The ITER design

Abstract

In 1998, after six years of joint work originally foreseen under the ITER engineering design activities (EDA) agreement, a design for ITER had been developed fulfilling all objectives and the cost target adopted by the ITER parties in 1992 at the start of the EDA. While accepting this design, the ITER parties recognized the possibility that they might be unable, for financial reasons, to proceed to the construction of the then foreseen device. The focus of effort in the ITER EDA since 1998 has been the development of a new design to meet revised technical objectives and a cost reduction target of about 50% of the previously accepted cost estimate. The rationale for the choice of parameters of the design has been based largely on system analysis drawing on the design solutions already developed and using the latest physics results and outputs from technology R&D projects. In so doing the joint central team and home teams converge towards a new design which will allow the exploration of a range of burning plasma conditions. The new ITER design, whilst having reduced technical objectives from its predecessor, will nonetheless meet the programmatic objective of providing an integrated demonstration of the scientific and technological feasibility of fusion energy. Background, design features, performance, safety features, and R&D and future perspectives of the ITER design are discussed.

Operational conditions in a W-clad tokamak

“…These properties would lead to a considerable higher load on the support structures if bulk tungsten tiles were used. Although the higher forces would be technologically manageable (see for example the ITER divertor design [1]), they would make the transition from a device designed for C-based PFCs very costly and time consuming. Therefore ASDEX Upgrade has chosen the coating solution.…”

“…PFCs in present day devices are mostly designed to optimise fusion performance, partially neglecting the technical needs of a future fusion power plant. Even the next step device, ITER [1], follows a conservative approach using beryllium for the main chamber PFCs in order to minimise the risk of high power loss through impurity radiation in the central plasma. In a reactor however, Be will not be a viable solution due to its high erosion yield and high-Z components may have to be used [2].…”

Final steps to an all tungsten divertor tokamak

“…This is especially true for the plasma wall interaction since the extrapolation from present day devices towards ITER and beyond is largest. ITER [1] still follows a conservative approach using beryllium for the main chamber PFCs in order to minimise the risk of strong power loss through impurity radiation in the central plasma. In a reactor however, Be will not be a viable solution due to its high erosion yield and high-Z components may have to be used [2,3].…”

Hydrogen retention in ITER relevant mixed material layers

“…beryllium (Be) is the primary candidate for the first wall material, whereas tungsten (W) is chosen for the divertor armour, except for the area around the strike points where carbon fibre composite (CFC) is designated as armor material [1]. Based on observations in present fusion devices, it is expected that the wall material will be eroded due to the interaction with plasma particles and subsequently deposited on other material surfaces.…”