Nuclear Fusion
Reactors//safety and radiation protection
considerations for demonstration
reactors that follow the ITER facility
IRSN Report 2017/199
ISRN/IRSN-2017/199 – FR+ENG

1© IRSN/2018 – All rights reserved

L’IRSN//in brief

The Institute for Radiological Protection and Nuclear Safety is a
public institution with industrial activities. Its missions are definedin French Act no. 2015-992 of 17 August 2015 pertaining to theenergy transition for green growth. Its structure and governance aredefined in Decree no. 2016-283 of 10 March 2016. IRSN operates
under the joint authority of the Ministry of the Environment, theMinistry of Defence and the Ministries in charge of Energy, Researchand Health.
It is the nation’s public service expert in nuclear and radiation risks,
and its activities cover all the scientific and technical issues related
to these risks. IRSN interacts with all parties concerned by theserisks to contribute to public policy issues relating to nuclear safety,
human and environmental protection against ionizing radiation,and the protection of nuclear materials, facilities, and transport
against the risk of malicious acts. Its work also actively contributes
to other major public policies in research, innovation, occupational
health and environmental health.

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Foreword
The Institute for Radiological Protection and Nuclear Safety
is a public service expert in nuclear and radiation risks.
It develops research programmes and conducts studies in its fields
of scientific and technical expertise, and provides technical support
to the public authorities in charge of nuclear safety and security,
and radiation protection. In fulfilling its missions in risk assessment
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In its desire to share its knowledge and provide high-quality
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IRSN publishes “reference documents” that reflect its position on agiven subject on the day they are published.
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Jean-Christophe NIEL
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Preface
The goal of this document, which falls under the field of forward
planning, is to explain the safety and radiation protection issues
which need to be examined while designing future nuclear fusion
reactors after ITER. Such reactors constitute the preparatory stage
before building industrial nuclear fusion power generation facilities.
These reactors will follow on the heels of ITER (International
Thermonuclear Experimental Reactor), which is currently being built
at Cadarache (France) to demonstrate the technical and scientificfeasibility of controlled fusion. They are currently being studied invarious countries across the world (China, South Korea, India, etc.).
ITER is the first magnetic confinement fusion reactor to require
a construction licence decree under French nuclear facilities
regulations. The decree was passed in 2012 after the Institutefor Radiological Protection and Nuclear Safety conducted an indepth assessment of the safety and radiation protection measures
adopted by the operator.
Due to its design and operation, ITER presents unique safety
and radiation protection issues, including the risk of dust andhydrogen isotope explosion, plasma malfunction, magnetic system
failures, etc.
The Institute has put in place the following measures to manage
these specific features:•• collaboration with the Canadian Nuclear Safety Commission on
tritium confinement and the effects of tritium releases into theenvironment;•• technical support from a plasma physics expert;•• neutronic and structure activation calculations by the European
Organization for Nuclear Research in Geneva (Switzerland) andthe implementation and research and development programmes
within the Institute and/or in collaboration on specific safety
problems raised by a nuclear fusion facility.
In performing the safety analysis for ITER, IRSN has acquired specificexpertise in this field. The Institute drew from this experience andacquired knowledge and considered that it was wise to explore thesafety and radiation protection issues that demonstration reactors
currently under development could raise.

7© IRSN/2018 – All rights reservedFuture demonstration reactors will mainly differ from the ITER
reactor by seeking to attain tritium self-sufficiency and achievesignificantly longer operating times. These differences will havea significant impact on design and a direct influence on safety.
The thoughts expressed in this document highlight the growing
importance of safety and radiation protection issues in these new
designs with respect to the ITER facility.

Giovanni Bruna
Scientific Director

8© IRSN/2018 – All rights reserved

//Author
Didier Perrault							

Nuclear safety division
Senior Expert

9© IRSN/2018 – All rights reserved

Sommaire
1/		Introduction
2/		 The “tokamak” concept
3/		 Design of a nuclear fusion facility
4/		 The ITER experimental fusion facility
5/		 Nuclear fusion reactor projects
5/1 Intermediate experimental reactors
5/1/1 China’s “CFETR” project
5/1/2 America’s “FNSF-AT” project
5/1/3 India’s “SST-2” project
5/2 Fusion power reactor projects
5/2/1 South Korea’s “K-DEMO” project
5/2/2 Europe’s “DEMO” project
5/2/3 Japan’s project
6/		 The main differences between the planned reactors
		 and the ITER experimental facility
6/1 Self-sufficiency in tritium
6/2 Significant operating times
7/		 Safety and radiation protection issues to be examined		 from the design-phase of DEMO reactors
7/1 Residual heat removal
7/1/1 During operation without plasma
7/1/2 During sector transfer, and storage and maintenance				 in hot cells
7/2 Ionising radiation exposure risks
7/3 Types of accidents to consider
7/3/1 Changes in the quantities of tritium and dust				 in the vacuum vessel
7/3/2 Presence of tritium breeding blankets
7/3/3 Increase in the number of possible cases				 of loss of control of the plasma
7/3/4 Increased magnetic energy of the toroidal field coils
7/3/5 Increase in the quantities of helium used

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© IRSN/2018 – All rights reserved

7/3/6 Increase in the number of rooms where there could be				 significant quantities of hydrogen isotopes outside				 the vacuum vessel
7/3/7 Vertical port in the vacuum vessel
7/3/8 Specification of protection				 with regard to extreme events
7/4 Releases into the environment under normal operation
7/4/1 Reducing tritium quantities in the facility
7/4/2 Examination of the main possible paths for gaseous				 tritium releases
7/4/2/1 Releases associated with the cooling systems for tritium								 breeding blankets
7/4/2/2 Releases associated with transfers of internal components								 to the hot cells and their processing in these cells
7/4/2/3 Releases associated with waste detritiation equipment

7/5 Waste
8/		Conclusion
9/		Glossary
10/ References
11/ Summary of the IRSN report on the construction
		 license application for the ITER experimental facility
12/ Main past and scheduled milestones for the ITER facility

12© IRSN/2018 – All rights reserved

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Figures
Figure 1.

Principle of magnetic confinement of a plasma
in a tokamak © Georges Goué/IRSN.

17

Schematic diagram of the ITER facility tokamak.
© ITER Organization.

18

ITER facility vacuum vessel andinternal components. © ITER Organization.

19

Various types of tritium breeding blankets,from [1]. © Georges Goué/IRSN.

21

ITER facility vacuum vessel pressure suppression
system (VVPSS). © Georges Goué/IRSN.

22

Figure 6.

ITER facility cryopump. © ITER Organization.

23

Figure 7.

ITER facility cryostat. © ITER Organization.

24

Figure 8.

ITER facility tokamak building.
© ITER Organization.

27

ITER facility tritium building.
© ITER Organization.

28

Figure 10. ITER facility hot cell building.
© ITER Organization.

29

Figure 11. ITER facility automated transfer cask [2].
© 2017 Elsevier B.V.

30

Figure 12. Design of the Chinese CFETR reactor,
from [3] and [4]. © DR.

36

Figure 13. Design of the American FNST-AT reactor
project [5]. © A. M .A. Garafalo/General Atomics.

37

Figure 14. Republic of India Roadmap, from [6]. © DR.

38

Figure 15. Design of the South Korean K-DEMO reactor,
from [7]. © DR.

39

Figure 16. Design of the European DEMO reactor [8].

40

Figure 17. Cooling for the blanket sectors
on the Japanese SlimCS reactor project [18].
© Kenji Tobita/JAEA.

50

Figure 2.
Figure 3.
Figure 4.
Figure 5.

Figure 9.

13© IRSN/2018 – All rights reserved

14© IRSN/2018 – All rights reserved

Figure 18. Fuelling system bypass, from [33].
© Georges Goué/IRSN.

63

Figure 19. Releases associated with the cooling systems
for tritium breeding blankets, from [36].
© Georges Goué/IRSN.

64

1/ Introduction

1/

Introduction
Current power reactors use fission of heavy nuclei, mainly uranium,
to produce energy. However, fusion projects are based on fusion of
the light nuclei of hydrogen isotopes, deuterium and tritium.
Since initial research on nuclear fusion began in the 1950s, thesystem most frequently used to produce it is the “tokamak”,
comprising a toroidal chamber in which a gaseous mix of hydrogenisotopes in the form of plasma is confined using the helical magnetic
field resulting from the combination of magnetic fields producedby the field coils. Other projects aim to produce a helical magnetic
field by giving the toroidal chamber and coils a helical form, as
in the W7X stellerator developed in Germany. Another approachhas been developed, which consists of using lasers to exert highpressure on a target of hydrogen isotopes, mainly in the context of
nuclear weapons development (the CEA’s(1) Megajoule Laser facility[LMJ], the National Ignition Facility [NIF] developed by Lawrence
Livermore National Laboratory, etc.). This report only covers themost common approach, magnetic confinement in a tokamak.
Since the 1950s, over 200 tokamaks have been developed aroundthe world to perform research on fusion. In the late 1980s, adecision was made by the European Atomic Energy Community(Euratom), Japan, the Soviet Union and the United States toconstruct ITER (International Thermonuclear Experimental Reactor)
with a view to “the scientific and technical feasibility of fusion
power”. This decision is part of a long-term initiative which thenplanned construction of a second research reactor, called DEMO
(DEMOnstration power plant), more similar to a power reactor,
prior to industrialisation of nuclear fusion.

Fusion is the combinationof two nuclei of light atoms,
here deuterium and tritium,
to form a heavier nucleus
which releases a large
amount of energy carried by
the reaction products (nuclei,
particles and radiation).
The most commonly usedsystem for producing a fusion
reaction is the tokamak, aRussian acronym meaning“toroidal chamber withmagnetic coils”, which uses
magnetic fields to create,confine and control a hot
plasma inside which thefusion reaction can occur.
(1)
Commissariat à l’énergie
atomique et aux énergies
alternatives.

15© IRSN/2018 – All rights reserved

ITER is the first fusion facility
using magnetic confinement
requiring authorisationin accordance with theregulations applicable tonuclear installations.
(2)
Institut de radioprotection etde sûreté nucléaire.

Various reactor projects,
called “DEMO reactors”,
using the tokamak concept
are under development
worldwide. IRSN is using
the experience acquired
during the ITER facility safetyassessment to identify thesafety issues they need totake into account from thedesign phase.

16© IRSN/2018 – All rights reserved

Due to the quantities of radioactive substances used, far greaterthan those used in the tokamaks previously built, the ITER
facility is the first fusion facility using magnetic confinement
to require a construction license under French law governing
nuclear facilities. Consequently, its construction license wasissued via Decree no. 2012-1248 of 9 November 2012, following
a detailed examination of the provisions adopted by the operator
to prevent or adequately mitigate the risks and disadvantages it
presents. Assessment of these provisions was performed by the
French Institute for Radiological Protection and Nuclear Safety(IRSN(2)). A summary of this assessment is given in Appendix 1 ofthis document. The ITER facility is currently under construction.
A timeline of the main past and scheduled milestones is given in
Appendix 2 of this document.
Currently, no new international projects are planned following
the ITER facility. Different nuclear fusion reactor projects are
currently being studied in various countries around the world(People’s Republic of China, Republic of Korea, Republic of India,etc.). Most of these projects are based on the same “tokamak”
concept. For simplicity, in the rest of this document, all these“tokamak” projects will be called “DEMO reactors”. Thesefacilities involve more significant safety and radiation protection
issues than for the ITER facility.
The purpose of this document is to present some considerations
on the safety and radiation protection issues that should be
examined from the design stage of “DEMO reactors”, based on
publications currently available and experience acquired during
the ITER facility safety assessment.

