Plasma Disruption Events Research Articles

Electromagnetic force evaluation on the water cooled ceramic breeder blanket for CFETR

The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). To investigate the electromagnetic (EM) force (magnetization force and Lorentz force) on the WCCB blanket, EM analyses were carried out using ANSYS Multiphysics code. At first, finite element (FE) static magnetic analysis under normal operation condition was accomplished, adopting the magnetic scalar potential (MSP) method. The magnetization force can be obtained as a result. Then transient magnetic analysis under plasma disruption event was performed using the magnetic vector potential (MVP) method. The eddy current on blankets was subsequently analyzed. Finally, the Lorentz force applied on blankets can be calculated with cross-product of the eddy current field from transient magnetic analysis and the magnetic field from the static magnetic analysis. The EM forces will be used as mechanical loads for the future structural analysis of blanket under integrated load.

Gas Gun Model and Comparison to Experimental Performance of Pipe Guns Operating with Light Propellant Gases and Large Cryogenic Pellets

Injection of multiple large (~10 to 30 mm diameter) shattered pellets into ITER plasmas is presently part of the scheme planned to mitigate the deleterious effects of disruptions on the vessel components. To help in the design and optimize performance of the pellet injectors for this application, a model referred to as “the gas gun simulator” has been developed and benchmarked against experimental data. The computer code simulator is a Java program that models the gas-dynamics characteristics of a single-stage gas gun. Following a stepwise approach, the code utilizes a variety of input parameters to incrementally simulate and analyze the dynamics of the gun as the projectile is launched down the barrel. Using input data, the model can calculate gun performance based on physical characteristics, such as propellant-gas and fast-valve properties, barrel geometry, and pellet mass. Although the model is fundamentally generic, the present version is configured to accommodate cryogenic pellets composed of H2, D2, Ne, Ar, and mixtures of them and light propellant gases (H2, D2, and He). The pellets are solidified in situ in pipe guns that consist of stainless steel tubes and fast-acting valves that provide the propellant gas for pellet acceleration (to speeds ~200 to 700 m/s). The pellet speed is the key parameter in determining the response time of a shattered pellet system to a plasma disruption event. The calculated speeds from the code simulations of experiments were typically in excellent agreement with the measured values. With the gas gun simulator validated for many test shots and over a wide range of physical and operating parameters, it is a valuable tool for optimization of the injector design, including the fast valve design (orifice size and volume) for any operating pressure (~40 bar expected for the ITER application) and barrel length for any pellet size (mass, diameter, and length). Key design parameters and proposed values for the pellet injectors for the ITER disruption mitigation systems are discussed.

Engineering analyses of Divertor Thomson scattering system

The ITER Divertor Thomson Scattering system is designed to provide an instrument capable of measuring the profiles of electron temperature and relative profiles of electron density in the outer divertor plasma. Main components of construction are front rack, back rack, neutron shielding and diagnostic mirrors.The whole Divertor Thomson Scattering system has been analyzed to be consistent with thermal, mechanical, electromagnetic and seismic loads which can affect construction. All the obtained results, such as displacement and stress maps, temperature distribution and values of forces and torques in construction during plasma disruption events, can lead to some significant changes in construction design in order to improve the system.

Design and testing of Crowbar Protection System for the JT-60SA superconducting magnet power supplies

The Crowbar System is a key protection system for the Super Conducting Magnet Power Supplies (SCMPSs) of JT-60SA, it is a Safety Relevant component and it is designed to be seismic resistant. The Crowbar System is operated in case of faults, quench of superconducting coils, plasma disruption or earthquake event, and it has to carry the current flowing through the superconducting coils in order to protect them and the SCMPS from over-voltages. The circuit configuration is based on a hybrid-making-switch, composed by a bidirectional-static-switch with a mechanical-switch and a varistor in parallel. The static-switch is composed by eight thyristors in bidirectional configuration (four connected in parallel and four in antiparallel), and it is designed to be able to safely intervene even if one thyristor or the mechanical-switch is not operating. The mechanical-switch (rated 50 kA making capacity) takes some tens of milliseconds to close, limiting the conduction time of thyristors. The overvoltage protection is performed by a voltage transducer, a Break-Over-Diode (BOD) and a varistor assuring a fast intervention with a high redundancy protection level.

