Plasma burn-through simulations using the DYON code and predictions for ITER

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied or of the allowable stress() in RCC-MR code at the maximum stress region in the FW.

This article describes a neutral particle analyzer/isotope separator (ISEP) developed for measurement of the relative hydrogen isotope composition of Joint European Torus (JET) plasmas. The ISEP deployed on the JET can be regarded as a prototype of an instrument proposed for measurement of the spatial profile of the ratio of the density of deuterium and tritium ions in the plasma, nD(r)/nT(r), in the International Thermonuclear Experimental Reactor (ITER). The ISEP makes simultaneous measurements of the energy distribution of efflux of hydrogen isotope atoms (H, D, and T) from the plasma. From such measurements it is possible to deduce the radial profile of the relative hydrogen isotope ion composition of the plasma and radial transport of ions of one isotope across the plasma of another isotope species. The main elements of the ISEP are (a) use of a thin carbon foil for reionization of the incident atoms, thereby eliminating gas stripping cells and gas sources of conventional neutral particle analyzers (NPAs), (b) acceleration of secondary ions in order to access the regime of higher detection efficiency of the NPA and to better separate ion pulses from neutron/γ-ray induced pulses in scintillator detectors, (c) E‖B dispersion of the secondary ions in specially designed nonuniform magnetic and electric fields to provide focusing in the detector plane, increased throughput and greater contrast between neighboring isotopes, and (d) counting of energy and mass analyzed secondary ions using detectors consisting of thin [1⩽t (μm)⩽7] CsI(Tl) scintillators deposited directly on miniature thin window photomultiplier tubes mounted in vacuum. The ISEP has high contrast between atoms of neighboring masses (⩾103 for E≈5 keV and much greater at higher energies), and high detection efficiency (0.06⩽ε⩽0.83 for atoms of 5⩽ (keV)⩽150. ISEP detectors have very low sensitivity to neutrons and γ rays (⩽10−7 of ion sensitivity), making it feasible to use the ISEP in JET DT experiments without any shielding. Only a modest amount of neutron/γ-ray shielding would be required in the ITER for similar applications of the ISEP. The initial experiments on JET plasmas using the ISEP demonstrate well the capabilities of the instrument for measurement of the hydrogen isotope composition of the plasma and the energy distribution function of isotope ions.

A Compact Neutron Spectrometer (CNS) has been developed to measure the neutron emission spectra in Joint European Torus (JET) fusion plasma experiments. The spectrometer, based on a liquid scintillation detector (BC501A), is equipped with a Digital Pulse Shape Discrimination (DPSD) acquisition system for neutron (n) and gamma-ray (γ) separation. The CNS enables recording the n and γ pulse height spectra (PHS) up to total count rates of ~10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">6</sup> s <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">-1</sup> . Energy resolution, after PHS unfolding, will be <; 2% for 14 MeV neutrons and <; 4% for 2.5 MeV neutrons. The work done by ENEA-Frascati and Physikalisch-Technische Bundesanstalt (PTB) respectively in the assembly and test of DPSD and scintillation detector, along with the first results obtained by the spectrometer in JET plasma experiments are presented. The experience obtained with CNS in JET will contribute to the development of neutron spectrometers suitable for applications in the International Thermonuclear Experimental Reactor (ITER).

Since the start of the ITER (International Thermonuclear Experimental Reactor) design effort, safety experts have attempted to guide designers in developing a machine that will demonstrate the potential safety and environmental advantages of fusion. Offsite dose limits (based on regulatory requirements) have been translated into release rates or radiological inventories, which provide useful guidance to designers. The ITER team is attempting to protect the public, to the extent possible, against major accidents with passive means, that is, without relying on active systems. The concept of passive safety is developed by the US safety team for internal events and for external events that are within the envelope of design-basis accidents selected for the ITER. In addition to overall safety guidance, implications of passive safety for ITER systems (as envisioned by the US safety team) are discussed. An approach to implementing passive safety in the ITER with reference to the blanket, plasma-facing components, fuel-cycle systems, plasma shutdown, and magnet design is described. The need for a confinement building is addressed, and the role of such a building in the context of passive safety is also discussed. >