2/ The “tokamak” concept

2/The “tokamak”
concept
In a nuclear fusion facility using the “tokamak” magnetic
confinement concept, fusion reactions take place inside a toroidal
plasma, a state of matter made up of ions and electrons. Magnetic
confinement of the plasma (Figure 1) is mainly produced by atoroidal magnetic field (Bt) produced by toroidal field coils and by
a poloidal magnetic field (Bp) produced by an electric current in
the plasma. Helical magnetic field lines are thus produced aroundwhich ions and electrons spiral, thereby containing the plasma
within a sealed toroidal vessel, called the “vacuum vessel”, into
which deuterium and tritium have been introduced.

Plasma is a hot, low-density
gas produced under theaction of a strong magnetic
field. It is made up of positive
ions and electrons, whichhave been stripped off due tothe temperature.

Figure 1. Principle of magnetic confinement of a plasma in a tokamak.
© Georges Goué/IRSN.

17© IRSN/2018 – All rights reserved

The induced current, called the “plasma current”, heats the plasma
but is not in itself enough to reach the temperature conditions
needed for the fusion of deuterium and tritium. Additional
heating is therefore needed, which will be described later.
The fusion reaction of deuterium () and tritium (), whichproduces a particles () and neutrons (n) is as follows:

Tritium is the radioactive isotope of hydrogen. Its nucleus decays
via a beta emission with an average energy of 5.7 keV to produce
stable helium-3. Its half-life is 12.3 years.

High energy neutrons areproduced during fusion in theplasma. 80% of this energy isconverted into heat.

The fusion reaction produces high-energy neutrons (14 MeV)
whereas the energy of neutrons produced by fission reactions isless than 2 MeV. A fifth of the energy is released into the plasma
via the α particles, while four-fifths is converted into heat via
neutron interactions with the materials comprising vacuum
vessel internal components and the vessel itself.
A “tokamak” (Figure 2) is made up of the following main
components:

Figure 2. Schematic diagram© ITER Organization.

18© IRSN/2018 – All rights reserved

of

the

ITER

facility

tokamak.

2/ The “tokamak” concept•• the vacuum vessel and its internal components (blanket anddivertor);
The vacuum vessel is a sealed toroidal metal chamber which
encloses two main components: the blanket and the divertor(Figure 3).

Figure 3. ITER facility© ITER Organization.

vacuumvesselandinternal

components.

The blanket provides neutron protection for the metal wallsof the vacuum vessel. It is cooled via a coolant to transfer theradiative heat of the plasma and a large fraction of the heat due
to the slowing of neutrons in the structure to cooling towers (forthe ITER facility) or to a turbine generator (for “DEMO reactors”).
It also removes the heat produced by nuclear reactions in its
materials caused by neutrons from the fusion reactions in theplasma (“neutron reactions”). Furthermore, very sparsely for the
ITER facility and comprehensively for the “DEMO reactors”, theblanket is tritium breeding, i.e. it contains lithium (Li) which,under the neutron flux from the plasma, produces tritium via
the following reactions:

Most of the heat removed
from ITER, which will
ultimately be converted into
electricity in a DEMO reactor,
comes from the slowing of
neutrons in the structuresof the vacuum vessel and its
internal components.

19© IRSN/2018 – All rights reserved

(3)Tungsten is a highly
refractory metal which offers
very high resistance to heat
and wear.

Tritium breeding blankets
promote tritium production
while also providing neutron
protection for the metal
walls of the vacuum vessel.

The wall of the blanket that faces the plasma, called the firstwall, is subject to heavy loads due to the neutron flux, thermal
flux and thermal shocks, which erode it and create dust(see Chapter 3). Furthermore, this first wall adsorbs tritium.
It is made of beryllium or tungsten(3), and must be changed
regularly during operation. To do this, various sized parts ofthe blanket must be removed from the vacuum vessel using
robots, then transported, also by robotic means, into hot cells
in which the first wall is changed.
The tritium breeding blankets (see Section 6.1) are essential
equipment for future nuclear fusion facilities. They are thesubject of intense research and development activities
worldwide. Several types of blanket are being studied (someexamples are given in Figure 4). They always include thefollowing:−−lithium which, under neutron flux from the plasma, reacts
to form tritium. The lithium may be present under various
forms (liquid lithium, lithium-lead eutectic, lithium-basedoxide ceramics, molten lithium salts, etc.);−−a neutron booster material, such as beryllium or lead, which
can increase tritium production. Interactions between
neutrons and these materials cause nuclear reactions,called (n, 2n) reactions, which lead to the emission of twice
as many neutrons;−−a coolant (water, helium, liquid metal, etc.) to transfer partof the radiative heat of the plasma, heat due to the slowingof neutrons in the structure and heat from “neutron
reactions” to cooling towers (for the ITER facility) or to aturbine generator (for “DEMO reactors”).

The divertor, located in thelower part of the vacuum
vessel, extracts helium, fuelthat has not undergone
fusion (and can be reused)
and impurities.

20© IRSN/2018 – All rights reserved

The divertor, located in the lower part of the vacuum vessel,
extracts helium-4 (α particles from the fusion reactions thathave captured plasma electrons), fuel (tritium and deuterium)that has not undergone fusion and impurities (in particular
the dust from erosion of the first wall of the blanket and thedivertor). As for the blanket, the divertor is cooled by a coolant
to transfer part of the radiative heat of the plasma, heat due
to the slowing of neutrons in the structure and heat from“neutron reactions” to cooling towers (for the ITER facility) or
to a turbine generator (for “DEMO reactors”). The divertor also
has a first wall facing the plasma, whose material (tungsten)
may different from that of the first wall of the blanket, due

2/ The “tokamak” concept
to the greater thermal flux from the plasma. This first wall mustalso be changed during operation, in the same way as for theblanket.

Figure 4. Various types of tritium breeding blankets, from [1]. TES: tritium
extraction system; CPS: coolant purification system; PCS: power conversion
system). © Georges Goué/IRSN.

The vacuum vessel is a metal chamber. Cooling water flows
between the two walls to transfer the heat due to the slowingof neutrons in the walls and in the borated stainless steel blocks
installed between these walls, along with heat from “neutron
reactions”, to cooling towers (for the ITER facility) or to a turbine
generator (for “DEMO reactors”). Penetrations are used to installequipment for monitoring plasma behaviour and performance,
for additional heating, for the fuelling system and for cooling andvacuum systems (see Figure 3). Penetrations are also plannedfor extracting parts of the blanket or divertor for maintenancepurposes. These penetrations can have large cross-sections when
the parts of the blanket or divertor to be extracted are verylarge, which has a significant impact on the overall design of thetokamak.

21© IRSN/2018 – All rights reserved

The vacuum vessel has a pressure-suppression system to copewith all foreseeable accident situations involving a rise ininternal pressure (water ingress, air ingress, etc.). This system
is made up of a relief line fitted with valves and rupture discs,
leading to a relief tank half filled with water, which uses
sparging to condense most of the tritiated water (HTO) formed
via reaction with the tritium in the event of a cooling waterleak into the vacuum vessel. The non-condensed fraction of
the tritiated water would mostly be trapped by the ventilationand detritiation system which keeps the atmosphere of therelief tank under negative pressure. The fraction that is nottrapped would be released into the environment via a facility
stack (Figure 5). In the event of air ingress into the vacuum
vessel, the non-condensable tritiated gases (HT) would mostly
be trapped by the ventilation and detritiation system alone.
Other radioactive materials present in the vacuum vessel
and cooling systems (activated dust and activated corrosion
products) would mostly be trapped by the ventilation anddetritiation system filters.

Figure 5. ITER facility vacuum vessel pressure suppression system(VVPSS). © Georges Goué/IRSN.

22© IRSN/2018 – All rights reserved

2/ The “tokamak” concept•• the cryopumps;
The cryopumps (Figure 6), located under the vacuum vessel
downstream of the divertor, remove gases and dust from thevacuum vessel (mainly tritium and deuterium that have notreacted, helium-4 from fusion reactions and dust from erosionof the first wall of the blanket and the first wall of the divertor)
and transfer them to an auxiliary building.

Figure 6. ITER facility cryopump. © ITER Organization.

•• the magnetic system;
The magnetic system is made up of superconducting field coils
cooled by circulation of liquid helium at a temperature of 4.5 K.
It is mainly made up of toroidal field coils, poloidal field coils andthe central solenoid, which produce the magnetic field needed
to confine the plasma within the vacuum vessel.
•• the cryostat;
All magnetic field coils are enclosed in a cylindrical stainlesssteel vacuum enclosure, the “cryostat” (Figure 7), whose mainfunction is to maintain the extremely low temperature conditions
needed for superconducting coils to operate. The cryostat itself
is entirely enclosed in a biological shield housing which limits
worker exposure to ionising radiation from the vacuum vessel.

23© IRSN/2018 – All rights reserved

Figure 7. ITER facility cryostat. Approximately 30 m in height anddiameter, with a total mass of 3,800 tonnes of steel, it has 280 penetrations
to provide access for pipework and electrical power and for heating,diagnostics and remote handling systems. © ITER Organization.

•• equipment for
performance;

monitoringplasmabehaviourandThe main equipment for monitoring plasma behaviour andperformance in the vacuum vessel is as follows:
The vacuum vessel issurrounded by a large
number of auxiliary
systems – equipment forcryogenics, measurement
and monitoring,
additional heating, fuelsupply, cooling, vacuum
and electrical power –which work together tocreate and maintain theconditions required forfusion.

24© IRSN/2018 – All rights reserved−−magnetic measurement systems for monitoring plasma
shape and position,−−optical systems for monitoring plasma temperature anddensity profiles,−−spectroscopic devices and particle analysers to determine
the characteristics of various particles (ions, electrons,
α particles, impurities, etc.),−−neutron measurements to calculate the quantity of fusion
power produced.

2/ The “tokamak” concept•• equipment for additional heating of the plasma;
Additional heating of the plasma is provided by various
components using two different heating methods: injectionof a beam of high-energy neutral atoms and emission ofelectromagnetic waves.
•• fuel supply lines (for deuterium and tritium) ;The deuterium and tritium introduced into the vacuum vessel
come from the tritium building, which will be described later inthis document.
•• cooling systems;
Radiative heat from the plasma, heat due to the slowing of
neutrons and heat produced by neutron reactions in the vacuum
vessel and its internal components is removed by three primarycooling systems: one for the vacuum vessel, one for the blanket
and one for the divertor. These primary cooling systems arethemselves cooled via heat exchangers by a secondary coolingsystem, which transfers its heat to a third system, made up
of cooling towers for facilities like ITER which do not produce
electricity, or feeding a turbine generator for “DEMO powerreactors”.
•• disruption mitigation systems;
A plasma is subject to multiple types of instability. Small-scaleinstabilities lead to turbulence similar to that observed in a fluid.
This has the effect of mixing the hot particles (ions and electrons)from the centre of the plasma with the colder ones nearer to theedges, but does not destabilise the plasma enough to make itlose its magnetic confinement. However, large-scale instabilities
affect the plasma as a whole. These include oscillation, wave
propagations and vertical displacements up or down. If theplasma touches the first wall of the blanket or the first wall of
the divertor, it completely loses its magnetic confinement ina few milliseconds. This sudden plasma termination is called“disruption”.

