gas in COMPASS tokamak is a vital part of the research program. Unfortunately plasma breakdown in helium is quite difficult owing to heliums high ionization potential. In the absence of impurities with lower ionization energies helium plasma breakdown is very sensitive to start-up parameters and thus unreliable. To mitigate this issue several methods have been...
A new high magnetic field tokamak, COMPASS-U [Panek et al., Fusion Eng. Des. 123 (2017) 11 – 16], replacing the currently operating COMPASS tokamak at IPP Prague is being designed and constructed nowadays. As a result of high magnetic field in the machine (B0 = 5 T) large forces (up to 6.5 MN) acting especially on the TF coils are expected. In order to keep the loading of the coils, which will...
An upgrade of the RFX-mod experiment is presently in the final design phase, with the main objectives of improving the control of magnetic confinement, plasma density and plasma wall interaction in both RFP and Tokamak configuration.
The achievement of these aims implies a major change and reconfiguration of the internal components of the machine assembly: the present toroidal support...
In 2017 the Wendelstein 7-X stellarator (W7-X) has performed the second experimental phase (OP 1.2a), introducing divertor operation with a inertially cooled test divertor unit (TDU), and graphite-covered first wall.
The divertor operation resulted in a substantial improvement of the plasma parameters and the inertially cooled TDU and the graphite walls also allowed much longer discharges with...
Neutral Beam Injectors (NBI) for DEMO-like reactors will need deuterium neutrals at a high energy (>0.8MeV) and a fair injector overall efficiency (>50%) for plasma heating and current drive. The neutralization efficiency of positive ions drops for energy higher than 100keV/nucleon and so NBIs based on negative ions are required. A conceptual design of injectors (so called Siphore at the IRFM...
In ITER, each Heating Neutral Beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a negative ion beam up to the energy of 1 MeV for a duration of 1 hour.
To this purpose, a large RF-driven plasma source is required to generate a 40A D- or 46A H- ion current, with low electron/ion ratio (<1) and high uniformity over the extraction area (800 mm x 1600 mm). SPIDER...
In arc discharge plasma, an ion species ratio depends on plasma density and ion confinement time. To increase an atomic ion species ratio in an NBI (neutral beam injector) ion source, various ion sources have been designed to have higher ratio of the plasma volume to the effective ion loss area so that the ion confinement time could be longer. For that reason, most of NBI ion sources have a...
A 3.7GHz lower hybrid current drive (LHCD) system was built on HL-2M in 2017. A Full-Active Multijunction (PAM) launcher is installed. The peak parallel refractive index is 2.25 with a range from 2.1 to 2.4. ALOHA code predicts a low Reflection Coefficient (RC) within a large range of the HL-2M plasma density. TE10 to TE30 mode converters are designed for the antenna to divide the power in...
Electron cyclotron heating and current drive (ECHCD) system with a 28 GHz gyrotron has been prepared for non-inductive electron cyclotron (EC) plasma ramp-up in the QUEST spherical tokamak (ST). Non-inductive plasma start-up using the EC waves is a key issue for advanced tokamak reactor concepts as well as for the ST concept. There are two important aspects of conducting the present ECHCD...
The KSTAR's NBI heating system achieves an output of about 5.46 MW, 8.7s through the 2017 KSTAR Campaign. The second NBI system (NBI2) has being installed to increase the output of the NBI heating system. The NBI2 system consists of one on-axis and two off-axis ion sources, and the maximum output is designed to be 6 MW based on the neutral beam output. The ion source of NBI2 is in the same...
Radio Frequency (RF) waves in the Ion Cyclotron Range of Frequency (ICRF) are successfully used to heat fusion plasmas. For a better understanding of the ICRF wave propagation, absorption and other processes in the plasma good in-vessel diagnostics are essential.
In recent years, a number of high frequency B-field probes have been installed in the ASDEX Upgrade tokamak. These probes, arranged...
A short-pulse and high power neutral beam injection (NBI) system is developed for the Versatile Experiment Spherical Torus (VEST) as a main plasma heating device. The NBI system is designed to inject above 0.5MW neutral beam heating power at the hydrogen ion beam energy of 20 keV in the pulse length of 10 ms. In addition, a design feature of VEST NBI system is changing the beam injection...
The small scale prototype negative ion source has been designedfor KSTAR negative ion source. The target performance of the ion source is to extract 0.5 A of 200 keV D-. The aperture geometries of the accelerator gridsare based on the ITER HNBI reference design and optimized for the small scale prototype ion source.The accelerator consists of plasma grid (PG), extraction grid (EG), and ground...
To achieve the high performance plasma in the Korea Superconducing Tokamak Advanced Research (KSTAR) tokamak, Neutral Beam Injection (NBI) system has been installed. The first NBI (NBI-1) was installed in 2010, which provides a 100 keV deuterium neutral beam of 5.5 MW maximum using three ion sources. The second NBI (NBI-2) with another 6.0 MW will be constructed until 2019. In this process,...
Production of negative ion plays an essential role in Neutral Beam Injection (NBI). Research on a cesium-free negative ion source using sheet plasma has being carried out. The sheet plasma is suitable to produce negative ions because the electron temperature in the central region of the plasma is as high as 10 -15 eV, whereas in the periphery of the plasma, a low temperature of a few eV is...
The ITER Heating Neutral Beam injector will be equipped with a beam source that will provide a negative beam of 40A (H2 or D2). The R&D activities undertaken in Europe to pursue this challenging goal comprise three experiments:
ELISE the half size ion source experiment operating in IPP Garching
SPIDER the full size ion source experiment at Neutral Beam Test Facility (NBTF) site in Padua...
The ITER ECRH system consists of 24 gyrotrons and their associated power supplies providing up to 24 MW millimeter wave heating power at a frequency of 170GHz, a set of transmission lines connecting the gyrotrons with the equatorial and the four upper launchers. With its high frequency this heating system provides the unique capability of driving locally current due to the small beam focus of...
The paper presents the ICR-heating system of the TSP TRINITI complex: purpose; features of implementation; characteristics; power supply system; physical condition; and also discusses the possibility of upgrading the system of ICR-heating of the TSP TRINITI complex for the Ignitor project.
According to the project of tokamak Ignitor to accelerate the plasma ignition and facilitate the access...
Negative ions play an essential role for Neutral Beam Injection (NBI) system of steady state magnetic nuclear fusion. We have development of negative ion sounrce in a cesium-free discharge by the magnetized sheet plasma device, TPD-Sheet IV [1]. Negative ions are formed by volume-production, that is, the dissociative attachment of low energy electrons (Te = 1-2 eV) to highly vibrationally...
A Test Facility (TF) has been designed and installed at SPC to allow for the commissioning of the EU gyrotrons developed in view of their integration to the ITER EC system. The first phase of operation of this TF was dedicated to the development of the EU 2MW coaxial cavity gyrotron [1,2]. The EU gyrotron development for ITER has been reoriented since then and is presently advancing a...
Fenix, the ASDEX Upgrade (AUG) flight simulator under development, is based on the Plasma Control System Simulation Platform (PCSSP) developed for ITER and the ASTRA transport code. Fenix will give a session leader the possibility to check whether the discharge will meet experimental goals prior to execution. It reads the AUG discharge program and checks if all the parameters and reference...
The electron cyclotron heating system in DIII-D comprises four 110-GHz gyrotrons and one 117.5-GHz gyrotron able to inject more than 3 MW into the plasma for an administrative limit of 5 s during the 2018 experimental campaign. In a major controls upgrade, the gyrotron high voltage reference waveform is no longer generated by an obsolete pre-programmed waveform generator but by a newly...