Given the urgent need to converge on precise guidelines for accident management in nuclear fusion plants, in this paper, the authors will analyze the problem related to the choice of possible candidate materials for the nuclear fusion plants like International Thermonuclear Experimental Reactor (ITER), DEMOnstration power plant (DEMO), or PROTOtype power plant (PROTO). Fusion power is a promising long-term candidate to supply the energy needs of humanity. From the safety point of view, nuclear fusion holds inherent and potential safety advantages over other energy sources. In magnetic confinement devices, the plasma edge and surrounding material surfaces provide a buffer zone between the hightemperature conditions in the plasma core and the normal “terrestrial” environment. The interaction between the plasma edge and the surrounding surfaces profoundly influences the conditions in the plasma core and is the principal key engineering issue. Robust solutions to plasma–material interactions (PMIs) issues are required to realize a commercially attractive fusion reactor. PMIs critically affect tokamak operation in many ways. Erosion by the plasma determines the lifetime of plasma-facing components (PFCs) and generates a source of impurities,which cool and dilute the plasma. Deposition of material onto PFCs alters their surface composition and can lead to long-term accumulation of large in-vessel tritium inventories. Retention and recycling of hydrogen from PFCs affect fueling efficiency, plasma density control, and the density of neutral hydrogen in the plasma boundary, which impacts particle and energy transport. Dusts are currently produced in the existing plants like JET (and will also be in the future ones like ITER, DEMO, and PROTO) by PMIs. Thus, the issues related to the PFCs of the first wall of the nuclear fusion plants area topic that the Quantum Electronics and Plasma Physics Research Group have been studying for more than a decade. Regarding the selection of PFCs, there are two strong school of thoughts that work on fusion plants materials for the first wall: (1) one composed by experts working mainly on nuclear fusion and that are focusing their attention on certain materials (like tungsten, beryllium, carbon, tritium, stainless steel, and other), and (2) the other one composed of experts working mainly on the fission plants that want to demonstrate the “dual-use” characteristics of the “fission materials” for fusion plants applications. In this review paper, the authors will (1) review both the approaches mentioned above comparing the main characteristics of all the proposed and tested materials, (2) analyze the consequences (experimentally revealed and numerically predicted) of PMIs that provokes serious damages to the plants and dangerous consequences like radioactive/toxic dust production, and (3) analyze (also thanks to our experimental activities on STARDUST-Upgrade facility) the risks (for the operators) related to the dust re-suspended in case of accidents. The authors will critically comment each section of the paper together with a detailed state of the art.

The major purpose of the International Thermonuclear Experimental Reactor (ITER) is to develop an experimental fusion reactor for the scientific and technological feasibility demonstration of fusion power. ITER Conceptual Design Activites (CDA) began in the spring of 1988 through the united efforts of the European Community, Japan, the Soviet Union and the United States and ended in December 1990. The objectives of this phase were to develop the design of ITER, to perform a safety and environemental analysis, and to define, the future research and development. This paper gives the design concepts developed for the ITER basic devices such as containment structures, remote handling tools, superconducting coils and plasma facing components.

An accurate and reliable measurement of poloidal β is essential to assess the performances of Joint European Torus (JET). The diamagnetic loop can measure β values as low as 0.1 in JET discharges with a plasma current larger than 2×106 A. The instrumentation used includes a flux loop rigidly fitted on a toroidal field (TF) coil, a large Rogowski coil measuring the TF busbar current, and a displacement gauge measuring the TF coil expansion. The fluxes to be compensated originate, in order of importance, from the TF current, the eddy current in the vessel, the TF coil expansion, and the stray coupling with the poloidal fields. The TF and eddy currents must be particularly well compensated on JET since the plasma current starts before the toroidal field has reached its plateau value. Comparison between the diamagnetic and other evaluations of β shows a good agreement.

Plasma control at JET

The International Thermonuclear Experimental Reactor (ITER) is an international project aimed to build a fusion reactor using a plasma magnetic confinement (Tokamak) in Cadarache, France, which will demonstrate that such a machine can produce at least 10 times more energy than the one spent to sustain the fusion reaction. The project involves 7 partners: China, European Union (EU), India, Japan (JA), South Korea, Russian Federation and United States of America, all of whom will provide “in-kind” contributions to the central ITER Organization (IO) in the form of components required to build the machine. Each of the seven partners is represented by a Domestic Agency (DA) in order to comply with their in-kind contributions. The European Joint Undertaking for ITER and the Development of Fusion Energy or 'Fusion for Energy' (F4E) is a type of European organization known as a Joint Undertaking created under the Euratom Treaty by a decision of the Council of the European Union and it is the ITER European DA . F4E has three main objectives: a) Providing European contributions to the ITER international fusion energy research project being built in Cadarache, France; b) Providing European contributions to a number of joint projects with Japan that aim to accelerate the development of fusion - the Broader Approach; c) Coordinating a program of activities to prepare for the first demonstration fusion reactors that can generate electricity. ITER superconducting magnet system, working at 4.5K, consists of 18 Toroidal Field (TF) coils, 6 Central Solenoid (CS) modules, 6 Poloidal Field (PF) coils and 18 Correction Coils (CC). F4E is responsible for the procurement of about 25% of the magnet system. In this paper we are reporting on the main challenges to be faced during the manufacturing processes in terms of weld and their inspection.