Disruption is the namegiven to loss of plasma
confinement in the vacuum
vessel. Systems to mitigate
this are being studied.

Under the current operating conditions of a tokamak, disruptions
are frequent, which is one of the reasons why operating sequenceswith plasma, called “plasma discharges”, remain very short (usuallyaround a few tens of seconds, although the Tore-supra tokamak
achieved a duration of around 7 minutes in 2003).
Given the large amount of energy stored in the plasma, when adisruption occurs it causes physical phenomena (thermal shocks,

25© IRSN/2018 – All rights reservedIn summary:
ITER will be the largest
facility in the world using atokamak and is designed todemonstrate the scientificand technical feasibility offusion power. The hot plasma
in which the nuclear fusion
reaction occurs is formed
by a strong magnetic field.
The fusion reaction between
deuterium and tritium atoms
is the one that produces themost energy at the “lowest”
temperatures. The plasma will
be confined in the vacuum
vessel. The blanket protects
the vacuum vessel and thesuperconducting magnets
from the energy carried by
the neutrons produced. For
ITER, this energy will be
removed but ultimately, fora DEMO reactor, it will be
converted into electricity.
The divertor located in thevacuum vessel removes
fusion reaction products. Allof the vacuum vessel internal
components are subject
to very large energy fluxes
which require penetrations
to allow their removal for
maintenance. Auxiliary
systems located aroundthe central vessel create
the conditions required forfusion. A major future issue isthe development of tritium
breeding blankets to provide
reactors with self-sufficiency
in tritium.

26© IRSN/2018 – All rights reserved

electron beams, eddy currents, etc.) which have definite
consequences (electromagnetic loads in the vacuum vesselinternal components and the vessel itself, production of dust
from erosion of the first wall of the blanket) or lead to a risk ofwater leaks.
To prevent the occurrence of disruptions as far as possible,
systems are being studied to shut down the “plasma discharge”(for example by a large injection of gas) without a disruption.
Such systems could be added to the ITER tokamak and will most
likely be planned from the design phase for future nuclear fusion
facilities.

3/ Design of a nuclear fusion
facility

3/

Design of a nuclear
fusion facility
In a nuclear fusion facility, the tokamak, located in the main
“tokamak building”, is surrounded by rooms that mainly house theprimary cooling systems (Figure 8) and the vacuum vessel relief
tank.

Figure 8. ITER facility tokamak building. © ITER Organization.

27© IRSN/2018 – All rights reserved

The vacuum vessel fuelling lines (deuterium and tritium) and the
cryopumps that remove the fusion reaction products, are locatedin the “tokamak building”, but are only part of the fuelling system.
This system recycles any unburnt fuel present in the products
removed from this vessel, along with any produced in the tritium
breeding blankets back to the vacuum vessel. Other fuellingsystem equipment is located in an auxiliary building, called the“tritium building” (Figure 9). This equipment (see Figure 18):•• processes products extracted from the vacuum vessel andtritium breeding blankets, i.e. separates hydrogen isotopes(hydrogen, deuterium, tritium) from other gases (He, Ar, etc.)
and impurities (dust, etc.);
The buildings around thevacuum vessel provide
cooling for the components
located around the vessel,and extract the reaction
products for removal as
waste or for recycling backinto the vessel.

•• performs hydrogen isotope separation. The isotope separation
system comprises cryogenic distillation columns, which
operate based on the different boiling points of the various
isotopes. The separated hydrogen isotopes or mixtures ofthese isotopes are then stored. Fuel is stored in metal hydride
beds (zirconium, titanium, uranium etc.);•• provides fuelling for the vacuum vessel.

Figure 9. ITER facility tritium building. © ITER Organization.

As seen above, vacuum vessel internal components require
periodic maintenance. This maintenance is performed in the28© IRSN/2018 – All rights reserved

3/ Design of a nuclear fusion
facilitycells located in an auxiliary building, called the “hot cell building”(Figure 10). This building may also be used to process and store
intermediate level long-lived waste (ILW-LL) and tritiated waste.

Figure 10. ITER facility hot cell building. © ITER Organization.

It should be noted that the dose rates resulting from activationof vacuum vessel internal components are such that parts of thisequipment can only be removed by robotic means and that theirtransfer from the vacuum vessel to the cells can only be performedby automated transfer casks (Figure 11).
Other waste, whether low and intermediate level short-lived waste(LILW-SL) or very low level waste (VLLW), is processed and stored inanother auxiliary building, called the “radwaste building”.
All buildings are fitted with ventilation systems which, along withthe static confinement provided by the process equipment, rooms
and buildings, provide confinement of radioactive substances. The
ventilation systems of a large number of rooms and equipment
items (glove boxes, etc.) include equipment for air detritiation
under normal and accident operating conditions.

Due to high dose rates,
robotic means are used totransfer equipment requiring
periodic maintenance.
All nuclear buildings(tokamak, tritium, radwaste,
hot cell) have ventilation
systems that provide
confinement of radioactive
substances.

29© IRSN/2018 – All rights reserved

Figure 11. ITER facility automated transfer cask [2]. © 2017 Elsevier B.V.

Two types of radioactive substance are found in a nuclear fusion
facility:•• tritium;
Tritium is most often found as a gas (T2), which easily diffuses
across seals and certain materials, or as tritiated water (HTO),
which is often corrosive for the equipment involved.
The tritium needed for fusion reactions will come from outside
the ITER facility, and will be produced in tritium breedingblankets for DEMO reactors.
Most of the tritium is located in the “tritium building”(several kg) but a significant quantity is also found in thevacuum vessel (in the first wall of the blanket and in thelithium-bearing part of the blanket), and to a lesser extent,in the hot cell building due to the maintenance operations
performed on vacuum vessel internal components and waste
storage.
Without intervention, the quantities of tritium adsorbed onto
the first wall of the vacuum vessel internal components could
become very large. The operators of nuclear fusion facilities
limit the quantities of tritium adsorbed in order to avoid losing
too much fuel and to avoid excessive consequences in theevent of accidents involving tritium. To reduce the quantitiesof tritium adsorbed, they use thermal desorption to remove
the tritium from the first wall of the vacuum vessel internal

30© IRSN/2018 – All rights reserved

3/ Design of a nuclear fusion
facility

components, by operating the primary cooling systems at atemperature much higher than under normal operation.
•• activated materials or products;
Interactions between neutrons from fusion reactions in theplasma and the nearby environment lead to the production of
the following:−−activated structural materials,
The blanket and divertor are highly activated, as are to alesser extent the vacuum vessel, the field coils, the cryostat
and the equipment for monitoring plasma behaviour andperformance, for additional plasma heating, for the fuellingsystem, and for the cooling and vacuum systems. If theblanket is tritium breeding, it will also contain activatedfluids, gases and solids, which vary depending on the type oftritium breeding blanket used.
−−activated dust,
The dust resulting from erosion of the first wall of the vacuum
vessel blanket and the first wall of the divertor is activated;
it also contains adsorbed tritium. Without intervention, thequantities of activated dust in the vacuum vessel could
become very large. As with tritium above, operators of nuclear
fusion facilities limit the quantities of dust as it disturbs theplasma. In addition, the consequences of accidents that could
involve this dust must not be too serious. For these reasons,dust is regularly removed from the vacuum vessel using
robotic vacuum cleaners. However, this results in significant
storage of activated dust in the hot cell building.

Two types of radioactive
substance are produced
during fusion: tritium which
will be produced in DEMO
reactors and is introduced in
ITER, and activated materials
or products. As with tritium,
activated dust must be
periodically removed, inparticular to mitigate theconsequences in the event ofan accident.

−−water activation products,
The activation products of the water in the primary coolingsystems are tritium, carbon-14, nitrogen-16 and nitrogen-17.
However, tritium production by activation of the waterin primary cooling system is negligible compared with thequantities of tritium which diffuse from the vacuum vessel
into the cooling systems via their walls.
−−activated corrosion products in water,
Corrosion products formed in the primary cooling systems
which could be mobilised in the event of a water leak are ionsin solution and non-fixed deposits on the walls, which areactivated during passage through the vacuum vessel internal
components and the vessel itself.

31© IRSN/2018 – All rights reservedIn summary:The buildings around thevacuum vessel provide
cooling for the components
located around the vessel,and extract the reaction
products for removal as
waste or for recycling backinto the vessel.
Due to high dose rates,
robotic means are used totransfer equipment requiring
periodic maintenance.
All nuclear buildings(tokamak, tritium, radwaste,
hot cell) have ventilation
systems that provide
confinement of radioactive
substances.
Two types of radioactive
substance are produced
during fusion: tritium which
will be produced in DEMO
reactors and is introduced in
ITER, and activated materials
or products. Activateddust and tritium must be
periodically removed, inparticular to mitigate theconsequences in the event ofan accident.

32© IRSN/2018 – All rights reserved−−activated gases,
The air present between the cryostat and the biologicalshield housing, and to a lesser extent, the air in the tokamakbuilding rooms, is activated by the neutron flux produced by
the fusion reactions. The main isotopes entailing irradiation
risks are carbon-14 and argon-41.
Inert gases (nitrogen, neon, etc.) which are introducedin small quantities between the plasma and the divertor
to reduce radiative heat transfer to the first wall of thedivertor are also activated.

4/ The ITER experimental
fusion facility

4/

The ITER experimental
fusion facility
The general goal of the ITER facility is to “demonstrate the scientificand technological feasibility of fusion power by producing around
500 MW for pulses of several hundreds of seconds”.
During the 20 years of operation planned for this facility, around
40,000 plasma discharges could be performed.
Other than the fusion power targeted, the main characteristics ofthe ITER facility tokamak are as follows:•• a vacuum vessel with a volume of approximately 1,600 m3;•• a nominal plasma current of 15 MA;•• a beryllium first wall for the blanket;•• a tungsten first wall for the divertor;•• six tritium breeding modules out of the 440 modules of theblanket; the six places reserved for these modules mean that theseven members of the international ITER Organization for fusionenergy (the People’s Republic of China, the European Atomic
Energy Community, the Republic of India, Japan, the Republic ofKorea, the Russian Federation and the United States of America)
can test the various types of tritium breeding blankets they aredeveloping.

33© IRSN/2018 – All rights reservedIn summary:
The general objectiveof the ITER facility is to“demonstrate the scientificand technological feasibilityof fusion power by producing
around 500 MW for pulses ofseveral hundreds of seconds”.
Almost 40,000 plasma
discharges are planned,
constrained by the limited
quantities of tritium
authorised in the facility.
This facility offers thepossibility of studying theperformance of the auxiliary
systems (heating, remotehandling, cryogenics, etc.)
and performing tritium
production experiments via
the six places reserved fortritium breeding modules
among the 440 blanket
modules.