Mechanical analyses of the ITER electron cyclotron upper launcher first confinement system

The Electron Cyclotron Upper Launcher is an eight beamline ITER antenna whose main goal is to drive current locally, inside magnetic islands that may form on the q=3/2 or 2 rational magnetic flux surfaces, in order to stabilize neoclassical tearing modes (NTMs). The primary vacuum boundary at the port plug extends into the port cell region through the ex-vessel mm-wave waveguide components, defining the so-called First Confinement System (FCS). Thermal expansion, seismic events and plasma disruption events result in displacements of the vacuum vessel, relative to the tokamak building, that are transferred to the FCS at its interfaces with the port plug. In absence of suitable inline waveguide bellows, the adaptation to such imposed displacements is provided by bending compliance of the straight waveguide sections. An analysis methodology for the ex-vessel components was developed which covers a preliminary set of applicable load combinations throughout the system life-cycle. This methodology describes the FCS numerical model developed to assess the system against independent events as well as the load combination strategy followed to combine the results. Some simplifications like the simulation of the Vertical Displacement Events (VDEs) by a static approach are considered in order to develop this methodology. The analysis of the results shows that the miter bends and the pieces of waveguides rigidly attached to the port cell ceiling are the most demanded components for most of the considered load combinations.

Solidification and Acceleration of Large Cryogenic Pellets Relevant for Plasma Disruption Mitigation

The technology for producing, accelerating, and shattering large pellets (before injection into plasmas) for disruption mitigation has been under development at the Oak Ridge National Laboratory for several years, including a system on DIII-D that has been used to provide some significant experimental results. The original proof-of-principle testing was carried out using a pipe gun injector cooled by a cryogenic refrigerator (temperatures $\sim 8$ –20 K) and equipped with a stainless steel tube to produce 16.5-mm pellets composed of either pure D2, pure Ne, or a dual layer with a thin outer shell of D2 and core of Ne. Recently, significant progress has been made in the laboratory using that same pipe gun and a new injector that is an ITER test apparatus cooled with liquid helium. The new injector operates at $\sim 5$ –8 K, which is similar to temperatures expected with cooling provided by the flow of supercritical helium on ITER. An alternative technique for producing/solidifying large pellets directly from a premixed gas has now been successfully tested in the laboratory. Also, two additional pellet sizes have been tested recently (nominal 24.4 and 34.0 mm diameters). With larger pellets, the number of injectors required for ITER disruption mitigation can be reduced, resulting in less cost and a smaller footprint for the hardware. An attractive option is longer pellets, and 24.4-mm pellets with a length/diameter ratio of $\sim 3$ have been successfully tested. Since pellet speed is the key parameter in determining the response time of a shattered pellet system to a plasma disruption event, recent tests have concentrated on documenting the speeds with different hardware configurations and operating parameters; speeds of $\sim 100$ –800 m/s have been recorded. The data and results from laboratory testing are presented and discussed, and a simple model for the pellet solidification process is described.

Mechanical analyses of the waveguide flange coupling for the first confinement system of the ITER electron cyclotron upper launcher

The four electron cyclotron (EC) upper port antennas (or “upper launchers” —UL) will be used to drive current locally inside magnetic islands located at the q=2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs), as well as heat inside of ρ of about 0.4. Each antenna consists of eight beam lines that are designed for the transmission of 1.5 MW of mm-wave power at 170GHz. The First Confinement System (FCS) is formed by the ex-vessel mm-wave waveguide components, for which SIC-1 classification requirements apply.The beam lines in the FCS comprise a Z shaped set of straight corrugated waveguides with a nominal diameter of 50mm connected by miter bends. This system is subjected to imposed displacements coming mainly from the thermal expansion of the vacuum vessel, seismic events and/or plasma disruption events. In absence of suitable SIC-1 waveguide bellows, the FCS waveguides must provide the necessary mechanical functional compliance. This has required the development of a dedicated, flange type coupling system with double metallic seals, capable of resisting the generated external loads while maintaining vacuum tightness and alignment. This paper presents the results of the design, analysis and pre-qualification experimental work done on the waveguides and the integrated SIC-1 compliant coupling system.

Transient Electromagnetic Analysis of Blanket Modules 14 and 15 in Different Sectors of the ITER Blanket System due to Plasma Disruption

The ITER blanket system provides shielding of the plasma controlling field coils and vacuum vessel from the plasma heat flux as well as nuclear heating from the plasma. In addition to the thermal requirements the blanket module attachment scheme must withstand the electromagnetic forces that occur during possible plasma disruption events. During a plasma disruption event eddy currents are induced in the blanket module (first wall and shield block) and interact with the large magnetic fields to produce forces which could potentially cause mechanical failure. For this reason the design and qualification of the ITER blanket system requires appropriate high-fidelity electromagnetic simulations that capture the physics of these disruption scenarios.The key features of the analysis procedure will be described including the modeling of the geometry of the blanket modules and the plasma current during disruption.The electromagnetic calculations are performed using the Opera-3d software. This software solves the transient 3D finite element problem from which the eddy currents are calculated. The electromagnetic loads due to these eddy currents are then calculated and translated to the local coordinate system of the blanket module of interest.