The development of a conceptual design for a demonstration fusion power plant (DEMO) is a key priority of the recent European fusion program. The DEMO design faces an even higher challenge taking into account that, compared to ITER, the European DEMO design has a fusion power that is four times higher and a major radius that is only 1.5 times larger than ITER. From a first review of the wall...
Plasma control design increasingly depends on fast simulations able to connect to operational plasma control system (PCS) software for iterative development. GSevolve is a nonlinear free-boundary simulation that evolves the Grad-Shafranov equilibrium including current and pressure profiles, and can connect to all versions of the DIII-D PCS operational on devices around the world. Its ability...
A careful control of poloidal field (PF) coil currents is indispensable to assure a successful plasma initiation in ITER. This requires the development of an accurate modeling tool which can evaluate PF coil current and voltage waveforms leading to a satisfactory breakdown condition. In particular, it is of most importance to provide the necessary loop voltage with a sufficiently large field...
The Southern Europe Thomson Backscattering Source for Applied Research (STAR) is a compact hard X-ray source designed by INFN for advanced applied materials-science research and founded by Progetto
MaTeRiA. MaTeRiA has been realized thanks to a partnership between the University of Calabria and CNISM (Consorzio Nazionale Interuniversitario per le Scienze fisiche della Materia).
The machine...
The magnetic diagnostic system in RFX-mod [1] includes local field probes, flux loops and saddle loops. It plays an important role in real time plasma control and in several off line studies of the plasma configurations and behaviours. The failure or malfunction of one or more components can cause negative consequences in several applications, including the plasma equilibrium and stability...
The reconstruction of the Plasma Boundary (PB) in fusion reactors is a critical task in several applications, including plasma control and several off-line studies of the plasma configurations.
A widely shared definition for PB is the Last Closed Magnetic Surface (LCMS) within the vacuum vessel. In the limits of plasma equilibrium conditions, the PB reconstruction should imply the solution of...
Integrated control of the toroidal current density profile, or alternatively the q-profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q-profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics...
The JET Real-Time Protection Sequencer (RTPS) co-ordinates responses for magnetic and kinetic actuators to protect the ITER-Like Wall from possible melting events and other undesirable scenarios. It allows programmable stop responses per pulse, based on alarms raised by other systems.
The architecture combines a modular run-time application developed using MARTe (Multithreaded Application for...
The Satellite Tokamak Programme (STP) is the main project within the Broader Approach agreement. The STP includes the construction of the JT-60SA superconductive tokamak and its exploitation as an ITER “satellite” facility. In view of JT-60SA operations, Japanese and European scientists are developing different tools to support preliminary studies. In this context, a set of tools for the...
Mitigation of the intense heat flux to the divertor is a crucial issue for safety operation of ITER and next-step devices. The divertor heat fluxes can be significantly reduced by operating detached plasmas, where a huge amount of energy carried by plasma particles is converted into isotropic radiation. In COMPASS-U the power decay length is expected to be small due to high magnetic field,...
ST40 is a new high field spherical tokamak built by Tokamak Energy Ltd. (TE). When operating at full power, the main parameters of ST40 will be: R0=0.4–0.6m, A=1.7–2.0, Ipl=2MA, Bt=3T, κ=2.5, and it will have up to 2MW of auxiliary heating power, and a pulse length of 1-2s. The first operational phase of ST40, at limited performance, took place in early 2018, and its results are presented by...
The Real Time Protection Sequencer allows the Session Leader to program magnetic and kinetic actuators in response to alarms. We will deal with the State Diagram, actuator conflicts and Jump to Termination, which drove the changes to the software and architecture in [1].
The State Diagram determines what happens when one response type follows another, e.g a main chamber hot spot response...
The runaway electron generation during plasma disruptions presents a danger to the vacuum vessel and associated instrumentation. The presented work concerns application of semiconductor detectors for study and characterization of runaway electrons events. The recent advances in the field of semiconductor detectors allow for the development of new diagnostic methods, utilizing their particle...
C/O monitor system is a specially dedicated spectrometer for Wendelstein 7-X, which is planned to be installed before the second Operational Phase (OP 2). It will be a high throughput and high time resolution system which will be able to monitor the main low-Z impurities in the plasma. It will be fixed at nearly horizontal position for the observation of Lyman-alpha of H-like ions of carbon...
Neutron spectrometer based on diamond detector has proposed for the D-D fusion neutron measurements at the KSTAR tokamak in future. Neutron flux monitoring and neutron spectrometry at the KSTAR tokamak as well as at fusion reactors are one of important diagnostics. Diamond has excellent properties in environments such as radiation harsh, high temperature, small size, and so on.
The aim of...
KSTAR plasma experiments have been carried out without baking the inner wall (called as cold wall condition) until 2016. The drift (ΔVsen/Vsen) in magnetic sensor signal was able to satisfy with the requirement for the plasma control by adjusting the offset level of the input in the integrator for several the vacuum shots. Here, the required value of (ΔVsen/Vsen) should be below 2 % during 60...
In this paper the method of complex geometrical optics (CGO) is applied to the analysis of polarymetric laser beam propagation and diffraction in tokamak plasma. The paraxial CGO reduces the problem of beam diffraction in inhomogeneous medium to the set of ordinary differential equations, which significantly simplifies calculation as compared with full-wave approach. Numerical simulations for...
Shutter systems for various diagnostics at the Wendelstein 7-X stellarator
K. Höchel, R. Laube, C. Brandt, M. Otte, M. Schülke and the W7-X Team
Max-Planck-Institut für Plasmaphysik, Greifswald, Germany
The stellarator experiment Wendelstein 7-X (W7-X) has completed the first two operational phases. During the last operation phase OP1.2a in 2017 a maximum energy limit of 80 MJ within a 24 s...
The High Field Side Reflectometry (HFSR) is one of the ITER diagnostics, which provides information about plasma state by measuring the electron density profile. HFSR diagnostics undergoes a strong action of different physical nature loads.
Five various designs have been developed and studied since 2014 as a result of interaction of Peter the Great St. Petersburg Polytechnic University, NRC...
Since impurities determine the plasma performance in magnetically-confined fusion plasmas, analyzing their behavior and controlling their confinement is significantly important for a stable and steady-state operation of fusion devices. A comparison of the impurity transport in the core region of large stellarators, such as Wendelstein 7-X and Large Helical Device (LHD) is quite useful to gain...
First mirrors (FMs) of optical diagnostics in ITER will operate in a harsh environmental conditions. Deposition of plasma-facing materials on the mirror surface will cause a degradation of the nominal diagnostic performance. During ITER lifetime the mirror optical characteristics are intended to be periodically recovered by an appropriate cleaning technique. Cleaning system based on the RF...
In a tokamak device magnetic diagnostics play a key role in the understanding of plasma physics, for control and safe operation. JET tokamak has hundreds of magnetic sensors distributed over the torus, designed to withstand neutron fluence. By the end of 2016 experimental campaign C36B, JET lost several pick-up coils used both for equilibrium reconstruction (slow coils) and MHD analysis (fast...
Paper describes recent advances in Heavy Ion Beam Probing (HIBP), a unique diagnostics to measure the core plasma potential in the T-10 tokamak (R = 1.5 m, a = 0.3 m, B_tor = 1.5 - 2.5 T). Fine focused (< 1 cm) and intense (<130 mkA) Tl+ beams with energy up to E = 330 keV, equipped by advanced control and data acquisition system provides the measurements in the wide density interval n_e =...
Magnetic measurements at long pulse magnetic confinement fusion devices require implementation of the true steady state magnetic field sensors in order to achieve required precision of plasma position measurement. Inductive sensors can suffer from a range of temperature gradient and radiation induced offsets which together with the intrinsic offsets of analogue integrators can lead to unwanted...