34© IRSN/2018 – All rights reserved

The authorised operating domain for the ITER facility is limitedby:•• a quantity of tritium in the whole facility less than or equal
to 4 kg,•• a quantity of tritium in the vacuum vessel less than or equal
to 1 kg,•• a mass of dust in the vacuum vessel less than or equal to1,000 kg (tungsten and beryllium).

5/ Nuclear fusion reactor
projects

5/

Nuclear fusion reactor
projects
Different countries have taken different approaches with regard to
how to follow on from the ITER experimental facility with a view tonuclear fusion power generation. However, two general strategies
emerge:•• before building a fusion reactor for power generation, somecountries are aiming to demonstrate its feasibility using
smaller experimental reactors (fusion power below 1,000 MW);
this is the case for the People’s Republic of China, the UnitedStates of America and the Republic of India. In the remainderof this document, these projects are collectively referred to as“intermediate experimental reactors”;•• other countries are aiming to directly build an experimental
fusion reactor for power generation, with a fusion power output
over 1,000 MW; this is the case for the Republic of Korea, theEuropean Atomic Energy Community and Japan. In the remainderof this document, these projects are collectively referred to as“fusion power reactors”;

Internationally, countries
have opted either to directlybuild fusion power reactors
or to first build experimental
reactors.

To date, the Russian Federation has no known plans for a tokamak
nuclear fusion facility.
5/1

Intermediate experimental reactors
5/1/1

China’s “CFETR” project
The People’s Republic of China considers that before building afusion power reactor, it is necessary to:35© IRSN/2018 – All rights reserved

The Chinese project aims
to study the behaviour of
the tritium breeding blanket
and the resistance of thematerials, and to optimise
the design of the divertor,
with a fusion power output
of up to 1,000 MW.
Maintenance via middleplane, upper and lower
penetrations has beenretained.

•• demonstrate the real effectiveness of the implementation ofa tritium breeding blanket (self-sufficiency in tritium, tritium
processing, blanket cooling);•• demonstrate the resistance of the materials used under highneutron flux;•• improve the design of the divertor.
In view of this, the People’s Republic of China plans to build afacility called the China Fusion Engineering Test Reactor (CFETR).
Operation will start with a fusion power output of 200 MW, rising
to 1,000 MW in the second phase of CFETR facility operation.
This project has been the subject of quite a large number ofpublications.
The People’s Republic of China has not yet determined the typeof tritium breeding blankets to be used on the CFETR facility. They
have been studying several types of tritium breeding blanket.
The tokamak configuration adopted is one with middle-plane,
lower and upper penetrations of the vacuum vessel, used forextracting very large sectors of the blanket and divertor for
maintenance (Figure 12).

Figure 12. Design of the Chinese CFETR reactor, from [3] and [4]. © DR.
5/1/2

America’s “FNSF-AT” project
To date, there has been no US government decision to build a
DEMO reactor. However, a strategy has been proposed by the
Department Of Energy (DOE). This suggests that, before deciding
to build an experimental fusion power reactor, a facility called
the Fusion Nuclear Science Facility Advanced Tokamak (FNSF-AT)
should be built, to demonstrate that:36© IRSN/2018 – All rights reserved

5/ Nuclear fusion reactor
projects•• nuclear fusion facilities can be self-sufficient in tritium while
producing electrical power;•• the blanket and divertor can be used for a relatively long time
without needing maintenance;•• the physical phenomena involved are well understood (damage
caused by neutron interactions with materials, tritium behaviourin the first wall of the blanket and divertor, etc.).
Even if design studies for the FNSF-AT were launched in the next
few years, the facility would not be operational before 2030. An
experimental fusion power reactor is therefore not conceivable
before 2050.
To be adaptable based on developing knowledge, the FNSF-AT
is to be modular, in particular thanks to implementation ofvertical penetrations for the extraction of vacuum vessel internal
components (vertical maintenance, Figure 13). The fusion power
output of 125 MW at commissioning could be raised to 250 MWthen 400 MW. The target is for a mean operating time with plasmaof approximately 30% per year.

The American project aims
to develop a system that is
self-sufficient in tritium bystudying two types of tritium
breeding blanket which also
allow for electrical power
generation, to optimise theservice life of vacuum vessel
components and to study
the physical phenomena
involved, for fusion power
output of up to 400 MW.
Maintenance via vertical
penetrations is preferred.
The quantity of tritium used
would be approximately 4 kg
for an operating time withplasma of approximately
30% per year.

Figure 13. Design of the American FNST-AT reactor project [5].
© A. M .A. Garafalo/General Atomics.

37© IRSN/2018 – All rights reserved

Two types of tritium breeding blanket are under consideration:•• Helium Cooled Ceramic Breeder (HCCB ), using a lithiumbased oxide ceramic for tritium production, beryllium as aneutron booster and helium to cool the structures;•• Dual Coolant Lithium Lead (DCLL ) using a LiPb eutectic fortritium production, as a neutron booster and for cooling thestructures, along with helium for cooling the structures. The
DCLL blanket seems to be the preferred option.
The quantity of tritium present in the whole facility would bearound 4 kg.
5/1/3

India’s “SST-2” project

India’s project aims to study
various types of tritium
breeding blanket and various
components for a fusion
power output of up to
500 MW.

The Republic of India’s programme envisages the constructionof a facility called the Steady State Superconducting Tokamak-2(SST- 2) by 2027, and a fusion power reactor with fusion power
output of 3,300 MW by 2037 (Figure 14). The main purposeof the SST-2 facility would be to test the equipment for the
future power reactor, in particular the tritium breeding blankets.
Operation will start with a fusion power output of 100 MW,
rising to 500 MW in later SST-2 facility operation. Very little has
been published on this programme to date.

Figure 14. Republic of India Roadmap, from [6]. © DR.

38© IRSN/2018 – All rights reserved

5/ Nuclear fusion reactor
projects

5/2

Fusion power reactor projects
5/2/1

South Korea’s “K-DEMO” project
By 2037, the Republic of Korea plans to build a reactor called“K-DEMO” whose vacuum vessel will have a similar volume to thatof the ITER facility. It would be operated in two phases:•• an initial phase would serve to test various components (tritium
breeding blankets, equipment for monitoring plasma behaviour
and performance, etc.). During this phase, the fusion power
output would be limited to 10% of maximum capacity, with anavailability of around 10-20%;•• following replacement of certain components, the aim of thesecond phase would be to produce significant electrical power,with availability of at least 70%, with a view to the design offuture industrial fusion power reactors.

The South Korean project
aims to build a DEMO reactorof similar size to ITER, which
would ultimately produce
significant electrical power,with availability of at least
70% for fusion power output
of up to 2,400 MW.
Vertical maintenance ispreferred. The tritium
breeding blanket would bewater cooled and made oftritium-producing lithiumbased oxide ceramic.

Two fairly similar options are being studied, with 1,700 MW and
2,400 MW fusion power output respectively, and corresponding net
power of 100 MWe and 300 MWe. This project has been the subjectof quite a large number of publications.

Figure 15. Design of the South Korean K-DEMO reactor, from [7]. © DR.

39© IRSN/2018 – All rights reserved

Water Cooled Ceramic Breeder (WCCB) tritium breedingblankets would be used, with a lithium-based oxide ceramic forthe production of tritium, beryllium as a neutron booster andpressurised water for cooling. Vertical maintenance is preferred(Figure 15).
5/2/2

Europe’s “DEMO” project

The European Atomic Energy
Community is aiming tobuild a DEMO reactor with adesign similar to ITER, which
would ultimately produce
500 MW of electrical powerwith good availability for afusion power output of up to
1,950 MW.
Vertical maintenance seems
to be preferred. Several typesof tritium breeding blanket
are under consideration.

The European Atomic Energy Community aims to operate
a European DEMO reactor around 2040. Its operation will startwith an initial DEMO 1 phase, during which, based on designs
similar to those of the ITER facility, the tritium breeding blankets
and the divertor will be improved and the availability will bemuch higher (approximately 30%). The purpose of the second
operating phase (DEMO 2) will be to supply a net electrical powerof around 500 MWe from fusion power output of approximately
1,950 MW, with good availability. This project has been thesubject of a large number of publications.
Several types of tritium breeding blanket are being studied witha view to making a choice around 2020.
For the general design of the European DEMO project tokamak,
vertical maintenance seems to be preferred (Figure 16).

Figure 16. Design of the European DEMO reactor [8].

40
© IRSN/2018 – All rights reserved

5/ Nuclear fusion reactor
projects
5/2/3

Japan’s project
Until 2014, Japan had long presented the SlimCS (“slim” central
solenoid) project, with fusion power output of approximately
3,000 MW, as the next fusion facility for this country. Then, Japan
turned to a new project with fusion power output of 1,300 to
1,500 MW. For want of an official name, this project, which has
been the subject of a few publications, will be referred to as the“new Japanese project” in the remainder of this document. Finallyin late 2015, Japan published a roadmap, drawn up by a team madeup, at the request of the of Ministry of Education, Culture, Sports,
Science and Technology (MEXT), of members of the National
Institute for Fusion Science (NIFS) and the Japan Atomic Energy
Agency (JAEA), which foresees construction in around 2030 of a
DEMO reactor with fusion power output of around 1,000 MW [9].

The Japanese project aims tobuild a DEMO reactor withfusion power output of up to
1,000 MW.

41
© IRSN/2018 – All rights reserved

6/ The main differences
between the planned reactors
and the ITER experimental
facility

6/

The main differences
between the planned
reactors and the ITER
experimental facility
While the conceptual design of the tokamak is the same for the
ITER experimental facility and the planned nuclear fusion facilities,
the latter differ mainly by aiming to be self-sufficient in tritiumand by significantly longer operating times. These differences havea significant impact on the design of the new facilities.
6/1

The main development
priorities associated withplanned nuclear fusion
reactors are aiming to obtainself-sufficiency in tritium
and to significantly increase
operating times with plasma.

Self-sufficiency in tritium
Tritium from the detritiation of heavy water from current CANDU
reactors, which will be used for the ITER facility, will not be available
in sufficient quantities to supply DEMO reactors which will needmuch more tritium. A DEMO reactor with fusion power output of
1,000 MW operating continuously will consume approximately
60 kg of tritium annually, whereas the ITER facility will only need
20 kg of tritium over its 20 years of operation during which it will
operate with plasma on average approximately 1% of the time.
To supply tritium for DEMO reactors, the tritium breeding blanketsof these reactors are now set to produce more tritium than thereactor consumes, in order to take account of tritium losses due to
its radioactive decay and those due to its permeation in the reactor
equipment. A nuclear fusion facility should also be able to produce

43© IRSN/2018 – All rights reserved

For self-sufficiency in tritium,
the ratio between tritium
production and consumption
in the facility must be greaterthan 1 and as low as possible.

