J. Plasma Fusion Res. SERIES, Vol. 5 (2002) 137-142

Status of the ITER Project
TSUNEMATSU Toshihide and NAGAMI Masayuki
Japan Atomic Energy Research Institute, Ibaraki 311-0193 Japan(Received: 14 December 2001 / Accepted: 10 Aprll 2002)

Abstract
The status of the ITER Project is presented. The objective of ITER is to demonstrate the scientificand technical feasibility of fusion energy for peaceful purposes. The Engineering Design Activities(EDA) was concluded in July 2001 to finalize the nine-years design and R&D activities. The design ofthe ITER has been developed not only to satisfy these requirements but also to have the flexibility ofplasma operations. This flexibility will provide an advantage for coping with uncertainties in the physics
database, in studying burning plasmas, in introducing advanced features. The govemmental negotiations
has been initiated by aiming at the ITER Construction, Operation, Exploration and Decommissioning
Activities in early 2003.

Keywords:
ITER, engineering design, R&D, operation region, negotiation

1. Introduction
The history of co-operation on ITER started in1985 when government leaders in summit meetings

and the detailed technical objectives and cost target asoriginally set. With the approaching end of the original

period set for the EDA, the Parties negotiated anagreement extending the EDA period for three years

called for more substantial international co-operation inorder to increase the efficiency and minimize the cost offusion power development. In response to the summit
initiatives, the Conceptual Design Activity (CDA) began
in 1988 after the four Parties formally accepted thewhich was signed by three Parties, Euratom, Japan andthe Russian Federation. The USA agreed to participate
for one additional year. At that time, by recognizing thepossibility that they might be unable, for financial
reasons, to proceed to the construction of the theninvitation from the Director-General of the IAEA toparticipate, in accordance with the terms of reference,
under the auspices of the IAEA. The CDA weresuccessfully completed in 1990. After completing theforeseen device, the Parties adopted a less demanding
set of detailed technical objectives that would still meetthe overall programmatic objectives but at aCDA, the four Parties, Euratom, Japan, Russian

significantly reduced cost (about 507o). The EDAFederation and the USA, entered into negotiations on
how ITER should proceed further, resulting in the ITER

extension period was primarily devoted to developing adesign under the revised technical objectives set in1998, as well as completion of planned technology R&D
projects. The results of these activities have beenreflected in the report of the ITER design [] presented
by the ITER Director in July 2001. At the last moment

Engineering Design Activity (EDA). The Agreement
was signed on July 21, 1992 in Washington byrepresentatives of the four Parties and entered
immediately into force. Joint work during the six yearsof the EDA period, Euratom, Japan and Russian

period originally set for the Agreement led, by July
1998 to the development of a design for ITER that wasadjudged to futfill the overall programmatic objective

Federation initiated a non-committal discussion on the@20O2 by The Japan Society of Plasma

Corresponding author's e-mail: toshit@ naka.jaeri. go.jp

Science and Nuclear Fusion Researchr37

Tsunematsu T. et a/., Status of the ITER Proiect

implementation of the construction, operation anddecommissioning of ITER and the formal negotiation

MeV neutron average power load on the first wall

has started in November 200lamone the above threea the option for later installation ofa tritium breeding> 0.5 MWm2 and fluence > 0.3 MWa/m2,

Parties and Canada.

blanket on the outboard of the device should not be
precluded.

2. ITER EDA Objectives3) Operation requirements:o the operation anticipated over an approximately 20year period should address the issues of burning

The purpose of the ITER Agreement is to produce
a detailed, complete, and fully integrated engineering
design of ITER and all technical data necessary forplasmas, steady-state operation and improved
modes of confinement, and testing of blanket

future decisions on the construction of ITER. Suchdesign and technical data shall then be available formodules.

each of the Parties to use either as part of an
intemational collaborative programme or in its domesticprogramme.