For an ITER optical diagnostic, the components located within the port interspace area must be equipped with the proper remotely controlled positioning/alignment mechanisms meeting the variety of functional, environmental and load requirements. For the H-alpha diagnostics the major requirements are: 1) high structural stiffness and positioning stability under the ~100kg weight of Long Focus...
For an ITER optical diagnostic, the components located within the port interspace area must be equipped with the proper remotely controlled positioning/alignment mechanisms meeting the variety of functional, environmental and load requirements. For the H-alpha diagnostics the major requirements are: 1) high structural stiffness and positioning stability under the ~100kg weight of Long Focus...
The WEST thermal fusion reactor is currently an upgraded evaluation environment for the diagnostics to-be-deployed in the constructed ITER tokamak. In particular, fast metallic impurities diagnostics is tested, that in future can increase reaction efficiency and provide safe mode of operation with the divertor. A diagnostic system has been provided in the framework of the collaboration and the...
A Neutron Activation System (NAS) is a highly useful and reliable tool for neutron flux and spectrum measurements in fusion devices such as ITER or ENS. The underlying process involves neutron irradiation of chosen material probes followed by their gamma spectroscopy. The gamma spectra and the further data analyses produce quantitative information on the neutron field characteristics in the...
Understanding of resonant interaction between particles and Alfven eigenmodes is one of the most important issues for fusion plasma research. Destabilization of modes by energetic ion population and energetic ion redistribution by unstable modes have been studied by many researchers. However, the interaction of bulk ions with modes has almost not been investigated. Because some bulk ions also...
Understanding core MHD activities is of particular interest on ITER. The fast measurement of core density fluctuations might be a viable avenue for the exploration of plasma core activities in some of the ITER operational scenarios. Fluctuation beam emission spectroscopy (BES) would utilize the diagnostic neutral beam shot into the plasma, where beam atoms get to excited states due to...
In magnetically confining plasma experiments, measurement of ion dynamics is of great
importance to study the impurity transport in the scrape-off-layer (SOL) and divertor.
Impurities from the divertor areas may degrade the core energy confinement, if transported too
far upstream in the SOL. The Doppler coherence imaging spectroscopy (CIS) is a relatively
new...
The Collective Thomson Scattering (CTS) will be the ITER diagnostic responsible for measuring the alpha-particle velocity distribution. Using mirrors, a 1 MW microwave beam is directed into the plasma via an opening in the plasma-facing wall. The microwaves will scatter off fluctuations in the plasma, and the scattered signal is recorded after transmission through a series of mirrors and...
Due to the successful program of ASDEX Upgrade (AUG) the high current converters come to their limits. At the same time the risk of failure increases because of ageing. At the moment all converters are in use and thus the area of operations is fixed. Low power, high density discharges of AUG are limited by the OH transformer flux. Therefore a new thyristor converter group, called “Group 7”, is...
Within the framework of the ITER project, CEA/SBT is in charge of the design, manufacturing and delivery of 277 Venturi tube flowmeters for the control of the superconducting magnets. Six types of flowmeters were developed for either the control of the supercritical helium flow in the magnets at cryogenic temperature (4.5 K) or the operation of the current leads at room temperature. The last...
The Experimental Advanced Superconducting Tokamak (EAST) poloidal Field Power Supply has recently implemented the upgrade of the control system. Inspired by the ITER Control Data Access and Communication (CODAC) system, experimental physics and industrial control system (EPICS) has been chosen for the control system. The Data Acquisition (DAQ) system is an important subsystem of the overall...
In the Poloidal Field (PF) Power-supply System of International Thermonuclear Experimental Reactor (ITER), the phase-control thyristor converter is used to supply power to the PF superconducting coil. A precise zero scale is provided for the thyristor trigger by synchronization technology. The method of extracting the synchronization signal on the rectifier transformer primary side is widely...
This paper describes the development and testing of a 15 kA Solid Circuit Breaker (SCB) applied to the Switch Network Unit (SNU) of the Experimental Advanced Superconducting Tokamak (EAST). The circuit scheme composed of parallel connected integrated gate-commutated thyristors (IGCT) and the diode bridge to realize bidirectional breaker ability is presented firstly. In this paper, the stray...
Current start-up experiment and Steady-State Tokamak Operation (SSTO) are performed by Electron Cyclotron Heating (ECH) in QUEST spherical tokamak. SSTO discharge can be maintained longer than 2 hours using continuous 8.2 GHz ECH. For a new ECH for SSTO, 8.56 GHz high power klystron system is in preparation. The klystron can work continuously, and its incident RF power is 250 kW maximum....
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. Fusion for energy (F4E) is responsible for supply five of them(PF2-PF6) as in-kind contributions to ITER project. ITER PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and F4E.
The...
The Quench Protection Circuit (QPC) can provide a fast and reliable protection for superconducting magnet. According to the conceptual design of China Fusion Engineering Test Reactor (CFETR), a high power QPC with mechanical-static hybrid switch is proposed, which is designed for the superconducting Toroidal Field (TF) magnets. The rated currents and maximum reapplied interruption voltages for...
The external bypass, as an important components of the international thermonuclear experimental reactor (ITER) poloidal field converter unit (PFCU), will provide a freewheeling loop for the load current to protect the magnets and PF converter modules from being damaged by over-current and over-voltage under fault conditions. The triggering bypass protection test is used to verify that the...
The SPIDER experiment, currently starting to operate at the Neutral Beam Test Facility (NBTF) in Padova, is the full-size prototype of the radiofrequency (RF) ion source for ITER Neutral Beam Injector (NBI). It features a 1MHz RF system including 4 generators, each one rated for delivering up to 200kW of RF active power. A single generator is composed of 2 tetrodes connected with the push-pull...
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High-temperature superconductor (HTS) CrossConductors (HTS CroCo) are twisted stacked strands built from HTS tapes, optimized for high engineering current density and easy long length production [1,2].
The production for a 35 kA DC cable demonstrator required increased amounts of HTS CroCos for which the production of HTS CroCos was extended to 8 m length. This milestone shows clearly that...
The electron cyclotron resonance heating group installs four new gyrotrons to enlarge the heating power at ASDEX Upgrade (Axially Symmetric Divertor Experiment). Gyrotrons are microwave oscillators for additional plasma heating. Therefor two new direct current power supplies are needed. One power supply is able to feed two gyrotrons.
The engineering data for both power supplies are:
- dc...
The ELM triggering with H-mode will cause energy loss of plasma in Tokamak device. The phenomenon results in the timely response of control system and motivates the rapid variation of current in PF coils which affects the AC loss and stability of superconducting magnet. The cryogenic parameters of the superconducting magnet will be analyzed according to the EAST experimental data for recent...
The Power Supply system for Resistive Wall Modes (RWM) control in the JT-60SA experiment is an Italian in-kind contribution to JT-60SA within the Broader Approach Agreement.
A very efficient control of RWM is necessary to access plasma currents with high βN values (3-5) sustained for a hundred seconds, as foreseen in the JT-60SA research plan.
The development of the whole RWM control system...
Magnet power supply (PS) system of JT-60SA is being implemented within the Broader Approach Agreement aiming to achieve first plasma in 2020. The system consists of several components such as DC power supply, switching network unit, quench protection circuit and motor-generator. All the components have their own internal controller (LCC) for stand-alone operation so the individual...
The plasma breakdown and ramp up in fusion experiments require high peaks of power, increasing with the size of the machines and plasma current value. In ITER, the power peak reaches 600 MW and in DEMO the present estimation gives values higher than 1 GW, far beyond the limits generally fixed by the power grid operators, even for special plants.
Experience from existing fusion experiments,...