Everything that increases
the tritium burn-up ratein the plasma, the tritium
production rate in theblankets and the surface area
covered by these blankets
inside the vacuum vessel,
along with everything that
reduces tritium recycling
time, facilitates the facility
being self-sufficient intritium.

44© IRSN/2018 – All rights reserved

the tritium needed for start-up of the new nuclear facilities built
after it, until these facilities themselves become self-sufficient.
The tritium breeding blankets are therefore characterised by theratio between tritium production and consumption in the facility,
called the Tritium Breeding Ratio (TBR), which should be greaterthan 1. In various publications, the minimum TBR for a DEMO
reactor to be self-sufficient ranges from 1.1 to 1.2.
The lower the minimum TBR to be achieved, the easier it will beto obtain the desired self-sufficiency. In this respect, an importantparameter is the percentage of tritium introduced into the vacuum
vessel that undergoes a nuclear fusion reaction with deuterium,
called the “burn-up fraction” in the plasma. This fraction is low
because the fuel that is introduced into the vacuum vessel does
not all reach the interior of the plasma and because only partof the fuel that enters the plasma undergoes fusion reactions,as the presence of helium in the plasma (product of the fusion
reactions) hinders fusion reactions. When fuel is introduced into
the vacuum vessel as gas, only 5% of the fuel enters the plasma.
However, when the fuel is introduced as pellets, 50-90% of thefuel reaches the interior of the plasma. For the ITER facility,
the burn-up fraction in the plasma is not expected to exceed
0.3%. Each plasma discharge of several minutes will involve
approximately 100 g of tritium, of which approximately 0.3 g will
be burnt. Several grams will remain adsorbed on the first wallsof the vacuum vessel internal components, and the rest must be
recycled in the fuelling system. For DEMO reactors, the projectedburn-up fraction in the plasma ranges from 1 to 5% in various
publications [10-12]. For the same fusion power, the greater theburn-up fraction, the less tritium needs to be introduced into
the vacuum vessel and consequently the lower the quantity oftritium to process in the tritium building. The quantities of tritium
lost by radioactive decay and by permeation are also lower andthe minimum TBR to be achieved is lower. In summary, the higher
the burn-up rate in the plasma, the lower the minimum TBR to beachieved and the more feasible self-sufficiency in tritium appears.
Another factor, which is less influential than the burn-up ratein the plasma, may be considered for reducing the minimum
TBR to be achieved. This is the time needed to process tritiumin the tritium building before recycling; the shorter this time,
the lower the minimum TBR to be achieved. In the ITER facility,
there is no tritium recycling during plasma discharges given theirshort duration. For DEMO projects, the time needed for tritium
recycling has been estimated, as an initial approximation, to6/ The main differences
be between 1 and 24 hours, depending on the recycling processadopted and the associated technologies [13].

between the planned reactors
and the ITER experimental
facility

Furthermore, to achieve the minimum TBR, tritium production
should be as high as possible. For this, the choice of tritium breedingblankets, among the numerous types being studied, will be essential,
and the surface area covered by the tritium breeding blankets inside
the vacuum vessel should be as large as possible. However, thetritium breeding blankets cannot cover the whole internal surfaceof the vacuum vessel due to the presence of the divertor andvarious components that access the interior of the vacuum vessel
to inject fuel and additional heat, and to monitor plasma behaviour
and performance, along with disruption mitigation systems andthose for removing and replacing internal components.
Numerous publications covering the ability of a nuclear fusion
facility to be self-sufficient in tritium show that this objective,
which is essential for industrial operation of tokamak fusion powerreactions, is difficult to achieve.
IRSN considers that the impact on safety and radiologicalprotection of the choices made to become self-sufficient intritium also need to be taken into account. As will be highlighted
later in this document, tritium releases into the environment
will strongly depend on the type of tritium breeding blanket
adopted and, to a lesser extent, on the target burn-up fractionin the plasma.
6/2

Significant operating times
For the ITER experimental facility, the average operating time withplasma will only be about 1% of the time. For the planned DEMO
reactors, the target operating times with plasma range from 30%
to 70% of the time.
One of the consequences of this much greater operating time withplasma compared with that of the ITER facility, is that the numberof displacements per atom(4) (dpa) due to 14 MeV neutrons in thematerials of the structures surrounding the plasma, around 2 to
3 dpa for the ITER facility, could reach 150 dpa for a high-power
DEMO with high availability. Furthermore, the quantities of helium
and hydrogen produced from neutron reactions in these materials,
which will be negligible for the ITER facility, will be significant for
DEMO reactors and will constitute an additional source of damage
for these materials. The choice of materials able to withstand(4)
Mean number ofdisplacements to which
the atoms of a material are
subject under a neutron flux.
This number measures themodification of the structureof the material, which leads
to a deterioration of its initialproperties.

45© IRSN/2018 – All rights reservedintense neutron bombardment is therefore essential for thedesign of DEMO reactors.

To increase the plasma
operating time it is necessary
to reduce the duration ofmaintenance operationsin the vacuum vessel. Thisconsists in increasing thesize of the components thatmake up the blanket andtherefore the penetrations,the transport vehicles andthe hot cells where they will
be processed.

46© IRSN/2018 – All rights reserved

The significant operating times sought with plasma also imply
major design modifications compared with the ITER facility.
The time needed to extract some or all of the ITER facility’s
440 blanket modules and 54 divertor cassettes from the vacuum
vessel and transport them to the hot cells is totally incompatible
with the long operating times targeted with plasma for DEMO
reactors (for the ITER facility, it is estimated that it will takeabout two years to replace the entire first wall of the blanket andthe first wall of the divertor; changing the divertor alone would
require around six months of shutdown). To reduce maintenance
operation times in the vacuum vessel, the designs of the planned
DEMO reactors are leaning towards the installation of blankets
made up of a few sectors or half-sectors that can be removed fromthe vacuum vessel via very large penetrations (see Figures 12, 15
and 16). The volume and weight (30-720 tonnes) of these sectors
or half-sectors implies having automated transport vehicles thatare themselves extremely large and heavy. The hot cells thatwill receive the sectors or half-sectors for maintenance will also
therefore be very large (six times the surface area of the ITER
facility hot cells for the European DEMO project).

7/ Safety and radiation
protection issues to be
examined from the designphase of DEMO reactors

7/

Safety and radiation
protection issues to be
examined from thedesign-phase of DEMO
reactors
7/1

Residual heat removal
7/1/1

During operation without plasma
For the ITER facility, there is relatively little residual heat to be
removed (11 MW at plasma shutdown and 0.6 MW after one day).
The internal structures of the cryostat are at very low temperatures.
In the event of failure of all cooling systems, the temperature risein the tokamak structures would be slow, especially since in thissituation air would be introduced into the cryostat to further
slow this rise in temperature. It would take about four months forthe temperature of the divertor to become hot enough to cause
failure of its cooling system and consequent water ingress into thevacuum vessel. There could be a large proportion of dust sufficiently
hot to be oxidised by water (temperature above about 350°C) andhydrogen production in the vacuum vessel would be significant.
A hydrogen or dust explosion in the vacuum vessel would release
tritium and dust into the environment.

47© IRSN/2018 – All rights reserved(5)Roughly speaking, a low
activation material should be
able to be handled without
special precautions after aperiod of 100 years.

For DEMO reactors, the residual heat to be removed would bemuch greater than for the ITER facility, due to the longer operatingtimes with plasma and, for some projects, the greater fusion
power. Thus, for a high power DEMO reactor with high availability
using the same materials as the ITER facility, the residual heat to
be removed could be one or two orders of magnitude greaterthan for the ITER facility [14]. However, use of lower activationmaterials(5) (martensitic steel, vanadium alloy, silicon carbide
composites, etc.), which is planned for all DEMO reactor projects
to limit risks of exposure to ionising radiation, should also reduce
the residual heat to be removed. In these materials, elementsthat undergo rapid radioactive decay after activation, such as
tungsten or tantalum, would replace certain elements currently
used such as molybdenum, niobium and nickel.
Table 1 below gives the residual heat values from a number ofrecent publications for several DEMO reactors. They are difficult
to compare as the operating times with plasma are rarely
specified.
Residual heat

Planned facility

Fusion power

At shutdown

After oneday

Publication

Européen DEMO

2,000 MW

80 MW

30 MW

[15]

New Japanese project

1,350 MW38 MW

8.3 MW[16]

Former Japanese
project (SlimCS)

3,000 MW54.4 MW

11.3 MW[17–18]ITER

500 MW

11 MW

0.6 MW

Tableau 1. Some residual heat values from a number of recent
publications for several DEMO reactors.

To give a sense of scale, for a French N4-series pressurised water
reactor, providing 1,450 MWe of electrical power, the residual
heat is 264 MW immediately after reactor trip and approximately
24 MW after one day.
On the basis of estimates of the residual heat to be removed,
designers should assess the possible consequences of total
failure of the tokamak’s cooling systems. In this regard, a study
performed by the Materials Assessment Group (MAG) for theEuropean DEMO reactor project calculated a temperaturerise for the tungsten first wall of the blanket of approximately
1,000°C after 10 days, which in the event of probable air
ingress into the vacuum vessel due to failure at a penetration
given the temperatures reached, could lead to the formation of48© IRSN/2018 – All rights reserved

7/ Safety and radiation
significant quantities of radioactive tungsten trioxide (WO3): 10 to
100 kg/hr depending on the type of tungsten alloy for a first wall
surface area of 1,000 m2. These highly volatile aerosols could be
dispersed into the environment in the event of deterioration of theconfinement barriers following total failure of a tokamak’s coolingsystems [19, 20].
An assessment by analogy should be made for other DEMO reactor
projects, including those for which the residual heat to be removed
is equivalent to that of the ITER facility, as the consequences of therising temperature in the tokamak structures could be different, inparticular due to the presence of tritium breeding blankets.

protection issues to be
examined from the designphase of DEMO reactors

The consequences of the
failure of cooling systems
in nuclear fusion facilities
depend on the residual heatto be removed. For DEMO
reactors, the residual heat
will be greater than for the
ITER facility.

The tokamak design, and in particular the design of its coolingsystems, is therefore very heavily dependent on the assessmentof issues concerning residual heat removal. IRSN therefore
considers that it is essential to cover these issues from thesafety options stage.
7/1/2

During sector transfer, and storage and maintenance in hot
cells
As mentioned above, DEMO reactor blanket sectors to be transferred
to hot cells will be very large. For these reactors (or at least someof these reactors), it is therefore probable that the residual heat ofeach of these blanket sectors will be significant and that, in contrast
to what is planned for the ITER facility, these sectors will need to becooled during the various transfer operations between the vacuum
vessel and the hot cells.
For example, the residual heat from one sector of the former
Japanese SlimCS DEMO reactor project was estimated at 4.55 MW
at shutdown and 0.26 MW after one month. Under these conditions,the designers considered that it would be necessary to wait one
month before transporting such a sector, and that it would need
to be cooled during transfer because, without cooling, the sector
would reach a temperature of approximately 1,000°C after around40 days [18, 21].
Designers should therefore examine, from the safety options
stage, the possible consequences of total failure of the coolingof a sector when it is transferred between the vacuum vessel
and the hot cells, and during processing and storage in the hotcell building (Figure 17).

