SIV-2NIFS 
Helical reactor design and engineering research bases in NIFS FER ProjectAkio SAGARA
National Institute for Fusion ScienceFusion Engineering Research Project (FERP) has
been launched from the FY2010 in NIFS. This projectfocuses on both the conceptual design of a steady-statefusion demonstration reactor and various engineeringresearch and development, which are needed before
entering into the engineering design activities forDEMO as shown in Fig.1. Therefore, this project
consists of three research groups, (1) reactor systemdesign, (2) superconducting magnets, and (3) in-vesselcomponents, with the total 13 task and 44 sub-task
groups as shown in Fig.2.
The LHD-type device does not need plasma
current, and this excellent feature gives a great
advantage for realizing a steady-state reactor.
Therefore, along with a conceptual design of thehelical DEMO reactor FFHR-d1 by integrating designbases established so far on the designs of the FFHR
series for commercial power plants, the project iscarrying out research on key components, such as thesuperconducting coil system, high performance blanket,
first wall and divertors, and so on. As the center offusion engineering research for universities, the project
enhances domestic and international cooperation toadvance reactor design studies as well as to expand
basic research lying in interdisciplinary areas.
The large-scale superconducting magnet system
requires high-performance superconductors of 100
kA-class current capacities. Research is being
conducted for developing such an advanced conductor
using metallic low-temperature superconducting
materials
like
Nb3Sn
and
Nb3Al
and/orhigh-temperature superconducting materials likeYBCO. Actual environment testing is also carried out
to estimate the characteristics of superconducting
materials
underconditions
withcryogenic
temperatures, intense magnetic field and neutron
irradiation. Components in the magnet systems aresubject to huge electromagnetic force. Research is
ongoing so as to precisely evaluate the expected stresson component materials and to seek the optimum coilsupporting structure. The engineering design of thewinding and fabrication method for magnets is also inprogress.
It is essential in fusion blankets to use structural
materials whose radioactivity is low and decays swiftly
after irradiation by neutrons. Vanadium alloys are oneof the major candidate materials. A low-activation

vanadium alloy (NIFS-HEAT) was produced and its
evaluation has been carried out in collaboration withuniversities. Development of various components
using vanadium alloys is also progressing along withfurther improvement of the material. The experimental
results confirmed that creep deformation is effectively
suppressed by strengthening the material withhigh-density precipitation.
The blanket is a key component to shield neutrons,
to convert fusion energy to thermal output, and tobreed tritium fuels. In order to develop liquid breeder
blankets using molten salts or liquid metals,experimental investigations are being carried out on
the high temperature strength of materials and control
of corrosion, which are important for a long-life
blanket. One of the integration studies is constructing aforced convection type molten salt loop Orosh2i-1. In
this work, feasibility study on material science
includingchemistry,
fluid
dynamics
andthermo-mechanical engineering for the fusion blanket
are integrated.
Divertor heat flux in a steady-state fusion
demonstration reactor is considered to be maximum
10MW/m2 in a steady state. Here extremely highheat-tolerant divertor plates need to be developed.
Three important subjects in the research anddevelopment are material selection, development inbonding technology between armor tiles and coolant
systems, and design studies of the 3D-shape of thehelical divertor with sufficient pumping ducts.

Fig.1 NIFS FERP activities towards DEMO reactor

The first wall of a magnetic fusion reactor isrequired to maintain high vacuum as well as to act aspart of the blanket structure, facing the edge plasma,
under 14MeV neutron irradiation at elevatedtemperatures. The choice of material for the first wallis thus of critical importance for the successful
development of fusion energy.
Hydrogen isotopes such as deuterium and tritium
will be utilized as a fuel in fusion plants. Tritium is aradioactive isotope and therefore should be managedwith safety. The project includes development oftritium handling and safety technologies, such astritium decontamination and an advanced tritium
removal system. A sensitive detector is also underdevelopment for safety management of tritium. Those
collaboration experiments are performed using tritium
facilities of universities.
In the first three years, as the first round of design
integration, primary design parameters of FFHR-d1
have been selected with a helical system code byintroducing core plasma design with the Direct Profile
Extrapolation (DPE) from LHD experimental data andby reducing blanket thickness with advanced shielding
materials, resulting in reactor size optimization forblanket space and magnetic stored energy < 160GJ.
The detailed 3-D design of in-vessel components,
mechanical supporting structures, divertor pumping
configurations and replacing scenarios are in progressas the second round.
Intensive R&D works are also in progress onsuperconducting magnets, such as innovative
superconducting current leads, basic study on High
Temperature Superconductor (HTS), properties ofsuperfluid HeII, remountable joint of YBCO conductor,
HTS low porosity bulks, analysis of joints between
Cable in Conduit (CIC) conductors, high efficiency
cryocooler, analysis of a normal-zone propagation inthe LHD helical coil, next generation power devices,
properties of Cu addition MgB2 wires, series
compensatedthyristor
converters,inter-strand
resistance in CIC conductor, temperature control withhigh-precision, transposed tape conductors, and so on.
As for in-vessel components, low activation
vanadium alloys and ferritic steels were improved onhigh temperature mechanical properties and radiation
resistance by compositional and microstructuralcontrol such as addition of nano-particle dispersion andreduction of grain sizes by mechanical alloying.
Coating and surface modification methods wereinvestigated for MHD pressure drop mitigation, tritium
permeation reduction and corrosion protection for
liquid breeder blankets.
Regardingtritium
handlingandsafety
technologies, many kinds of investigations have beensuccessfully carried out as collaboration with many
universities, research institutes and companies onhydrogen isotope separation and removal technology,
tritium measurements, fueling in fusion reactor, safety
in environment, and so on.

Fig.2 NIFS FERP with the total 13 task and 44 sub-task groups.