The European DEMO, which will follow ITER according to the Roadmap of the European fusion program, is presently under Pre-Conceptual Design. The DEMO Plant Electrical System (PES) will include the Power Supply (PS) and the Electrical Power Generation systems. In all the present tokamaks, the main magnets, which constitute the major loads, are powered by thyristor rectifiers; in DEMO they...
In large superconducting magnet system, the release of inductive energy stored inside the superconducting coils provides a considerable potential of dc-arc hazards in case of an accident such as unmitigated quench. Safety analysis and numerical calculation have raised great concerns. This paper will present an electrical arc simulation method using Matlab/Simulink software to simulate the...
The pre-conceptual design of the European DEMO fusion tokamak is currently being developed under the coordination of the EUROfusion Consortium. This paper reports the mechanical analysis of the central solenoid (CS), which comes right after the phase of definition of the winding pack proposed by the CEA. An analytical model is firstly developed in order to approximate the required axial...
The ITER PF6 coil is composed by 9 double pancake(DP).The vacuum pressure impregnation（VPI）technology is an important process for double pancake manufacturing.The double pancake winding is contained within the VPI mould consisting of vacuum chamber and hard mould.The vacuum chamber provide the vacuum and pressure conditions required by VPI process,The hard mould provide the heating and shape...
The low-temperature superconducting (LTS) joint box is an important part of ITER HTS current leads. Enabling to provide the required functionality, the LTS joint boxes are made out of Copper-316L bi-metallic explosion plates. The bimetal interface of the joint has the direct effect on the mechanical properties of the joint and testing performance at low temperature. For this reason, the...
JT-60SA is a tokamak device using superconducting coils to be built in Japan, as a joint international research and development project involving Japan and Europe. One design object of JT-60SA is maximization of the plasma volume and the instrumentation port availability in the condition that several buildings, heating instruments, and diagnostics of JT-60U are reused. The cryogenic pipe...
Tokamak operations requires high-current coils. The systems typically used to supply these coils (self-commutated converters, resistive switching network units) present several drawbacks: high reactive power, harmonics, discontinuous loads, dead times, energy dissipated as heat during breakdown. Some traditional solutions involve large flywheels and power factor compensators. The achievable...
The thermal hydraulic analysis of the DEMO cryo-magnetic system has the main objective to minimize the refrigeration power. The cryo-magnetic system includes the superconducting magnets cooled by forced flow supercritical helium at about 4.4 K, the cryo-distribution lines and valve boxes, and the cryogenic system with several cold boxes. The DEMO cryogenic system is at the pre-conceptional...
The long plasma pulse duration and the large thermal loads expected in the EU DEMO reactor, currently under design, represent a challenge for the power exhaust and ask for a new, robust design of the divertor. For this reason, the design of a satellite fusion experiment, the Divertor Tokamak Test (DTT) facility, is being pursued in Italy. This fully superconductive compact tokamak, which must...
The construction of the full-superconducting tokamak JT-60 Super Advanced (JT-60SA) is in progress under the JA-EU broader approach agreement. During cool down, nominal operation and warm-up the thermal shields, the superconducting magnets and their structures, the high temperature superconductor current leads (HTS CL) and the divertor cryo pumps have to be supplied with helium at specific...
The Reduced Activation Martensitic/Ferritic steel (RAFM steel) is known as one of the most important materials for future fusion reactor blanket and the welding process is unavoidable for blanket structure, so it is extremely necessary to research his welding performance. Some simulations and experiments about anti-fatigue of RAFM steel welding specimens have been done in this paper. The...
For a recent Japanese (JA) DEMO reactor design (Rp: 8 m size), the exhausted power to the SOL (Psep) is expected to be 200-300 MW, where large power handling (Psep/Rp = 24-35 MW/m) is required in the SOL and divertor. The conventional divertor design with the divertor leg length of 1.7 m was proposed, where the peak heat load at the divertor target was simulated to be less than 10 MW/m2 with...
Due to the higher melting point and lower sputtering yield, tungsten (W) is considered as the candidate for plasma facing materials (PFMs) in the future fusion reactors. Under working condition, W will be exposed to 14 MeV neutrons produced by D-T fusion reaction, as well as energetic particles such as hydrogen isotope, helium ion. The damages introduced by charge-exchanged particles are...
Tungsten (W) is a refractory metal foreseen as plasma-facing material in future magnetic confinement fusion devices. Furthermore, W containing metallic composite materials, such as W particle- or fibre-reinforced composites, are currently regarded as promising advanced materials to enhance the performance and integrity of highly heat loaded plasma-facing components (PFCs). In principle, W is...
The use of a liquid metal in the plasma-facing divertor components of the DEMO project has attractive properties such as self-healing and neutron resilience and a liquid metal handling system is considered as a candidate technology. The linear plasma device Magnum-PSI is capable of producing DEMO-relevant plasma fluxes which well replicate expected divertor conditions. The present research is...
In this paper, effects of initial condition on seismic analysis for the HCCR TBM-set are evaluated using finite element analysis. Because of difficulty to predict when an earthquake occurs during operation, various scenarios are considered in the structural integrity assessment in ITER. To perform the simplified analysis, it is important to understand the effects of initial conditions on the...
Understanding of hydrogen isotope behaviors in plasma-facing wall is important from viewpoints of fuel control and tritium safety. Tungsten (W) is a candidate material of plasma-facing components. Although a sputtering rate of W is low, a certain amount of W deposition layer would be formed on the plasma-facing wall during a long time operation of a fusion reactor. However, hydrogen isotope...
The high heat flux test facility HELCZA move forward in commissioning phase to the high heat flux qualification tests. The purpose of the qualification tests is the demonstration of facility’s readiness for the testing of plasma facing components primarily first wall panels at high heat loads. For the qualification tests the flat cooper FW mock-up with representative size for electron beam...
A massive tungsten divertor, Div-III, was installed into ASDEX Upgrade (AUG) in 2014. Div-III is an adiabatically loaded component and consists of massive tungsten tiles clamped into their supporting structures. Before installing the new component, extensive studies, including Finite Element Modeling and high heat flux tests in the test facility GLADIS, were carried out. After the first...
Tungsten (W) is considered as a primary candidate material for plasma facing components, which endures high fluence plasma in future fusion reactor. Extremely insoluble helium(He) introduced into W exposed to high fluence He plasma tend to agglomerate into bubbles, which play a significant role in the radiation-induced micro-structural evolution and properties degradation. Therefore, to...
The GDOES (Glow Discharge Optical Emission Spectrometry) technique has been already used for investigation of erosion/re-deposition processes of W coated CFC tiles from JET. The challenge for GDOES was to measure the depth profile of deuterium together with the other elements from the W coatings exposed to JET plasmas. A tentative investigation using GDOES for deuterium analysis was performed...
Plasma facing components (PFC) within a fusion device are subjected to a harsh operating environment such as high heat fluxes and exposure to high flux of hydrogen isotopes. This exposure can lead to a high fuel retention that can raise serious concern from safety point of view. One of the reason for the use of W as a material for construction of the first wall is aimed to reduce the fuel...
At present, the study of structural and functional materials’ properties of fusion reactors is carried out as a part of implementation of ITER and DEMO projects.
The research of liquid metals application possibility as a plasma facing material (PFM) is one of the most important area. Studies carried out on this subject have shown that lithium is a good material for use as a PFM in a fusion...
In future fusion reactor, divertor is a key component where large heat flux from the confined plasma need to be exhausted. In CFETR Integration Design Platform, where the unified environment for both physics and engineering design is provided, a workflow is developed for the design of the geometry of divertor targets. According to the 0-D design and equilibrium of the designed plasma shape,...
In 1998 Makhankov [1] described the concept of modular exchangeable plasma facing components (PFCs) based on liquid metal heat pipes which are radiatively cooled. Here we present results from recent experiments with a lithium filled tubular heat pipe owned by Sandia National Laboratories. The tantalum envelope (~20mm diameter by ~200mm long) was heated on its side wall using a hydrogen plasma...