49
© IRSN/2018 – All rights reserved

Figure 17. Cooling for the blanket sectors on the Japanese SlimCS
reactor project [18]. © Kenji Tobita/JAEA.

•• During preparation of a sector for transfer:
During this operation, the sector is disconnected from its
normal cooling system to be connected to a transportspecific cooling system. The possible consequences of failureof the transport-specific cooling system should therefore be
examined, taking into account possible reconnection of thenormal cooling system.
Furthermore, it should be noted that when a sector to betransferred is disconnected from the normal cooling system,
the other sectors in the vacuum vessel must continue to bereliably cooled. The design of the cooling system(s) for blanket
sectors must take this requirement into account.
•• During sector transfer between vacuum vessel and hot cells:
Due to the expected size ofthe blanket sectors of DEMO
reactors, the residual heatto be removed will be high.
They should be specificallycooled at each step oftransfer. Transfer times,
cooling systems specificto automated transfer
systems and the places ofmaintenance and storage
must be considered.

50© IRSN/2018 – All rights reserved

Study of the potential consequences of a cooling failure during
the transfer must consider not only the maximum time sucha transfer could take, but also incidents that could lead to thetransfer being stopped for a long period. These assessments
should take into account the design of the cooling systems
installed on the automated transfer systems on which thedesign of these means of transfer depends, and that of the
facility areas to be crossed (size of the automated transfer
casks, possible emergency cooling system, etc.). For example,
the JAEA considered it necessary to install cooling panels on

7/ Safety and radiation
the walls of a corridor of the former Japanese SlimCS DEMO
reactor project so that the concrete of this corridor remained
at a temperature below 65°C to avoid deterioration of thisconfinement barrier due to evaporation of the water containedin the concrete [18].

protection issues to be
examined from the designphase of DEMO reactors•• During sector maintenance and storage operations in the hot
cells:
A certain number of sectors with potentially high total residual
heat may be grouped in the hot cell maintenance and storage
zones. The designers of the European DEMO reactor considered[8, 22] that stored waste from renovation of a complete setof vacuum vessel internal components should be cooled forapproximately 18 months. Again, the possible consequences offailures must be examined from the safety options stage.
The design of the blanket sectors, and in particular their coolingsystems in the vacuum vessel, in the automated transfer systems
and in their dedicated maintenance and storage areas, is highly
dependent on the assessment of residual heat removal issues.
IRSN therefore considers that it is essential to cover these issues
from the safety options stage.
7/2

Ionising radiation exposure risks
Like the ITER facility, exposure of workers to ionising radiationin DEMO reactors would mainly be associated with humanintervention during maintenance operations. Just as for the ITER
facility, DEMO reactors will undergo maintenance that largely
relies on robotics in order to reduce the doses received by workers.
Furthermore, as stressed above, research is underway regarding
the use of materials with lower activation under neutron flux thanthose used on the ITER facility.
The optimisation work conducted for the design of the ITER facility
adopted individual internal dose targets close to zero (a no dosetarget being incompatible with the presence of a large quantity oftritium, which easily diffuses through certain seals and materials)
and a mean individual dose of 2.5 mSv/year, with a maximum
individual dose of 10 mSv over a year and an annual collective doseof 500 person.mSv. For DEMO reactors, it is not certain that use of
low neutron activation materials will be enough to achieve similarvalues. It is therefore important that the design optimisation work
be widely developed, in particular for areas where the ionising

51© IRSN/2018 – All rights reserved

radiation exposure risks could be significantly greater than for
the ITER facility.
Particular consideration must be given to possible exposure risks
resulting from activation of tritium breeding blankets. Activationof the LiPb eutectic liquid used in some types of blanket underconsideration, produces mercury-203 (203Hg) and polonium-210(210Po), whose respective dose factors per ingestion are 100 and
100,000 times higher than for tritiated water. For maintenancephases requiring the opening of a system having carried LiPb,
the risk of dispersion of the volatile isotopes 203Hg and 210Po
into the room should be taken into account [23]. Furthermore,
LiPb is corrosive for the materials of the structures in which itis transported [24]. Precipitation of activated corrosion productsin the cold parts of systems (heat exchangers, etc.) is therefore
a possibility, which could lead to the creation of areas of highexposure to ionising radiation [13].
Due to the expected large
size of DEMO reactor blanket
sectors, activation of thetritium breeding blankets
and plasma operatingtimes much longer than for
ITER, the ionising radiation
exposure risks for operators
are greater.

Furthermore, there are exposure risks associated with highenergy γ radiation from nitrogen-16 (with a half-life of 7.1 s)
produced by neutron reactions with the oxygen in water. For
the ITER facility cooling system rooms, given its low operatingtime with plasma, these risks are avoided by simply prohibiting
access to the rooms during operation with plasma. For DEMO
reactors, given that operating times with plasma will be much
longer, it may be necessary to implement biological shielding in
the corresponding rooms [15].
The dose rates from DEMO reactor blanket sectors or half-sectors
will be much greater during their extraction for maintenancethan those associated with the ITER facility blanket modules.
Thus the dose rates associated with a European DEMO reactor
half-sector have been estimated, as an initial approximation, ataround 3 kGy/hr at contact [25]. Design provisions must be madeto limit worker exposure during transfer of these components
between the vacuum vessel and the hot cells and during theirmaintenance and storage in these cells.
As for any nuclear facility, the designer of each DEMO
reactor is responsible, from the design phase, for conducting
optimisation work for the doses received by operators, leading
to the proposal of dosimetry targets.

52© IRSN/2018 – All rights reserved

7/ Safety and radiation
7/3

Types of accidents to consider

protection issues to be
examined from the designphase of DEMO reactors

A tokamak is a complex device housing equipment carrying highenergy which could cause accidents or have an impact during an
accident. This could lead to failure of the first confinement barrier
for radioactive substances, made up of the vacuum vessel and its
extensions (relief tank, etc.), and even the second confinement
barrier, made up of the rooms surrounding the tokamak. The mainenergies to be considered are:•• plasma energy (approximately 700 MJ for the ITER facility),equally shared between thermal and magnetic energy. Plasma
disruption leads to thermal shocks in the first wall of the blanket,
with production of dust and the appearance of eddy currents
and electromagnetic loads, which are taken into account in thedesign of vacuum vessel internal components. In some cases,
such a disruption can create an electron beam which hits thefirst wall of the blanket creating damage which can cause
cooling system failure leading to water ingress into the vacuum
vessel, which could cause release of radioactive substances(tritium, activated dust and activated corrosion products) into
the environment (see Chapter 2);•• the magnetic energy of the field coils (around 50 GJ for the
ITER facility). In the event of loss of coil superconductivity (forexample due to a leak of their liquid helium coolant), the drop
in electric current in the coils could lead to the appearance ofeddy currents and electromagnetic loads, which are taken into
account in the design-basis used for the vacuum vessel internal
components and the vessel itself. Furthermore, a short circuit
in a coil could cause local deformation of this coil. Finally, anelectric arc could occur, causing local melting of the metal
structures of the coil and nearby equipment such as the vacuum
vessel or the cryostat. The integrity of the vacuum vessel is notaffected, unlike that of the cryostat. However, air ingress into thecryostat would not lead to a release of radioactive substances
into the environment;•• thermal energy from the primary cooling systems. Any waterleak from a cooling system into the vacuum vessel, into thecryostat or into the rooms surrounding the tokamak causes
vaporisation and rising pressure phenomena. Water ingress into
the vacuum vessel leads to a release of radioactive substances(tritium, activated dust and activated corrosion products) into
the environment (see Chapter 2). Water ingress into the cryostat

53© IRSN/2018 – All rights reserveddoes not affect its integrity. A cooling water leak into a roomaround the tokamak would lead to a release of activatedcorrosion products into the environment via the room’s
ventilation system;•• energy released in the event of an explosion of hydrogenisotopes (hydrogen, deuterium, tritium) or in the event of adust explosion:−−an explosion involving deuterium and tritium could occur inthe event of air ingress into the vacuum vessel. Furthermore,in the event of water ingress into the vacuum vessel, an
explosion of the hydrogen produced by oxidation of dust
by water, along with deuterium and tritium, could occur inthe vacuum vessel relief tank,−−a dust explosion could also be triggered by a deuterium andtritium explosion in the vacuum vessel.
•• the cooling energy from the system that supplies liquid heliumat 4.5 K (the tokamak building of the ITER facility contains
about 20 tonnes of liquid helium). Any leak of helium into thecryostat or the rooms surrounding the tokamak would cause
sudden vaporisation of the liquid helium, with large volumetric
expansion and a pressure rise in the affected area;•• energy from activation products (see Section 7.1.1).
Identification of accidents that could occur in the ITER facility
was made difficult by the novelty of the issues to be covered.
An initial list of accidents was drawn up pragmatically, based onforeseeable failures that could lead to significant consequences(water leak, air ingress, etc.). More detailed methods were thenused, which consist in starting either from component failures
or possible causes of releases. Finally, the risk of accidents that
could result from an external hazard (earthquake, etc.) or aninternal hazard (fire, etc.) were examined. Ultimately, 10 designbasis accidents were covered by the ITER facility operator in apreliminary safety report, which added 12 beyond design-basis
accidents, which are either design-basis accidents for which one
or more additional failures were considered, or accidents deemed
to be of very low probability. Technical discussions during
examination of this report led to the operator modifying certain
accident scenarios and studying a few more. The list of accidents
finally adopted for the ITER facility, and the doses which could
be received by the public 2.5 km from the facility are as follows:

54© IRSN/2018 – All rights reserved

7/ Safety and radiation
For design-basis accidents:

protection issues to be
examined from the designphase of DEMO reactors•• water ingress into the vacuum vessel (long-term dose at 2.5 km:
10-4 mSv),•• air ingress into the vacuum vessel (long-term dose at 2.5 km:
1.3 10-2 mSv),•• vacuum vessel primary cooling system leak outside the vacuum
vessel (long-term dose at 2.5 km: 10-5 mSv),•• divertor primary cooling system leak outside the vacuum vessel(long-term dose at 2.5 km: 1.8 10-2 mSv),•• leak in the isotope separation system (long-term dose at 2.5 km:
1.1 10-4 mSv),•• failure of a fuelling line (long-term dose at 2.5 km: 4.9 10-3 mSv),•• loss of confinement in the hot cell building (long-term dose at
2.5 km: 3.6 10-4 mSv),•• leak from a tritiated water tank (long-term dose at 2.5 km:
1.6 10-2 mSv),•• loss of confinement on an automated transfer cask (long-term
dose at 2.5 km: 4.3 10-3 mSv),•• failure of the largest opening of a glove box (long-term dose at
2.5 km: 3.1 10-5 mSv).
For beyond design-basis accidents:•• air ingress into the vacuum vessel and leak into the vacuum
vessel from the blanket primary cooling system (long-term doseat 2.5 km: 4.3 10-3 mSv),•• water ingress into the vacuum vessel and failure of a vacuum
vessel penetration (long-term dose at 2.5 km: 0.13 mSv),•• loss of control of the plasma without fusion power shutdown
and water ingress into the vacuum vessel (long-term dose at
2.5 km: 3.4 10-5 mSv),•• blanket primary cooling system leak outside the vacuum vessel
and failure of the fusion power shutdown system (long-term
dose at 2.5 km: 10-2 mSv),•• air ingress into the vacuum vessel causing an explosion (longterm dose at 2.5 km: 0.2 mSv),•• significant breach of vacuum vessel and cryostat (long-term
dose at 2.5 km: 0.3 mSv),

