4th WCSET-2015 at Japan
Special Session (Nuclear Fusion):
Title:
Options in the design of helical fusion reactor FFHR-D1
and C1
Authors:
J. MIYAZAWA, T. Goto, N. Yanagi, T. Tanaka, H. Tamura,
S. Masuzaki, J. Yagi, R. Sakamoto, A. Sagara
Abstract: The
design activity of a helical fusion reactor FFHR has
been conducted in NIFS (National Institute for Fusion
Science, Japan), based on the characteristics of
high-performance plasmas achieved in LHD (Large Helical
Device), which is the world largest superconducting
plasma confinement device using the helical magnetic
fields called heliotron. The FFHR has several options
named d1A (the basic design similar to LHD), d1B (w/
higher magnetic field to mitigate the requirement for
plasma), d1C (w/ an innovative magnetic configuration),
and c1 (w/ smaller size). Plasma operation scenarios
including the plasma physics assessment depending on the
device parameter are being investigated intensively. In
general, the central part of a fusion device is composed
of the vacuum vessel including the plasma facing
components called the first wall and the divertor, the
superconducting (SC) magnet coils, the tritium-breeding
blanket, and the neutron-shielding blanket. Each of
these components also has several options. As for the SC
coils, ReBCO (high-temperature SC material working at
~20 K) cooled by gas helium is considered as the first
option. A tritium-breeding blanket system using a molten
salt of FLiNaBe (melting point is ~600 K) is also the
first option in FFHR. As for the divertor, adding to the
conventional option of tungsten mono-block divertor with
copper heat sink as in ITER, a new option of liquid
metal X-point divertor (LMXD) has been proposed
recently. This LMXD has a possibility to drastically
ease the routine maintenance when coupled with the novel
divertor design that is also a new concept.
Keywords: Helical Fusion
Reactor, FFHR, LHD, Heliotron, Superconducting Magnet
Coils, Blanket, Divertor
Pages:
012-012