The European roadmap to fusion electricity attributes critical importance to the development of reliable plasma-facing component (PFC) technology. In the EUROfusion divertor project, a range of PFC concepts are explored, and two “Phases” of small-scale mock-up design, manufacture and high heat flux testing are underway. The rationale is to understand shortcomings revealed by the first phase in...
Erosion, deposition, and fuel retention on different plasma-facing components (PFC) are critical issues for next-step fusion devices such as ITER and DEMO. Since 2011, JET has been operated with the ITER-like wall (ILW): tungsten (W) in the divertor and beryllium (Be) in the main chamber. So far there have been three experimental campaigns in 2011-2012 (ILW-1), 2013-2014 (ILW-2) and 2015-2016...
The development of the future fusion reactor is highly associated to the improvement of the current materials, specially tungsten materials, to withstand the extreme conditions giving inside the reactor vessel during service life. Tungsten has extraordinary physical characteristics as plasma facing materials (high thermal conductivity, sputtering resistance and melting point). However, it has...
Several European First Wall (FW)/blanket concepts for DEMO are high-pressure (8 MPa) Helium-cooled systems. Typical radiative steady state loads delivered from the fusion plasma are predicted up to 0.45MW/m², but peak values could reach and excess 1MW/m². At such high heat fluxes, it is a major engineering challenge to dimension the cooling for moderate temperatures and moderate stresses,...
The first wall panels of the ITER main chamber will be completely armored with beryllium. The primary reasons for the selection of beryllium as an armor material for the ITER first wall are its low Z, high oxygen gettering characteristics and also high thermal conductivity. During plasma operation in the ITER, beryllium besides low cyclic heat loads (normal events) will be suffered by high...
Fusion fuel retention is a major issue for fusion devices. A set of divertor HFGC tile samples from first 2011-12 JET ITER-like wall (ILW-1) campaign were analysed in the research to complement the data obtained by other analysis methods [1].
Experiments were carried out using Hositrad MGT 6-300 Multi Gas Analyser with Thermal Desorption. The device operates with two Quadrupole Mass...
For the use in a fusion reactor, tungsten has unique properties such as a high melting point, low sputter yield and hydrogen retention as well as moderate activation. The brittleness below the ductile-to-brittle transition temperature and the embrittlement during operation are the main drawbacks for the use of pure tungsten in plasma facing components. Tungsten fibre-reinforced tungsten...
China Fusion Engineering Test Reactor (CFETR) is a tokamak reactor under design. Due to the maintenance, assembly of the blanket and heating, diagnostics of the plasma, ports must be opened on the vacuum vessel and the blanket module of CFETR. However, the ports affects the radiation shielding performance of CFETR, especially the equatorial ports where the neutron wall loading appears a...
The world’s largest modular stellarator, Wendelstein 7-X (W7-X), is in operation since 2015. Stepwise increase of operation parameters has its final goal in the demonstration of steady-state operation capabilities with pulses up to 30 minutes. Such pulses require a constant heating of the plasma due to losses by plasma-wall interactions, particle drift and radiation losses. The latter are...
Remote maintenance (RM) of highly activated components in the test cell (TC) is a key issue of design of accelerator-driven (IFMIF-type) fusion neutron sources. We newly propose a sophisticated RM method for the target assembly (TA) of an IFMIF-type Advanced Fusion Neutron Source (A-FNS), aiming at improving maintainability of the in-TC components in this study. Basic ideas of the proposed RM...
Remote Handling (RH) maintenance in hazardous environments is a challenging task to be performed on systems and components to ensure that they work as per design and that the requirements on the plant availability are fulfilled. According to this, systems and components have to be designed for easy assembly. This approach leads to an improvement and efficient maintenance process, reducing...
The China Fusion Engineering Test Reactor (CFETR) is the testing fusion device which would be the prototype for future commercial reactor. However, the traditional maintenance way is mainly remote handling which is time-wasting and low efficiency. To meet the demands for more complicated maintenance, it is no hesitation to start the more intelligent devices for the fusion device. The dual arm...
The ITER Cryostat, the largest SS vacuum pressure chamber ever built which provides the vacuum confinement to components operating in ITER ranging from 4.5 k to 80 k. Cryostat Design Model was qualified by ITER. As a Safety Important Class system, Design qualification at every change in its development and installation phases is mandatory. The Cryostat system is currently at manufacturing...
The ITER cryostat—the largest stainless steel vacuum pressure chamber ever built which provides the vacuum environment for components operating in the range from 4.5 k to 80 k like ITER vacuum vessel and the superconducting magnets. The Cryostat is divided into four section, of which, Base section is most complex because of its web shaped structure sandwiched between two 60mm thick plates with...
It is foreseen the most challenging tasks currently in remote maintenance of DEMO are the remote handlings of multi module blanket segment and divertor. Both of them are large and heavy, and must manoeuvre through precise trajectories in a limited space by remote manipulators. The manipulator deformation, due to the heavy payload, will deviate the handled component from the desired...
The full exchange of the ITER divertor is performed with the Divertor Remote Handling System during ITER long-term maintenance campaigns. Access to the vessel is possible through the lower maintenance port. The size of the port is highly constrained, therefore, high power density actuation systems are required to lift and transport the 10-tonne cassette assemblies in and out the vessel. It has...
Remote Handling (RH) equipment are deployed to exchange the ITER Divertor, segmented in 54 cassettes. The RH equipment are powered and controlled with water hydraulics, using self-supplied demineralised water as a pressure medium. Water hydraulics servo control technology has been successfully proven at Divertor Test Platform 2 (DTP2) with a full-scale prototype, namely Cassette...
The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) is in progress in the Power Plant Physics and Technology (PPPT) programme within the EUROfusion Consortium. In this work neutronics studies, fundamental for the nuclear design of DEMO, are presented for the horizontal lower port, an optional concept that is currently under investigation. Two possible configurations...
The fusion reactor produces a large amount of tritium dust in the vacuum vessel due to the plasma unstable events, decontamination is an indispensability dispose in the IVC maintenance and decommissioning, it also plays an extremely important role in reducing the radioactive pollutants diffusion, cumulative radiation dose and staff occupational exposure level, controlling radioactive effluent...
COMPASS Upgrade tokamak is a medium-size high-magnetic-field device currently in the conceptual design phase [Panek et al. Fus. Eng. Des. 123 (2017) 11-16]. Due to the high plasma current (up to 2 MA) and the strong magnetic field (up to 5 T), large electromagnetic forces on conducting structures surrounding plasma are expected during disruptions. To address this issue, electromagnetic loads...
One of the major challenge in the commercialisation of magnetic confinement fusion is maintaining the powerplant reactor in a sufficiently short period of time to achieve commercial levels of plant availability. To inform the development of an appropriate remote maintenance strategy for EU DEMO, a simplified, parametric model, called the Maintenance Duration Estimator (MDE), has been created...
The numerical analysis of robotic mechanisms for remote maintenance and inspection inside nuclear fusion reactors has to face several issues. Indeed, these robots are subject to large deformations, which are either induced by their own mechanical structure or by the heavy payloads which they usually handle. In many applications, robotic systems are usually modeled with rigid elements, although...
In this work the authors present the latest progresses in the conceptual design of the first wall and the main containment structures of DTT device. The previous DTT baseline design was reviewed in terms of both materials and plasma shape. This in turn led to a new all-welded double-wall vacuum vessel structure, made of AISI 316L(N) stainless steel. While the basic design has still 18 sectors,...