55© IRSN/2018 – All rights reserved•• vacuum vessel primary cooling system leak outside thevacuum vessel and loss of coolant flow in the other systems(long-term dose at 2.5 km: 3.6 10-4 mSv),•• water and helium leaks into the cryostat (long-term dose at
2.5 km: 2.4 10-3 mSv),•• failure of a fuelling line and failure of the detritiation system(long-term dose at 2.5 km: 1.6 mSv),•• fire in the tritium building (long-term dose at 2.5 km:
0.17 mSv),•• leak from the isotope separation system into its room thenexplosion (long-term dose at 2.5 km: 0.13 mSv),•• fire in the waste treatment area with propagation towards thestorage area (long-term dose at 2.5 km: 0.3 mSv),•• tritium explosion during the regeneration phase of a vacuum
vessel cryopump (long-term dose at 2.5 km: less than 0.3 mSv).
The vacuum vessel isthe first confinement
barrier while the rooms
surrounding the tokamak
constitute the second.
Foreseeable failures orexternal hazards that could
cause radioactive releases
into the environment havebeen considered for the
ITER facility. The increased
complexity of DEMO reactors
and the larger quantities ofmatter and energy involved
mean that the likelihood andconsequences of accidents
are different than those for
the ITER facility.

In addition, the following accidents were excluded:•• presence of an operator near an automated transfer cask
carrying a vacuum vessel internal component;•• fall of an elevator with an automated transfer cask carrying avacuum vessel internal component;•• fire jeopardising the storage of activated dust;•• fire jeopardising the storage of purely tritiated waste;•• spurious opening of a rupture disc or a valve on the vacuum
vessel relief line during a plasma discharge.
For the handful of DEMO projects that have been the subject ofpublications on the accidents considered, it would appear thatthe types of accidents considered are almost identical to thoseadopted for the ITER facility. However, the operator must take into
account the fact that the likelihoods and consequences could besignificantly different, in particular due to the greater quantities
of radioactive substances or energy involved. Furthermore, given
the increased complexity of DEMO reactors compared with the
ITER facility, it is possible that other types of accident may need
to be studied (see Sections 7.3.2, 7.3.3, 7.3.6 and 7.3.7 below).
These issues must be examined from the design phase. A nonexhaustive outline of such an examination is given below.

56© IRSN/2018 – All rights reserved

7/ Safety and radiation
7/3/1

Changes in the quantities of tritium and dust in the vacuum
vesselprotection issues to be
examined from the designphase of DEMO reactors

The consequences of accidental water or air ingress into the vacuum
vessel of the ITER facility have been deemed acceptable, given thelimitations on the quantities of tritium (1 kg) and dust (1,000 kg) in
the vacuum vessel adopted by the operator. On the basis of current
knowledge, these values could be reached after one or two yearsof operation, i.e. after 100-200 hours of operation with plasma.
The duration of the thermal desorption operations that will thenbe performed to reduce the quantities of adsorbed tritium and thecleaning operations to reduce the quantities of dust should notaffect the expected availability for this experimental facility.
For DEMO reactors, where the operating times with plasma aremuch longer, limitations identical to those adopted for the ITER
facility would require frequent shutdowns, incompatible with thedesired availability rates.
If the designers adopt the same limitations, they will need to put
forward provisions to reduce the adsorption of tritium on the
structures and to reduce dust creation, or reduce the time it takes
for vacuum vessel detritiation and dust removal. Operating with thefirst wall of the blanket at a higher temperature would promote
lower adsorption of tritium. In addition, a higher burn-up fractionin the plasma would also be favourable. Furthermore, the designers
could look into ways to control the inventory by attempting tocollect dust at the divertor during operation with plasma [26].

The design of DEMO
reactors, associated with alimitation on the quantitiesof tritium and dust in thevacuum vessel, must ensurethat the consequences ofwater or air ingress into thevessel are acceptable for theenvironment.

The designers could also examine other choices with regard to thelimitations on the quantities of tritium and dust in the vacuum vessel
and show that, with the new limitation adopted, the consequencesof accidents, in particular air ingress into the vacuum vessel followedby an explosion of hydrogen isotopes and dust, would be acceptable,
given the design provisions adopted elsewhere.
It is therefore necessary for the corresponding elements to beprovided from the safety options stage.
7/3/2

Presence of tritium breeding blankets
The presence of much more extensive tritium breeding blankets
in the DEMO reactors compared with the ITER facility means thatdesigners must re-examine the consequences of the accidents
considered, and even study other accidents.

57© IRSN/2018 – All rights reserved

To this end, it is necessary to determine the quantities ofradioactive substances that could be present in the tritium
breeding blankets , which could be very different depending on
the type of tritium breeding blanket adopted.

The accidents to beconsidered for DEMO
reactors depend on the typeof tritium breeding blanket
adopted.

For the ITER facility blanket, the main accident adopted was
the risk of a water leak from its cooling system into the vacuum
vessel. For the planned DEMO reactors, the tritium breedingblankets could be cooled by liquids (water, LiPb eutectic) or gases(helium). The designers should therefore examine the possible
consequences of leaks of these fluids into the vacuum vessel
and outside this vessel, and possible violent reactions (including
during transfer and maintenance in the hot cells). For example,
the risks of interactions between water and beryllium or the LiPb
eutectic (leading to hydrogen production) and between helium
and this eutectic are to be considered for certain types of blanket[28, 29].
The corresponding elements should therefore be provided
from the safety options stage.
7/3/3

Increase in the number of possible cases of loss of controlof the plasma
To manage the multiple types of instability that may occur inthe plasma, the designers plan to install equipment to monitor
plasma behaviour and performance (monitoring the neutron
flux, the magnetic field, etc.) and actuators (fuelling control
valves, additional means of heating, etc.) with the corresponding
instrumentation and control. In the event of malfunction ofplasma control, the plasma stops abruptly (disruption).
For the ITER facility, plasma discharges are short discharges, called“inductive scenarios”, because the plasma current is produced by
gradually increasing the current in the central solenoid. For DEMO
reactors, to obtain longer operation with plasma, permanentdischarges without any particular limitation of duration areplanned, called “non-inductive scenarios”. The plasma current is
produced by additional heating, to which is added a significant
fraction of the current generated within the plasma itself bycertain types of instability. This self-generated current is called“bootstrap current”.
The plasma control system should be much more developed thanfor the ITER facility to obtain good control of plasma stability

58© IRSN/2018 – All rights reserved

7/ Safety and radiation
under these operating conditions. This means that the possible
malfunctions of the plasma control system will be more numerous
and different from those considered for the ITER facility (sudden
increase in fuelling rate or additional heating, etc.), and some ofthese malfunctions could be more severe.
Furthermore, the designers must also examine all cases ofmalfunction of the disruption mitigation systems (gas injection),in particular the consequences of spurious trip of these systems,
which could lead to a plasma disruption whose severity needs tobe assessed.

protection issues to be
examined from the designphase of DEMO reactors

The number and type of
possible malfunctions of
the plasma control system
for DEMO reactors will
be different from thoseconsidered for the ITER
facility.

It would seem necessary that the possible consequences ofthe most severe disruptions that could occur, given the design
provisions adopted, be estimated at the safety options stage,
with suitable substantiation of the choice of scenarios studied.
7/3/4

Increased magnetic energy of the toroidal field coils
The ITER facility was designed with dimensions such that the
integrity of the vacuum vessel could not be affected by an electric
arc in one of the 18 toroidal field coils. For certain DEMO reactors,
the magnetic energy of the toroidal field coils could be significantly
higher than in those of the ITER facility. For example, the magnetic
energy of the toroidal field coil of the European DEMO project
could be around 10 GJ (compared with 2.28 GJ for an ITER facility
field coil). Loss of integrity of the vacuum vessel due to an electricarc on a field coil could lead to ingress of vacuum vessel coolingwater into both the vessel itself and the cryostat, an accident which
has not been examined to date.
For DEMO reactors, the risk of loss of integrity of the vacuum
vessel due to an electric arc in a toroidal field coil should
therefore be examined from the safety options stage.
7/3/5

Increase in the quantities of helium used
Any significant leak of liquid helium, a fluid used for cooling thesuperconducting magnetic field coils of the ITER facility, can put
the room or equipment where it occurs under pressure and affectits integrity. In particular, the operator examined if there is a room
where there could be a simultaneous occurrence of disseminationof radioactive substances and a helium leak threatening the59© IRSN/2018 – All rights reserved

The presence of helium in
DEMO reactors, in particular
as coolant for the tritium
breeding blankets, must be
examined given the risk ofpressure on the confinement
barriers in the event ofleakage.

integrity of the room. Specific provisions, in particular the designbasis earthquake used for the helium system, were taken to avoid
this possibility. Furthermore, for the ITER facility, the quantity ofhelium that could be accidentally spilt into the vacuum vessel islimited by design to 25 kg. This is because, in the event of water
ingress into the vacuum vessel and more than 25 kg of helium,
which is a non-condensable gas, the integrity of the vacuum
vessel and of the vacuum vessel relief tank could be threatened.
For some future DEMO reactors, the quantities of heliumin the tokamak building will likely be greater, in particular for
those which use gaseous helium as the coolant for the tritium
breeding blankets. Several publications stress the need to provide
expansion volumes to limit pressure on the confinement barriersin the event of helium leakage [30, 31].
It is therefore necessary that for the DEMO reactors concerned,
the designers take into account the risks of helium leakage
from the safety options stage.
7/3/6

Increase in the number of rooms where there could besignificant quantities of hydrogen isotopes outside thevacuum vessel
For the ITER facility, the operator studied the case of an explosionof all the hydrogen isotopes in the room housing the isotope
separation system for the fuelling system in the tritium building.
The total quantities of tritium used in some DEMO reactors will
be much greater than for the ITER facility. For these reactors, it
can therefore be expected that several rooms outside the vacuum
vessel could contain significant quantities of hydrogen isotopes.
The risks of explosion of these large quantities of hydrogenisotopes in these rooms must therefore be examined from thesafety options stage.
7/3/7

Vertical port in the vacuum vessel
For the DEMO reactor designs that plan for extraction of blanket
sectors or half-sectors via vertical penetrations, the risks of such a
sector or half-sector being dropped during its removal from thevacuum vessel must be examined from the safety options stage.