3. Validating Research and Development

2.1 Programmatac obiectaves

use established approaches and to validate their

The overall programmatic objective of ITER is todemonstrate the scientific and technological feasibilityof fusion power for peaceful purposes. ITER would

The overall design philosophy for ITER has been toignition and extended burn of deuterium-tritium

application through wide-ranging physics andengineering work. This included detailed analyses,experiments in existing fusion research facilities, anddedicated technology developments including
fabrication and test of full scale or scalable models ofplasmas, with steady-state as an ultimate goal, bydemonstrating technologies essential to a reactor in an
integrated system, and by performing integrated testingof the high-heat-flux and nuclear components required
to utilize fusion energy for practical purposes.

key components [2]. Unavoidable uncertainties remainin the extrapolation of performance from current
experience to the ITER size and parameters; these canonly be fully resolved through experiments at ITER
scale. On a voluntary basis, the Parties have conducted

2.2 Technical objectives

well-focused physics investigations that strengthen
further the physics database, reduce the ranges of
uncertainty in extrapolation and explore wider options
for possible ITER operation. The results of ITER

accomplish this objective by demonstrating controlled

Detailed technical objectives to determine the bestpracticable way to achieve the programme objectives
were adopted in 1992 by the ITER Council and served
as tlre focus for design work for the first six years of theEDA Agreement. In 1998 revised detailed objectives

Physics R&D has been summarized [3] and contributed
to the prediction and the design of the ITER plasma.

The design also addresses demanding technicalwere adopted by reducing the size of device as well as
the capital cost, while maintaining the programmaticchallenges including:oobjectives and by taking account of the progress inphysics and technology R&D. Detailed technicalobjectives adopted by ITER Council in 1998 are

the unprecedented size of the superconducting
magnets and structures,

osummarized as follows:oo1) Plasma Performance:o to achieve extended burn in inductively driven
plasmas with the ratio of fusion power to auxiliary.

power injected into the plasma Q)10 with anhigh neutron flux and high heat flux in the firstwall,
extremely high heat flux in the divertor,
remote handling for maintenance and interventionsin an activated tokamak structure,equipment unique to fusion reactors such as

fuelling and pumping, heating and current drive

inductive burn duration between 300 and 500 s,

o to aim at demonstrating steady-state operation

systems and diagnostics.

Major technology R&D projects therefore havebeen undertaken for the purpose of confirming

using non-inductive current drive with Q>5,o controlled ignition should not be precluded.
2) Engineering Performance and Testing:. demonstrate availability and integration of essential
fusion technologies,o test components for a future reactor,o test tritium breeding module concepts; with a 14

performance and understanding operating margins.
Assessment of results indicates, from an engineering andtechnology point of view, that the design is feasible, thatit can be manufactured to specifications, and that it wiU
be capable of meeting its operating objectives. In138

Tsunematsu T. et al., Status of the ITER Project

Table 1 Main Parameters of ITER
Total Fusion Power500 lvtW (700 l{W)Q - fusion power/additional heating power>10

Averase l4MeV neutron wall loadine>0.5MW/nr

Plasma inductive burn time)400 s

Plasma major radius (R)

6.2m

Plasma minor radius (c)2.0m

Plasma current (1p)15l/'A(17 MA)

Toroidal freld at6.2 m radius (Br)

5.3 T

Note: The machine is capable of plasma current up to 17 MA, with theparameters shown in parentheses, within some limitations on other parameters such as pulse length.

particular, seven large projects, undertaken jointly by

the Parties have demonstrated the feasibility,performance, and maintainability of the main
engineering components and systems of ITER under theintemational collaboration. Some key technologies suchas in the central solenoid and the vacuum vessel weremanufactured and tested by a commingled procedureamong several countries and industries. In addition tothe seven large R&D projects, development ofcomponents for fuelling, pumping, tritium processing,heating/current drive, power supplies, plasma
diagnostics as well as safety related R&D havesignificantly progressed.