The banana-shaped segment transport using all vertical maintenance ports (BSAV scheme) was selected as the primary maintenance option on Japanese DEMO. Among various engineering issues on the BSAV scheme, recent progress on remote maintenance (RM) focuses on in-vessel transferring mechanism of the segment, support structure of the segment and pipe connection. In the BSAV scheme, cooperative...
In Early 2018 a new challenging project of building unique Tokamak device Compass-U has begun. Apart from others, the technical attributes will include elevated operating temperatures of vacuum vessel and plasma facing components, active cooling of temperature sensitive diagnostics, cooling of the magnetic coils to LN temperatures and etc. The listed requirements of operating parameters pose...
The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at inspecting plasma facing surfaces of in-vessel components for both damage and erosion, both of which are related to the amount of mobilised dust present in the Vacuum Vessel.
Key design improvements from the on-going IVVS preliminary design have recently been incorporated into a...
One of the principal means of controlling welding distortion of metallic welded components is the use of assembly jigs. The knowledge about the effect of jigs on weld distortion is based in the qualitative, generally neither systematic nor documented experience of each manufacturer. Such knowledge needs to be complemented by specific research for components with the tight tolerances, intricate...
During welding of large, massive and complex assemblies, such as the ITER Vacuum Vessel PS3, the accumulated shrinkage of parallel welds may cause a cumulative distortion effect which significantly varies the dimensions and geometry of the welded portion while still unfinished. In case of the PS3, as consequence of such distortion, the space left to fit the outer shell plates into to the...
EAST Tokamak is a complex fusion device which requires high-quality first wall condition for the long pulse and H-mode plasma operation. During plasma phases, some of the first wall components are damaged due to high thermal stress and electromagnetic force and control is necessary in case of doubt about their condition. Detection and locating the damaged PFCs are the precondition to determine...
Tritium breeders is required to have a high lithium atom density from the viewpoint of tritium breeding ratio. The first candidate material being studied in Japan at the present time is Li2+xTiO3 which is Li2TiO3 containing excess Lithium [1], however development of a material having even higher Lithium atom density is still under way. Li8ZrO6 has the highest lithium atom density except for...
The development of new manufacturing methods for the production of key components for nuclear fusion reactors by selective laser manufacturing (SLM) is currently under investigation at Karlsruhe Institute of Technology. SLM offers great potential compared with conventional manufacturing methods, especially for fabrication of thin- and double-walled structures like sandwich-type flow channel...
Several manufacturing routines were developed in KIT for First Wall components of the Helium Cooled Pebble Bed concept: for the Test Blanket Module of ITER in 2010-2015 and the Breeder Blanket for DEMO since 2014. The overall fabrication strategy consists of two main steps: 1) the manufacturing of a semi-finished plate penetrated by channels, and 2) the forming of the plate with channels into...
Reduced Activation Ferritic Martensitic (RAFM) steel is currently under intense consideration as a structural material for the blanket applications in fusion reactor. The concept of blanket module utilizes both solid i.e. Li2TiO3 and liquid breeder material, i.e. Eutectic Pb-17Li operating at 320-480C. The critical issues like liquid metal corrosion of RAFM steel, tritium permeation into RAFM...
In a detritiation system, the function of recombiner is oxidation of tritium components. Detritiation system of a nuclear fusion facility should maintain its detritiation performance even in an event of accidents of the facility such as fire. The major technical background on recombiner design are detailed kinetics data, impact of gaseous impurities on kinetics, characteristics on pressure...
In dual-coolant lead lithium blankets, foreseen in fusion power plants, the liquid metal PbLi flows at sufficiently large velocity to guarantee a suitable removal of the volumetric heat generated in the fluid. The moving electrically conducting fluid under the influence of the plasma-confining magnetic field induces currents that create strong electromagnetic Lorentz forces and a high...
Water cooled ceramic breeder (WCCB) blanket has been considered as the primary design in Japan. There are many thin cooling pipes in the blanket and internal pressure would be applied to the blanket by the pipe rupture. A number of efforts have been made for keeping pressure resistance of the blanket with box structure. Pressure resistance of the box structure could be reinforced with internal...
This work concerns the conceptual design of the pipework and the main equipment of the Primary Heat Transfer System (PHTS) of EU-DEMO fusion power plant. EU-DEMO is considered to be the nearest-term reactor design to follow ITER; it shall be capable of demonstrating production of electricity, operation with a closed fuel-cycle and to be a facilitating machine between ITER and a commercial...
To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test-Blanket-Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. Each TBM and its associated shield form a TBM-Set that is mechanically attached to a TBM Frame. A TBM Frame and two TBM-Sets form a TBM port plug (TBM-PP). Actually different TBM versions will be...
In liquid metal blanket concepts for nuclear fusion reactors which are currently under development, the liquid metal PbLi serves as neutron multiplier, breeder material and shield against high neutron radiation. In helium-cooled or water-cooled blanket designs, the liquid metal may flow only at very small velocity since the entire heat released in the liquid metal is removed by water or...
The Water-Cooled Lithium-Lead Breeding Blanket is a candidate option for the European DEMO nuclear fusion reactor. The blanket is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the heat generated in the tokamak plasma. The last function is fulfilled by the First Wall and Breeding Zone independent cooling systems.
Several layouts of the...
The UK Government has invested ~€100M to create two new UKAEA centres for fusion research – Hydrogen-3 Advanced Technology (H3AT) and the Fusion Technology Facilities (FTF) both opening in 2020-21. FTF and H3AT will foster close cooperation with industry, academia and other international laboratories to develop and transfer knowledge between partners, offering opportunities to undertake R&D to...
A new facility, CLIPPER, is being constructed at CIEMAT to investigate tritium extraction from PbLi. It consists in a forced circulation loop with the main objective of validating the technique of permeation against vacuum. Originally, CLIPPER was designed with two zones operating at different temperatures and connected through a recuperator, which gives most of the required thermal jump. In...
In thermonuclear fusion reactor, tritium generated by nuclear reaction of lithium isotope with mass number six (6Li) and neutron is used as fuel. To maintain the nuclear reaction in the reactor, it is necessary to concentrate the 6Li isotope, which exists at only about 7.8mol% naturally, to 40–90wt%. The mercury amalgam method is the only practical method, but its environmental burden is large...
Plasma enhancement gases (PEGs) (such as: nitrogen, neon, argon and other inert gases) are injected into the plasma of several tokamaks in order to reduce the power load over the plasma facing component.
The exhaust gas in DEMO reactor consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He and impurities.
In DEMO reactor it is foreseen to recover the...
Cryogenic distillation (CD) of hydrogen in combination with Liquid Phase Catalytic Exchange (LPCE) or Combined Electrolytic Catalytic Exchange (CECE) process is used for tritium removal/recovery from tritiated water both for ITER/DEMO and for fission rectors like CANDU.
Tritiated water is being obtained after long time operation of CANDU reactors, or in case of ITER mainly by the...
The He-cooled Lithium Lead breeder blanket could be developed with the utilization of relatively mature material technology, which is used by Reduced Activation Ferritic / Martensitic (RAFM) steel as the structural material in China. It is necessary to analyse the first wall structure heated because of the thermal stress problems from two coolants in the blanket. The deviation from the thermal...
In future fusion DEMO reactors, coolant and breeding materials will flow along loops experiencing very different neutron irradiation environments, from those corresponding to in-vessel systems like blankets and divertor, up to regions without significant neutron radiation farther from the reactor. Moreover, in these loops there are material extractions due to the functioning of the tritium...
The research focuses on testing carbon molecular sieve membrane (CMS) for separation of the exhaust gas from fusion reactors.
The exhaust gas in tokamak demonstration reactor (DEMO) consists of more than 90% of unburned fuel gas (D and T) and the remaining part will be He, plasma enhancement gases (PEGs) and impurities. Plasma enhancement gases (PEGs) (such as: nitrogen, neon, argon and...