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

The merits of a mass spectrometer with crossed electric and magnetic fields are discussed. The electric field is assumed to be a toroidal one (field index c) and the magnetic field homogeneous. In the case c=-1 the stigmatic double focusing is obtained and the Barber's rule holds. If the toroidal electric field is produced by a cylindrical condenser having a pair of Matsuda plates, the directional focusing can be adjusted by changing the electric potential applied to the Matsuda plates and the velocity focusing can be adjusted by changing the ratio of electric and magnetic fields independently to each other. Therefore, all focal adjustments can be done electrically and no mechanical adjustment is necessary. This type of mass spectrometer works also as a single focusing apparatus under very weak magnetic field if appropriate potentials are applied to the condenser electrodes and Matsuda plates.

Fuel retention in carbon plasma-facing components (PFCs) is such a major concern for next-step operation that it could prevent the use of this material in the D-T phase of ITER. Because of its complete set of actively cooled PFCs, Tore Supra offers a unique opportunity to study this phenomenon in conditions where the plasma exposure time is much longer than the thermal equilibration time of the PFCs. In addition to the main characteristics of permanent retention measured during long-discharge operation, this paper discusses the different mechanisms possibly at work in the continuous increase of the in-vessel inventory and describes the morphology and physical properties of the deposits found at several locations in the vacuum chamber. The main results are (1) that D retention mainly depends on the lower hybrid power coupled to the plasma and, to a lesser extent, on the edge temperature and fueling method, (2) that permanent D retention is mainly due to codeposition, and (3) that the hydrogenated carbon deposits present at the surface of the different PFCs are strongly disorganized graphite carbons when they are exposed to high heat fluxes, whose formation occurs through a heterogeneous growth involving both codeposition of nanoparticles and basic structural unit vapor condensates.

On the Joint European Torus (JET), a key source of calibrated electron temperature Te profiles is from the measurements of the full electron cyclotron emission (ECE) spectrum made by a Fourier transform spectrometer (FTS). It is absolutely calibrated by using a calibration source inside the vacuum vessel. High spatial and temporal resolution ECE Te profiles are obtained using a 96-channel heterodyne instrument which is cross-calibrated on each JET pulse against the data from the FTS system. Residual systematic frequency dependent errors at the 5%–10% level can then be evaluated and corrected for, using specific discharges in which the toroidal field is varied while keeping the shape of the Te profile constant. This improvement in the calibration method has been systematically applied at JET for the first time improving both the smoothness and the symmetry of the Te profiles. The consequences of this improvement are discussed. In addition, it is shown that no deviation occurs in the FTS calibration for more than eight years, which is relevant for the International Thermonuclear Experimental Reactor.

Edge Localized Mode Control in TCV

A study is presented for the stability of alpha-particle driven shear Alfven Eigenmodes (AE) for the normal parameters of the three major burning plasma proposals, ITER (International Thermonuclear Experimental Reactor), FIRE (Fusion Ignition Research Experiment), and IGNITOR (Ignited Torus). A study of the JET (Joint European Torus) plasma, where fusion alphas were generated in tritium experiments, is also included to attempt experimental validation of the numerical predictions. An analytic assessment of Toroidal AE (TAE) stability is first presented, where the alpha particle beta due to the fusion reaction rate and electron drag is simply and accurately estimated in 7-20 keV plasma temperature regime. In this assessment the hot particle drive is balanced against ion-Landau damping of the background deuterons and electron collision effects and stability boundaries are determined. Then two numerical studies of AE instability are presented. In one the High-n stability code HINST is used . This code is capable of predicting instabilities of low and moderately high frequency Alfven modes. HINST computes the non-perturbative solution of the Alfven eigenmodes including effects of ion finite Larmor radius, orbit width, trapped electrons etc. The stability calculations are repeated using the global code NOVAK. We show that for these tokamaks the spectrum of the least stable AE modes are TAE that appear at medium-/high-n numbers. In HINST TAEs are locally unstable due to the alphas pressure gradient in all the devices under the consideration except IGNITOR. However, NOVAK calculations show that the global mode structure enhances the damping mechanisms and produces stability in all configurations considered here. A serious question remains whether the perturbation theory used in NOVAK overestimates the stability predictions, so that it is premature to conclude that the nominal operation of all three proposals are stable to AEs. In addition NBI ions produce a strong stabilizing effect for JET. However, in ITER the beam energies needed to penetrate to the core must be high so that a diamagnetic drift frequency comparable to that of the alpha particles is produced by the beam ions which induces a destabilizing effect.