4. Safety and Environmental Assessments
A central goal of ITER is to demonstrate the safety
and environmental potential of fusion as an energy
source. A consensus across the participant Parties on

Fig, 1 Cutaway view of the ITER device

safety principles has been attained, based oninternationally recognized safety criteria andradiological limits following ICRP and IAEA
recommendations, and in particular on the concept ofdefense in depth and the As Low As Reasonably

The joint work has yielded a mature design, costestimate and safety analysis. The main parameters ofITER are shown in Table 1. Figure I shows a cutaway
view of the ITER device inside the cryostat. Detailed
designs for major specific components have beendeveloped, coherence of parts or subsystems with thewhole has been achieved, and all major outstanding

Achievable (ALARA) principle. Comprehensive safetyand environmental assessments of the ITER design andoperation have been completed under generic site
assumptions. The conclusion indicates that ITER willmeet the objective of demonstrating the safety anddesign issues are resolved except site-specificenvironmental potential of fusion power.

adaptations, based on the design work and fabrication

experience of R&D components. The essential

5. ITER Design and Performanceengineering features on the tokamak core include:o A segmented center solenoid coil with six coils tocontrol strongly shaped plasmas;

5.1 ITER design
The design was developed to satisfy the technicalobjectives and to include the latest progress in physics
and technologies mainly from the ITER R&D activities.

.

An integrated structural arrangement in which
super conducting magnets (18 toroidal field coils,r39

Tsunematsu T. et al., Status of the ITER Proiectand 9 poloidal field coils) and vacuum vessel arelinked to provide an overall assembly whichsimplifies the equilibration of electromagnetic loads

3in all conditions, relying largely on the robustness

oof strong TF coil cases ; andModular in-vessel components (blanket modules
and divertor cassettes) designed to be readily andsafely maintainable by a practical combination ofo=o-.9g,

remote handling and hands-on techniques.

5ll-The tokamak is contained in a cryostat vessel,
situated in a pit, inside a building (the tokamak building)of about 50 m height. Peripheral equipment such as

1.0
0.7 0.8
0.7
0.8 0.9
0.9 1.0

fueling and pumping, heat transfer, auxiliary heating andremote handling systems are arranged in galleries
around the main pit. The seismic condition depends on

1.1

1.2

1 .3

Confinement lmprovement Factor (HH)
Fig.2 Operation region of ITER, o=l0,20 and 50 for

the site. Ifthe seismic ground peak acceleration is larger/"=15MA. The regions are bounded by the density

than 0.2 G, the tokamak building will be supported byflexible isolators. This seismic isolation minimizes thedesign changes due to different seismic conditions. The
main services required for ITER such as the electricalpower, cooling water, fuel treatment, information flow,
assembly and maintenance facilities, waste treatment,
etc. are distributed in ancillary buildings and otherstructures throughout a site of 25 hectares overall.

limit, beta limit and the power threshold to the Hmode transition.

Pr- n = 2.8 M-r Bl82 70 58 Pt

oooott,nc= Ip/7ta2
and5.2 Plasma performance

B*= BQo)l(IrlaBr) .

The design of the ITER tokamak has flexibility of

For Q=10 with 1p=15MA, the predicted

plasma operations which will allow exploration in alarge operation space of fusion power, plasma density,
plasma beta, pulse length and Q value in various
operation modes and will provide a wide range ofopportunities to study burning plasmas as well as todevelop reactor core plasmas. This paper describes
shortly possible plasma operation modes in ITER tostudy issues of burning plasma physics and fusion

performance includes about l5Vo of the margin in theconfinement. The burn time is 400-500 seconds. As Q
increases, the margin in HH decreases and abott l07o ofthe improvement in the confinement is needed for
Q=50. By increasing the plasma current up to 17MA,
the operation space has about lUVo of the confinement
margin shown in Fig. 3. The burn time of this operation
is limited to be about 100 second due to the capability inthe bum flux and in the cooling system. To prolong thepulse length, the ITER has a capability to install a non-reactor core development.

Figure 2 shows the operation space in the fusion
power vs the confinement capability for Q=19,20 and50. The HH factor denotes the deviation from the fitting
formula (HH=l) to the experimental data. The plasma
confinement time (zj, HH factor, L-H transition power

inductive current drive. Figure 4 shows the operation
points of Q=5 for different loop voltages. As an ultimate
state, a steady state operation in ELMy H-modes is(P1-s), Greenwald density (n6) and normalized beta (f*)

attainable for 100 MW. If the confinement is improvedin e.g. a negative shear configuration and bp can beincreases up to 3.5, the steady state operation with q=19
is available in ITER.
Many design studies of possible or desirable fusion
demonstration reactors and power plants have been done
and these provide requirements for reactor core plasmas

