ITER. Conceptual design of the ITER basic devices.

The current International Thermonuclear Experimental Reactor (ITER) design features a 6-m major radius, a 2.15-m minor radius, and a 22-MA plasma current. About 330 V-s in the poloidal field system inductively drive the current for hundreds of seconds. About 125 MW of neutral-beam, lower-hybrid, and electron-cyclotron power are provided for steady-state current drive and heating. Because the ITER design must account for the fact that the tokamak database is not exact, a reasonably cautious approach must be taken. There are some benefits of such an approach: the chances of success are enhanced, ITER traverses most of the distance toward the size and power of a commercial reactor, and it may be capable of more performance than promised. By extending the operating rules to the maximum expected range, the fusion power may be increased such that the machine might be capable of simulating a small net energy condition. If so, then considerable progress could be made toward fusion-power applications. >

Radiation shielding for superconductive toroidal field coils around the neutral beam injector duct in the ITER design

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. >

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ‘‘the Parties.’’ The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties’ four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

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.

The International Thermonuclear Experimental Reactor (ITER) represents a major step toward realizing the benefits of magnetic fusion energy. Four international parties, the European Communities, Japan, the Russian Federation and the United States are collaborating on the design of ITER and the development of the relevant technology. Because the long range goal is to construct a demonstration reactor, and ultimately to develop fusion power on a commercial basis, there is worldwide emphasis on enhancing industrial involvement throughout the Engineering Design Activities (EDA).

Consecutive absorption/desorption cycles of the ZrCo–H2 system were studied to simulate the real International Thermonuclear Experimental Reactor (ITER) hydrogen getter system. A ZrCo getter was used in this paper instead of the depleted uranium (DU) getter material, which has been recently considered as the hydrogen getter in ITER. In a cyclic pressure-composition isotherm (PCI) measurement, the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo–H2 system in detail, especially for the desorption stage. This high-pressure PCI apparatus, however, shows cause and effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo–H2 system or in case of the DU–H2 system having multiple getter bed battery, a similar affection by previous getter status might be related and a similar aspect could be shown to consider further ITER design; for example, a need for control logic from PCI measurements using a high-pressure Sievert apparatus.

ITER breeding blanket design for the enhanced performance phase

Effect of pressure on the radiation induced electrical conductivity of NBI insulator gases

This paper will discuss simulations of the full ionization process (i.e. plasma burn-through), fundamental to creating high temperature plasma. By means of an applied electric field, the gas is partially ionized by the electron avalanche process. In order for the electron temperature to increase, the remaining neutrals need to be fully ionized in the plasma burn-through phase, as radiation is the main contribution to the electron power loss. The radiated power loss can be significantly affected by impurities resulting from interaction with the plasma facing components. The DYON code is a plasma burn-through simulator developed at Joint European Torus (JET) (Kim et al and EFDA-JET Contributors 2012 Nucl. Fusion 52 103016, Kim, Sips and EFDA-JET Contributors 2013 Nucl. Fusion 53 083024). The dynamic evolution of the plasma temperature and plasma densities including the impurity content is calculated in a self-consistent way using plasma wall interaction models. The recent installation of a beryllium wall at JET enabled validation of the plasma burn-through model in the presence of new, metallic plasma facing components. The simulation results of the plasma burn-through phase show a consistent good agreement against experiments at JET, and explain differences observed during plasma initiation with the old carbon plasma facing components. In the International Thermonuclear Experimental Reactor (ITER), the allowable toroidal electric field is restricted to 0.35 (V m−1), which is significantly lower compared to the typical value (∼1 (V m−1)) used in the present devices. The limitation on toroidal electric field also reduces the range of other operation parameters during plasma formation in ITER. Thus, predictive simulations of plasma burn-through in ITER using validated model is of crucial importance. This paper provides an overview of the DYON code and the validation, together with new predictive simulations for ITER using the DYON code.

There is a need for a real time, reliable density measurement compatible with the restricted access and radiation environment on International Thermonuclear Experimental Reactor (ITER). Due to the large plasma path length, high density and field, refraction, and Faraday rotation effects makes the use of contemporary long wavelength (≳50 μm) interferometers impractical. In this paper we consider the design of a short wavelength vibration compensated interferometer which allows operation without a prohibitively large vibration isolated structure and permits the optics to be conveniently mounted directly in or on the tokamak. A density interferometer design for ITER incorporating a 10.6 μm CO2 interferometer with vibration compensation provided by a 3.39 μm HeNe laser is discussed. The proposed interferometer design requires only a small intrusion into the ITER tokamak without a large support structure, refraction and Faraday rotation problems are avoided, and it provides a density resolution of at least 0.5%. Results are presented from an interferometer installed on the DIII-D tokamak incorporating essential elements of the proposed ITER design including 10.6 and 3.39 μm lasers, a retroreflector mounted on the vacuum wall of the DIII-D tokamak and real-time density feedback control. In this paper we consider a short wavelength interferometer design that incorporates vibration compensation for use on ITER. Our primary concern is to develop an interferometer design that will produce a reliable real time density monitor. We use the ITER conceptual design activity report as the basis of the design.

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 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.

A conceptual design of a 4 K irradiation test facility has been conducted in support of the International Thermonuclear Experimental Reactor (ITER) magnet development program. A new research reactor designated as HANARO at the Korea Atomic Energy Research Institute has a Cold Neutron Source (CNS) port that is identified to be suitable for the fast neutron irradiation of metals and insulation materials for superconducting magnets at 4 K. A 40 hours of irradiation at full power will produce 2.5×1017 n/cm2 of the ITER magnet design neutron fluence with energy above 0.1 MeV. A material testing laboratory of Irradiated Materials Evaluation Facility (IMEF) that is located next to HANARO has been equipped with 77 K test machines and fracture analysis microscopes for radioactive specimens which can be upgraded for 4 K test without any intermediate warming. CNS radiation spectrum determined by Monte Carlo method is found to be more favorable for metal irradiation than for insulation materials with absorbed gamma dose that is 7–10 times the fast neutron dose. A lead-shielded irradiation capsule design with a 1 cm diameter specimen in 3 cm cold-bore diameter and 18 cm height will require about 120 watt cooling capacity at 4.6 K.

For the steady running of the first-wall (FW) in international thermonuclear experimental reactor (ITER), it’s essential to introduce highly effective heatsinks which can bear high heat flux and remove it up to 30MW m-2. Hypervapotron (HV) is one of the most promising devices in which the subcooled boiling plays a significant role. In this work, notable conclusions drawn by other researchers have been summarized, like “Vapotron effect” and “appearance of vortex”. Characteristics of different HV fabrications were compared at the same time. Especially, aiming at the optimization of Hypervapotron test section for FW in ITER, a series of selected HV heat transfer experiments were analysed thoroughly, including the working fluid parameters, structures of test section, CHF test results, etc. The most reasonable and reliable design of the test sections and corresponding parameters were chosen and optimized.