Lithium metatitanate (Li2TiO3) is one of the leading candidates for application as a breeder blanket material for the helium cooled pebble bed (HCPB) concept.
During operation, transmutation of lithium contained within the blanket material as a result of neutron capture produces both tritium and helium. For efficient tritium production, high ceramic density is desirable in order to increase...
Water Detritiation Systems (WDS) are used for CANDU (to remove tritium from heavy water) and also ITER. A water detritiaton facility was developed at ICSI Rm. Valcea based on Liquid Phase Catalytic Exchange (LPCE) combined with Cryogenic Distillation (CD). Initial application was for developing Cernavoda Tritium Removal Facility, but later these combined technologies were considered for fusion...
Gas adsorption processes are widely used in industrial applications including vacuum pumping in fusion reactors (getters and cryogenic pumps). In gas adsorption flows mass transfer occurs coupled with heat transfer. Modeling of such flows must be based on kinetic theory by applying the Boltzmann equation (BE) or kinetic model equations or alternatively the Direct Simulation Monte Carlo Method...
One of the main challenges for DEMO is to overcome the Power Conversion System (PCS) issues for a pulsed reactor. PCS is a complex system where the secondary circuits, connected to the Primary Heat Transport Systems (PHTS), should be integrated into an industrial and reliable system based, as much as possible, on proven technology. PCS should be designed to remove heat from PHTSs, during pulse...
Within the framework of the EU-DEMO Breeding Blanket (BB) research activities, the Water-Cooled Lithium Lead (WCLL) BB concept is the only one which adopts pressurized sub-cooled water as coolant and a Heavy Liquid Metal (HLM), namely Pb-15.7Li alloy, as breeder and neutron multiplier.
Cooling water, characterized by operative conditions typical of PWR fission reactors (temperature in the...
At the moment, the lead-lithium eutectic was chosen as the material of the reactor blankets of ITER and DEMO. During the reactor operation radioactive tritium will be produced in the blanket, but the interaction parameters of hydrogen isotopes with lead-lithium eutectic are poorly studied. The most significant processes affecting the release of tritium from the lead-lithium eutectic is its...
In the pathway towards the achievement of a demonstration nuclear fusion reactor DEMO, the construction of a neutron irradiation plant is a priority. Within the European framework, the construction of a facility that is capable of producing a significant amount of irradiation damage in materials as soon as possible was decided. The design for this Early Neutron Source (ENS) is DONES...
The Laboratory for Laser Energetics (LLE) at the University of Rochester supports a 35 kJ, direct-drive, laser to compress DT ice to study inertial confinement fusion physics. One millimeter diameter, hollow, 6 to 10 micron-thick wall, plastic spheres are charged with deuterium-tritium (DT) gas mixtures up to several hundred atmospheres. These targets are cooled to cryogenic temperatures in...
China Fusion Engineering Test Reactor (CFETR) is proposed to be the next generation fusion device of China. The conceptual design of CFETR has been completed and the engineering design is about to start. This work gives an overview of neutronics design analyses of helium cooled ceramic breeder blanket at institute of plasma physics Chinese academy of sciences (ASIPP), for CFETR phase 1 with...
The realisation of the breeding blanket system represents one of the crucial points in the design of future generation fusion reactors. At the time being, problems regarding materials compatibility persist. To protect structural materials from the harsh environment of breeding blankets, different materials have been studied as corrosion-resistant, anti-permeation coatings. Among them, Yttrium...
A conceptual design of breeding blanket module with pressure tightness (as 17.2 MPa) against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. Since the pressure loss of the cooling water system was 1.2MPa, the design pressure of the coolant is set to be...
The liquid lead-lithium (PbLi) blanket concept has become a promising design for fusion DEMO and power plant reactors. To promote the successful application of fusion energy, some R&D activities on the PbLi blanket have been performed, such as structure material corrosion, thermal hydraulics, magnetic-hydrodynamic (MHD) effect, coolant impurities technology and LOCA/LOFA, etc.. Therefore, it...
Beryllium pebbles produced by the rotating electrode method were selected as a reference neutron multiplier material for the DEMO fusion reactor blanket. Recently they were characterized for this application after neutron irradiation at relevant temperatures. Under neutron irradiation at elevated temperatures, beryllium suffers from significant volumetric swelling stimulated by transmutation...
Conceptual design activity for Advanced Fusion Neutron Source (A-FNS) is being carried out for neutron irradiation test of fusion DEMO reactor materials. We are to apply multi activation foils for A-FNS neutron monitor system in order to measure the neutron spectrum using an activation method with an unfolding code. It is important to evaluate the dosimetry cross section data above 20 MeV...
Plasma Laboratory for Fusion Energy and Applications at Instituto Tecnológico de Costa Rica (ITCR) plans the design of SCR-2; a Quasi-Toroidally Symmetric (QAS) two-field period modular Stellarator, aspect ratio ~5 formed by 24 modular magnetic coils. SCR-2 coils design is based on ESTELL QAS configuration (project cancelled) [1], supplied by Max Planck Institute for Plasma Physics,...
Liquid lithium can cause serious corrosion on the surface of metal structural materials that used in the blanket and first wall of fusion device. Fast and accurate compositional depth profile measurement for the boundary layer of the corroded specimen will reveal the clues for the understanding and evaluation of the liquid lithium corrosion process as well as the involved corrosion mechanism....
The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. Fusion for energy (F4E) is responsible for supply five of them(PF2-PF6) as in-kind contributions to ITER project. ITER PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and F4E.
An...
A plasma-facing components (PFC) such as a first wall or a divertor are exposed to high heat flux due to high thermal radiation and high energy particles emitted from a core plasma. Various heat removal techniques had been proposed and developed to remove the high heat load of the PFC. We newly developed a high efficient cooling technique by inducing a turbulent flow in a cooling channel with...
Tungsten is a refractory metal with very good thermal properties i.e. high thermal conductivity, high melting point, good high-temperature fatigue strength etc. Therefore, tungsten is of great interest in thermonuclear fusion research, mainly acting as a heat sink and an erosion protection on the first wall panels. However, mechanical properties and oxidation resistance of tungsten are rather...
Diffusion bonding methods as a candidate solution for Plasma Facing Components in fusion reactors involved significant investigations over the last decades. For diffusion bonding methods the Gleeble 3800 thermomechanical simulator provides a different method for the diffusion welding process: instead of having a furnace with radiation heating and axial forces in a vacuum chamber – the heating...
Tungsten (W) is the leading solid material for fusion plasma facing component (PFC) applications because it has many desirable properties. Future fusion power systems PFCs must tolerate an extremely hostile environment that includes severe heat loads, neutron damage, and surface modifications driven by energetic particle impingement. However, W and most W-alloys exhibit low fracture toughness...
The accelerator-driven fast neutron sources of broad- and quasi-monoenergetic spectra are operated at the NPI Rez Fast Neutron Facility utilizing the Be(thick) and 7Li(C) target stations and the variable energy proton beam (up to 37 MeV) from the isochronous cyclotron U-120M. The beryllium target station standardly operated with 35 MeV proton beam is used for production of IFMIF-like...
The CoCrFeNiMn high entropy alloy (HEA) and its low-activation variants strengthened by dispersion of nano-oxides prepared via mechanical alloying were investigated. The two ways of oxide dispersion creation were verified: direct adding of oxides and internal oxidation of oxidizable elements by adding the gaseous oxygen to the alloyed powder. The grain refinement of the one phase FCC alloy by...
Control of tritium permeation through structural materials is important in terms of fuel economy and radiological safety in fusion reactors; therefore, ceramic coatings have been researched as tritium permeation barrier for decades. Microstructural changes of the coatings, such as crystallization and deterioration, influence hydrogen isotope permeation behavior under reactor operation. In...