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

Transient Electromagnetic Analysis of Blanket Module 1 of the ITER Blanket System due to Plasma Disruption

The prediction of electromagnetic loads on blanket module 1 of the ITER device during a plasma disruption event is considered. This analysis is performed for a number of design variations (of the blanket module) and different disruption events.The key features of the analysis procedure will be presented including the geometric description of the blanket module composed of a first wall, shield block, and vacuum vessel. The modeling of the plasma current will also be described.The electromagnetic analyses are performed using the Opera-3d software. The transient eddy currents are first calculated, from which the electromagnetic loads are determined. Once these loads have been calculated they can also be exported for additional post-processing to assess the mechanical loading effects.

Electromagnetic analyses on radial field sector coils for JT-60SA

In the framework of EU – JA “Broader Approach” Agreement, the JT-60SA satellite tokamak will be built in Naka, Japan. In JT-60SA, to achieve the steady-state high beta plasma, suppression of Resistive Wall Modes (RWM) is necessary. Therefore, a passive stabilizing plate and a RWM active control system based on in-vessel coils are foreseen for JT-60SA. In the present design, this system consists in 18 sector coils, 6 in the toroidal, 3 in the poloidal direction, rated for a maximum current of 2.5 kA and fed by a dedicated fast power supply system. The coils shall produce sufficient radial magnetic field at the plasma edge, and the active control system shall be fast enough to react in real time to the plasma instabilities. In addition, during fast transients of the plasma current, significant loop voltage could be induced along the sector coils, which could result in over-currents in the coils and in the power supplies. This paper deals with the electromagnetic analyses on the sector coils of JT-60SA, with the main aim to determine the voltage/current characteristic of the coils and, at the same time, to have a first estimation of amplitude and shape of the produced magnetic field; different sector coil design solutions were considered and evaluated. The studies on the coil over-currents during plasma disruption or vertical displacement events are also presented. The results give useful indications in support to the development of the sector coils and power supply design and the identification of their electrical and functional requirements.

ITER vacuum vessel design and construction

According to recent design review results, the original reference vacuum vessel (VV) design was selected with a number of modifications including 3D shaping of the outboard inner shell. The VV load conditions were updated based on reviews of the plasma disruption and vertical displacement event (VDE) database. The lower port gussets have been reinforced based on structural analysis results, including non-linear buckling. Design of in-vessel coils for the mitigation of edge localized modes (ELM) and plasma vertical stabilization (VS) has progressed. Design of the in-wall-shielding (IWS) has progressed in details. The detailed layout of ferritic steel plates and borated steel plates is optimized based on the toroidal field ripple analysis. The procurement arrangements (PAs) for the VV including ports and IWS have been prepared or signed. Final design reviews were carried out to check readiness for the PA signature. The procedure for licensing the ITER VV according to the French Order on Nuclear Pressure Equipment (ESPN) has started and conformity assessment is being performed by an Agreed Notified Body (ANB). A VV design description document, VV load specification document, hazard and stress analysis reports and particular material appraisal were submitted according to the guideline and RCC-MR requirements.

The theory of the kink mode during the vertical plasma disruption events in tokamaks

This paper explains the existence of the m∕n=1∕1 kink mode during the vertical disruption event when the plasma has an electrical contact with the plasma facing conducting surfaces. It is shown that the kink perturbation can be in an equilibrium state even with a stable safety factor q>1, if the surface currents, excited by the kink mode, can flow through the conducting structure. The present theory suggests an explanation of the so-called sideway forces on the tokamak in-vessel components during the disruption event. For the general case of such wall touching kink modes, the energy principle is derived, which includes the current sharing effect.

Electromechanical analysis of the ITER ion cyclotron antenna structure and components

3D finite-element models of the ITER ion cyclotron RF antenna assembly have been implemented in order to perform electromagnetic and stress analyses and to calculate the applied electromagnetic loads, torques and forces on the antenna structure during plasma disruption events. The electromagnetic analysis has been carried out as follows: first the antenna outer structure is analyzed (support structure, neutron shield and Faraday shield), the magnetic field variation penetrating the outer structure is then computed and applied as forcing field to the antenna inner structure (four port junction, coaxial lines and straps). The stress analysis is carried out, for both the inner and outer structure, by computing in any element the cross product of the eddy current distribution with the local static magnetic field and by deriving stresses and deformations. Results of the analyses have shown that mechanical stresses are below the allowable limits in all the components of the antenna except the Faraday shield which needs some design improvements because of the very large displacements and stresses.