are given as follows [4]:Ta= H n tl{!sa3'zl
a IPB98(y,2)'E.th=0.1441f;.e3 Bot5 p-0.6e no4r Mo.te Rt.e1 e058 K0.18

and a good basis of experimental plans. The globalenergy confinement would not be a major issue, and thewhere the units are s, MA, T, MW, 1020 m-3, amu, mand r" - V/(2n2Ra2)

t40

Tsunematsu T. et al., Status of the ITER Project

lp=17MA, Q=50

Long Pulse/Steady State Operation of ITER
1Q00

=oo
3o6a(Lo4zC31250s 5000s ss800o 600
3on.9400a.9aIIu-200

0.7 0.8 0.9 1.0 1.1 1.2 1.3Confinement lmprovement Factor (Hg)
Fig. 3 The enhanced operation region for O=50 and/o=17 MA. The region has a margin for the con-finement. compared to that in Fig. 2.

0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5
Confinem ent lmprovement Factor (H H)
Fig. 4 The operation points of O=5 for different loop
voltages. As an ultimate state, a steady state op-eration in ELMy H-modes is attainable for 100
MW. lf the confinement is improved in e.g. a
negative shear configuration and B" can be in-enhancement factor (Hr) from the ELMy H-mode
confinement should be around 1. Reducing recirculating power, increasing fusion power density andincreasing reliability are major requirements. Theserequirements irnpose the following plasma parameters ina steady-state reactor (SSR) and in a pulse reactor (PR):-

6. Status after the EDA
Delegations from Canada, the European Union,
Japan and the Russian Federation started formalhigh normalized beta (B*) typically more than 3 innegotiations on the joint implementation of the ITER

SSR and 2:.5 in a PR,

project. Canada presented its bid to host the ITER
project in Clarington facing Ontario Lake. European-high bootstrap current (1ss) fraction typically more-

low divenor heat load typically less than 20Vo ofthan 6O7o in a SSR,
alpha and external heating power, and-creases up to 3.5, the steady state operation with
G10 is available in ITER.

avoidance of disruptions.

These requirenrents are not always necessary to achieve

the required performance in ITER. In order to explore
regimes close to fusion power plants with such as a highnormalized beta and a high bootstrap current fraction inITER, methods to improve parameters are needed especially in the confinement, the plasma density profile andthe plasma beta, compared to the assumed standard
mode in ITER, i.e., the ELMy H-mode. At this moment,
only indicative definite database exists for the steadystate of the improved plasma region. The exploration ofthis region in a burning condition is one of the research
items in ITER and the present design includes the capability in the magnet and control system for accommodat-ing such advanced research.

Union is preparing a proposal of Cadarache site in southFrance. The Russian Federation is preparing a proposalof in-kind contributions. In Japan, a possibility to hostthe project has been intensively studied. Following thework plan accepted by the Delegations, consensus onsite preference, cost sharing and procurement allocation,and understandings on Nominee Director General will

be achieved in 2002. The agreement is expected in
2003. The experimental operation is expected to start in2013. About 20 years operation is planned. Variety ofphysics and engineering researches will be done byvarious scientists and engineers. In order to use ITER
efficiently as a worldwide fusion research facility, theexploitation of ITER by the laboratories and universities
and personnel including graduate students participatingof the Parties is planned as well as by the ITER Legal
Entity staff. It is also important to enhance personnel
mobility between the project and the Parties domestic

Tsunematsu T. et al., Status of the ITER Projectprograms as well as to have remote access to thefacility.

t3l ITER Physics Basis Editors, ITER Physics Expert
Group Chairs and Co-Chairs, ITER Joint Central
Team and Physics Integration Unit, ITER EDA,

References

Nucl. Fusion 39,2137 (1999).
l4l Shimada, M. et al., "Physics Basis of ITERFEAT", Fusion Energy 2000 (CD-ROM, IAEA,Vienna), ITER p.105 (2001), IAEA-CSP-8/C, ISSN[]

"Summary of the ITER Final Design Report", to be
published by IAEA in the ITER EDA Documenta-tion Series, IAEA, Vienna (2001). http:ll
www.iter. org/ITERPublic/ITER/[2] ITER Joint Central Team and Home Teams, FusionEng. Des.55,97 (2001) Issues 2-3.

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