Fatigue behaviour of the oxide dispersion strengthened CoCrFeNiMn high entropy alloy was characterised at the first time. The initial powder was prepared via mechanical alloying and the material densification was done by the SPS technique. A microstructure, as well as basic mechanical characteristics, were obtained from the tensile test. These data allowed to set parameters for fatigue...
W-V alloys have already been considered for their use in the first wall armor components, but no information about their oxidation resistance is available. To guaranty passive safety of the future fusion reactors during a loss of coolant accident and the ingress of air in the fusion vessel, a full characterization of the oxidation behavior of new tungsten based materials is required. After a...
In the framework of future generation nuclear reactors, structural materials will face environmental conditions even more challenging with the highest radiation damage levels. To deal with this, oxide nanoceramics have been proposed. Oxide nanoceramics combine the enhanced radiation tolerance of nanocrystalline materials with the chemical inertness of oxides. In this work, the properties of...
Innovation in materials technology goes in parallel with advancements in material characterization techniques. Recent years have seen a large increase in use of transmission Kikuchi diffraction (tKD) to solving complex materials science problems, including nuclear materials and irradiation damage. A lack of high statistics in transmission electron microscopy (TEM) characterization of...
The development of plasma-facing components with a high laser-induced damage threshold (LIDT) is an important part of R&D program for laser-aided diagnostics in ITER. A number of papers have been published studying LIDT of ITER materials. Most of them, however, are the investigations of integrity of first wall materials using laser radiation for simulation of pulsed plasma impact during...
A purification system for lithium loop is one of critical issues to design the accelerator-driven (IFMIF-type) fusion neutron source. It is written in IFMIF Intermediate Engineering Design Report that the levels of nonmetallic impurities such as nitrogen, oxygen, carbon, hydrogen isotopes and tritium should be less than 10, 10, 10, 10 and 1 wt.ppm, respectively, although there is little basis...
Together with the outstanding high-temperature mechanical and physical properties, the highest melting point of all the metals and thermal stability against recrystallization make tungsten (W) one of the main armour and heat sink materials candidate. Nevertheless, its applicability as a high-performance structural material is somewhat limited due to its typically brittle character at low...
IFMIF-DONES (International Fusion Material Irradiation Facility- DEMO Oriented NEutron Source) is an IFMIF-based neutron irradiation facility which aims at providing the irradiation data required for the construction of a DEMO fusion power plant. Comparing with IFMIF, DONES consists of only one deuteron accelerators (40 MeV and 125 mA), and utilizes only the High Flux Test Module (HFTM) for...
In order to facilitate the modeling of thermal hydraulic transients and accident scenarios in fusion reactors, a number of additional fluids were added alongside water in the original ATHENA code, which was based on RELAP5. The same libraries were later ported to both RELAP5-3D and MELCOR 1.8.6 for Fusion, both in use today. The libraries included a number of liquid metals of potential...
In Large Helical Device (LHD) at National Institute for Fusion Science, deuterium plasma experiment with d (d, n) 3He reaction was performed from March to July 2017. The neutron generated in plasma is a direct evidence of this reaction. In addition, the neutron spectrum measurement will be useful in fusion engineering to be able to estimate the activation quantity of the fusion reactor...
The Helium Cooled Pebble Bed (HCPB) blanket concept is one of four EU DEMO (Demonstration Power Plant) blanket concepts currently under development. As part of the general strategy for the qualification of the design in view of licensing and operation, several experiments have been foreseen in to be carried out in the HELOKA facility at KIT.
This work presents the experimental results of a...
Knowledge retention and transfer strategies are crucial for the success of every project. Especially on complex innovation projects, within high-tech sectors such as the aerospace industry or fusion, inter-organizational knowledge management needs to be promoted during the entire project life cycle. Project-created knowledge is based on the learning from the day-to-day activities, planning...
Within the frame of European Fusion Technology Program, two neutronics mock-up (HCLL and HCPB) experiments have been performed using 14-MeV neutron generator (NG) to validate neutronics computational tools and nuclear data library. SuperMC, which is a nuclear design and safety evaluation software system developed by the FDS Team, was verified with HCLL mock-up experiment based on the latest...
Accurate calculation of the dose rate level around nuclear facilities after shutdown can provide important reference for the operation and maintenance of nuclear device, and also has important significance for the design of radiation shielding system and the disposal of nuclide waste. In this paper, based on the advanced neutron/photon transport calculation and activation calculation of...
This paper presents application and optimization of the method for the quantification of beryllium dust particles in experimental complex HELCZA (High Energy Load CZech Assembly) by selected available analytical instruments mainly by portable fluorometry as a procedure ensuing from the NIOSH (National Institute for Occupational Safety and Health) method. The experimental complex HELCZA, a high...
Super Multi-functional Calculation Program for Nuclear Design and Safety Evaluation, SuperMC, developed by FDS Team in China, is a large-scale integrated software system. Taking neutron transport calculation as the core, SuperMC supports the whole process neutronics calculation containing depletion, radiation source term/dose/biohazard, material activation and transmutation. Besides, SuperMC...
Radiation transport models for fusion neutronics analysis are becoming increasingly complex, further exacerbating problems in the creation and integration of neutronics models found in traditional analysis methods using MCNP. Serpent 2, an alternative radiation transport code developed at VTT Technical Research Centre of Finland, is considered as a potential method for neutronics analysis in...
The neutron-induced activation of materials is an important issue for fusion facilities. Photons emitted by the activated material result in a photon field, whose spatial distribution must be taken into account when planning maintenance during shutdown and for decommissioning. One of the approaches to calculate the shut-down dose rate (SDDR) is the rigorous 2-step method (R2S) which is based...
The occupational radiation exposure (ORE) assessment is one of the key aspects of the licensing process for International Tokamak Experimental Reactor (ITER) currently under construction in Cadarache (France). As this machine is the first of its kind, the maintenance activities for the replacement and repair of components are foreseen to be frequent and complex. In this context the remote...
Rigorous-Two-Steps (R2S) is one of the most important methodologies to estimate Shutdown Dose Rate (SDR) in relevant fusion facilities. This method is based on three calculations: first, a neutron transport calculation is performed to estimate the neutron flux in the facility during the irradiation phase; then, this neutron flux is used as input data for activation calculations, primary to...
The EU DEMO reactor, under pre-conceptual design within the EUROfusion Consortium, should produce several MW electrical power from nuclear fusion by the 2050s. DEMO shall be equipped with a Primary Heat Transfer System (PHTS) to remove the thermal power deposited in the plasma-facing components and convert it into electricity, and the associated safety-related components and subsystems require...
The identification of structures, systems and components (SSCs) performing safety functions is of a paramount importance for an EU DEMO development consistent with their implications in each phase of the project: design, fabrication, commissioning, operation, maintenance, inspections and tests. Moreover, this activity has to be performed at the early design stage for the correct definition of...
The DEMO preliminary safety and operating design requirements are being defined aiming at obtaining the license with a relatively large operational domain.
The DEMO design approach is being organized, by taking into account the Nuclear Power Plant ITER and Generation IV lesson learnt. Outstanding challenges remain in areas exhibiting large gaps beyond ITER. Those require a pragmatic approach,...
ICSI has completed in 2015 the conceptual design of the Cernavoda Tritium Removal Facility (CTRF). CTRF is located at CNE Cernavoda, a NPP subsidiary of SNN Bucharest, and is sized to process heavy water from 2 CANDU reactors, treating 40 kg/h heavy water over 40 years with a detritiation factor of 100. CTRF removes tritium using liquid phase catalytic exchange (LPCE) paired with cryogenic...