Ignitor plasma chamber structural design with dynamic loads due to plasma disruption event

During a plasma disruption in Ignitor [B. Coppi, IGNITOR Project Group, Engineering Design Description, Ignitor Programme, RLE Report no. PTP 96/03, December 1996], the electromagnetic (EM) loads are so large to require a dynamic elastic–plastic structural analysis of the plasma chamber (PC). A 30° sector of PC has been modeled. Static, dynamic and low cycle fatigue analyses under eddy and halo currents have been performed. It is shown that a PC wall thickness optimization limits the displacements within the clearance with toroidal coil. The fatigue lifetime is longer than the number of disruption events at the maximum performance.

Thermal performance of TF coils in HT-7U under plasma disruption

One of the most severe events for HT-7U TF coils is the plasma disruption event. Within a short period of the plasma disruption, the case and the coils will be heated up by the eddy current and coupling loss induced by the magnetic field variation. Certain amount of the deposited heat in coil cases would be conducted to the coil bundles. In this paper, the center plasma disruption is modeled and electromagnetic field is calculated by ANSYS software. The generated heat is thus directly coupled to the transient thermal analysis with the same software. It was found when the plasma current in normal condition is 1.5 MA and the disruption time constant is 3 ms, the eddy current will heat the case up to a temperature increase of about 15 K in the middle of the inner straight leg. The heat transferred from case to the cable can warm up the helium in the CIC bundle to a temperature increase of about 0.4 K 10 s after plasma disruption. The coupling loss in the coils is evaluated based on the magnetic field variation and the coupling time constant of the CICC cable. By applying GANDALF for the stability calculation, it is shown the coupling loss due to plasma disruption can reduce the temperature margin of coil by 0.63 K.

Effects of plasma disruption events on ITER first wall materials

In ITER, plasma disruption events may occur producing large fast thermal transients on plasma facing materials. Particularly important for the integrity of the first wall (FW) are relatively ‘long’ duration off-normal events such as plasma vertical displacement events (VDE) and runaway electrons (RE). An analytical methodology has been developed to specifically assess the effect of these events on FW plasma facing materials. For the typical energy densities and event duration expected for the primary and baffle FW, some melting and evaporation of the FW armor will occur without the beneficial effect of vapor shielding, and the metallic heat sink may also be damaged due to over-heating. The method is able to calculate the amount of melted and evaporated material, taking into account the evolution of the evaporated and melted layer and to evaluate possible effects of local temporary loss of cooling. The method has been used to analyze the effects of VDE and RE events for ITER, to study recent disruption simulation experiments and to benchmark experimental and analytical results.

High power density self-cooled lithium–vanadium blanket

A self-cooled lithium–vanadium blanket concept capable of operating with 2 MW/m 2 surface heat flux and 10 MW/m 2 neutron wall loading has been developed. The blanket has liquid lithium as the tritium breeder and the coolant to alleviate issues of coolant breeder compatibility and reactivity. Vanadium alloy (V–4Cr–4Ti) is used as the structural material because it can accommodate high heat loads. Also, it has good mechanical properties at high temperatures, high neutron fluence capability, low degradation under neutron irradiation, good compatibility with the blanket materials, low decay heat, low waste disposal rating, and adequate strength to accommodate the electromagnetic loads during plasma disruption events. Self-healing electrical insulator (CaO) is utilized to reduce the MHD pressure drop. A poloidal coolant flow with high velocity at the first wall is used to reduce the peak temperature of the vanadium structure and to accommodate high surface heat flux. The blanket has a simple geometrical configuration and low coolant pressure to reduce the fabrication cost, to improve the blanket reliability, and to increase confidence in its performance. Spectral shifter and reflector are utilized to improve the blanket shielding capability and energy multiplication, and to reduce the radial blanket thickness. Natural lithium is used to avoid extra cost related to the lithium enrichment process.

ITER reference breeding blanket design

The ITER reference breeding blanket design is water-cooled and is characterised by the use of the neutronic multiplier and breeder materials in the form of pebbles. Besides the achievement, with margin, of the tritium breeding ratio (TBR) minimum requirement, it exhibits an internal layout allowing it to withstand properly electromagnetic loads during plasma disruption and vertical displacement events, and pressure loads in case of rupture of an internal cooling channel (i.e. in-box LOCA). During the first part of 1998, the design has been optimised improving the performance in terms of TBR, enlarging the design margins with respect to the dimensioning loads and investigating in detail the global behaviour of the system during normal and off-normal conditions.

Magnetomechanical coupling: a robust approach for studying fusion machines behaviour during plasma disruptions

The mechanical design of the back-plate component of a Tokamak type fusion machine is strongly influenced by the magnitude of the electromagnetic forces induced by plasma disruption events. The purpose of this paper is to give a methodology of interfacing two distinct codes: an electromagnetic one (CARIDDI) and a mechanical one (ABAQUS) for the solution of problems concerning back-plate modelling during plasma disruptions. Finally, the results of the transient magnetomechanical computer simulations are described.