60© IRSN/2018 – All rights reserved

7/ Safety and radiation
Furthermore, use of a cover slab to protect the tokamak from
airplane crashes, as planned for the ITER facility, could be
difficult to implement for these reactors. This protection must
therefore be examined from the design options stage, taking into
account the provisions adopted for the extraction of internal
components.

protection issues to be
examined from the designphase of DEMO reactors

7/3/8

Specification of protection with regard to extreme events
Specification of all essential equipment designed to withstandextreme events must be considered for DEMO reactors from thedesign phase.
For ITER, this was, of course, only taken into account at a later stage.
Determination of the extreme events to be adopted shouldtake into account uncertainties for the phenomena considered,including the climate changes which could occur by the end ofoperation of the currently planned DEMO reactors.
7/4

Releases into the environment under normal operation
The radiological impact of releases into the environment due tonormal operation of a fusion facility mainly concern gaseous tritium
releases. The annual gaseous releases from the ITER facility should
not exceed 2.5 g of tritium in years of heavy maintenance (longterm shutdown to change the divertor, planned two or three times
in the service life of the facility) and 0.6 g of tritium for other years.
To calculate the consequences, tritium is assumed to be in the formof tritiated water (HTO), which is the worst-case inorganic form.
Organically-bound tritium (OBT) releases are not expected.
During heavy maintenance years, tritium releases will mainly be
due to the degassing of internal components in the vacuum vessel
during the long period of this being open and in the hot cells. The
release estimates stated above assume that a thermal desorption
operation will be performed before opening the vacuum vessel,
which should greatly reduce the quantities of tritium present. For
years without heavy maintenance, two-thirds of the tritium releases
will come from the Tritium Recuperation System (TRS), whose role
is to recover most of the tritium present in the activated dust andtritiated ILW-LL waste from the vacuum vessel internal components.

The radiological impactof releases into theenvironment due to normal
operation of a fusion facility
is mainly due to gaseous
releases of tritium, forwhich the greater quantitiesof tritium and the larger
number of release paths
implemented in DEMO
reactors must be taken
into account. Similarly, for
DEMO reactors there should
be a reduction in releasesof liquid effluents into theenvironment from the higher
number of air detritiation
systems and greater volume
of water detritiation systems.

61© IRSN/2018 – All rights reserved

The remaining releases could be caused by permeation of tritium
via equipment into the rooms.
For DEMO reactors, the use of greater quantities of tritium andadditional paths for gaseous tritium releases should be taken into
account.
In accordance with the optimisation principle, designers
should seek to reduce as much as possible the quantities oftritium in the facility and to examine the main release paths
from the design phase.
In addition, for the ITER facility, liquid effluents containing tritium
will mainly come from very slightly tritiated liquid effluents
produced by the air detritiation systems and from the treatmentof water from the cooling systems. The concentration of tritiumin these effluents is too low for them to be treated by the Water
Detritiation System (WDS), which recovers tritium for reinjection
into the fuelling system.
(6)
The estimated radiologicalimpact can be comparedto the annual permissible
dose limit for exposure ofartificial origin for the publicof 1 mSv/year specified
in Article R.1333-8 of theFrench public health code.

For DEMO reactors, which could contain numerous rooms fittedwith air detritiation systems and large volumes of cooling water,
the designers should examine the possibilities of reducing
the quantities of very slightly tritiated liquid effluents to bereleased into the environment [15].
For the ITER facility, the maximum radiological impact due toradioactive liquid and gaseous effluents is around 2.3 µSv/year(6).
7/4/1Reducing tritium quantities in the facility(7)
Seminar on 2 and3 dimensional modelling ofplasmas.

62© IRSN/2018 – All rights reserved

As stated above, the quantity of tritium present in the ITER facility
at any one time is limited to 4 kg. The quantities mentioned inpublications regarding DEMO reactors vary between 2.5 kg and7.5 kg [14, 26, 32, 33], most of which is found, just as for the ITER
facility, in the tritium buildings, with the remainder mainly in thevacuum vessel. In this regard, during the presentation made by
the University of Science and Technology of China at the “3Dversus 2D in hot plasma(7)” Seminar from 30 April to 2 May 2013,
it was shown that, when the quantities of tritium increase in afusion reactor, releases of tritium into the environment would
also increase as it is unlikely that lower leakage rates than thoseadopted for the ITER facility can be targeted in the design of
DEMO reactor equipment [3].

7/ Safety and radiation
In any case, to reduce the quantities of tritium present, several
publications note the interest of introducing a bypass into thefuelling system to supply the vacuum vessel with fuel that has
not passed through the isotope separation system (Figure 18).
One method considered would produce fuel pellets to be shot
into the centre of the plasma, using products extracted from thevacuum vessel by the cryopumps from which only the helium-4
and impurities (dust, etc.) would be removed [26, 33]. Such a
modification to the fuelling system could significantly reduce thequantities of tritium in the facility and consequent releases into
the environment.

protection issues to be
examined from the designphase of DEMO reactors

Figure 18. Fuelling system bypass, from [33]. © Georges Goué/IRSN.

Furthermore, it should be noted that, the higher the burn-up
fraction in the plasma, the less tritium needs to be introduced into
the vacuum vessel and consequently the lower the quantities oftritium to be recycled in the tritium building.
Finally, the quantity of tritium present in the tritium building will
be smaller the shorter the time needed for tritium recycling in thisbuilding.

In order to limit tritium
releases into theenvironment, the quantityof tritium used should be
reduced as much as possible(bypass, high burn-up
fraction, etc.).

63© IRSN/2018 – All rights reserved

7/4/2

Examination of the main possible paths for gaseous tritium
releases
7/4/2/1

Releases associated with the cooling systems for tritium
breeding blankets
In DEMO reactors, part of the tritium produced in the tritium
breeding blankets will diffuse via the walls of the cooling systems
that pass inside the blankets near the areas where tritium is
produced. Several publications mention the possibility of arounda hundred grams of tritium being found in these systems [34],while the quantities of tritium present in the two primarycooling systems of the ITER facility blanket and divertor areapproximately 0.7 g per system. Even if purification systems are
implemented to permanently extract tritium from the coolingsystems of DEMO reactor tritium breeding blankets, it is clearthat some of the tritium present in these systems will diffuse
via the cooling system heat exchangers into the environment, arelease path that does not exist for the ITER facility (Figure 19).
Research is underway regarding lining the walls of the coolingsystems inside the tritium breeding blankets with a protective
layer of alumina (Al2O3) or erbium oxide (Er2O3) [13], which
would reduce permeation by a factor of 50 to 1,000 [19, 35].
However, this research is still only at laboratory scale.

Figure 19. Releases associated with the cooling systems for tritium
breeding blankets, from [36]. © Georges Goué/IRSN.

64© IRSN/2018 – All rights reserved

7/ Safety and radiation
Initial estimates of releases from planned DEMO reactors vary
greatly depending on the type of tritium breeding blanket adopted.
Thus, for the European DEMO reactor, the releases associated withthe use of WCLL tritium breeding blankets (lithium-based oxide
ceramics for tritium production, beryllium as neutron booster
and water for cooling) are much lower than those that would be
associated with the use of HCPB blankets (lithium-based oxide
ceramics for tritium production, beryllium as neutron booster andhelium for cooling) or HCLL blankets (LiPb eutectic for tritium
production and as neutron booster, helium for cooling). The statedfactors are 55 and 160 respectively [37].

protection issues to be
examined from the designphase of DEMO reactors

In choosing the type of tritium breeding blanket to achieve selfsufficiency in tritium, there is a need to consider the tritium
releases into the environment associated with their coolingsystems.
7/4/2/2

Releases associated with transfers of internal components to the
hot cells and their processing in these cells
As mentioned above, under normal operation, tritium releases fromthe ITER facility are mainly associated with transfers of internal
components to the hot cells and their processing in the cells during
years of heavy maintenance, and with treatment of tritiated waste
in other years. It is difficult to get a relative appreciation of thereleases associated with planned DEMO reactors as the designof their blanket in sectors or half-sectors is so different fromthat of the ITER facility and because there are many factors that
could affect the volume of releases (quantity of tritium present
in the internal components transferred, frequency of transfers,
effectiveness of thermal desorption to reduce the quantities oftritium in the vacuum vessel, concentration of tritium in the waste
resulting from maintenance of this equipment, effectiveness of thewaste detritiation systems).
The designers of DEMO reactors should therefore carefully
examine, from the safety options stage, the tritium releases
into the environment which could be associated with transfersof internal components to the hot cells and their processing inthese cells.

65© IRSN/2018 – All rights reserved

7/4/2/3

Releases associated with waste detritiation equipment
If waste detritiation equipment is used (see Section 7.5 below),
this equipment could be a source of releases.
This path of releases into the environment should beconsidered from the design phase.
7/5

Waste

The type and quantity oftritiated waste produced
should be considered basedon the general policy of thehost country with regard towaste management.

Part of the waste from the ITER facility will contain quantitiesof tritium that are too high for it to be sent directly to waste
repositories in France. It is currently planned that they be
stored for approximately 50 years, in the planned INTERMED
facility, to reduce the quantities of tritium via radioactive decay.
Furthermore, the most tritiated waste will undergo pre-processingin the tritium recovery system (TRS).
It is probable that waste produced by DEMO reactors will be
more tritiated than that produced by the ITER facility. This could
lead designers to add specific detritiation equipment (thermal
furnace, melting furnace, incinerator, etc.). Furthermore, thetritium breeding blankets could be a source of waste different
from that from the ITER facility, depending on the type of tritium
breeding blanket adopted. In any case, management of DEMO
reactor waste will depend on the general policy of the hostcountry in this regard.
DEMO reactor designers should examine waste management
constraints from the safety options stage, taking into accountthe general policy of the country that will host the reactor.

66© IRSN/2018 – All rights reserved

8/ Conclusion

8/

Conclusion
The considerations above have been established on the basis of
the experience acquired during the safety assessment of the ITER
facility and publications available on planned DEMO reactors inlate 2017. Following this work, IRSN stresses that designers should
examine the following subjects as a priority:•• residual heat removal, taking into account the design envisaged
for the tokamak cooling systems and those for the blanket
sectors when they are transferred between the vacuum vessel
and the hot cells and during their storage in these cells;•• optimisation of the doses which workers receive depending on
robotisation and the choice of materials;•• the types of accident considered for the ITER facility and thespecific types of accident that could be associated with thedesign of DEMO reactors;•• possibilities for limiting the overall quantity of tritium present
in the facility and releases by various main paths for gaseous
tritium releases. In this respect, the choice of tritium breedingblankets would appear to be a key factor;•• identification of the waste management constraints based on
the general policy of the country hosting the reactor.

67© IRSN/2018 – All rights reserved

9/ Glossary

9/

Glossary
Blanket: a vacuum vessel internal component which provides
neutron protection for the metal walls of the vessel.
Collective dose: the dose is the quantity of energy absorbed bya medium from ionising radiation. The collective dose is the sumof the individual doses received by a given group of people (hereexpressed in person.millisieverts ([person.mSv]).
Cryopump: a vacuum pump system that removes helium andimpurities via condensation on a cold surface.
Cryostat: a metal chamber maintained at very low temperaturethat contains the tokamak.
Design-basis: determination of the characteristics of a facility
during its design to meet pre-established criteria and regulatory
practice.
Disruption: the name given to loss of plasma confinement in thevacuum vessel.
Divertor: an internal component located in the lower part of thevacuum vessel which extracts helium, fuel that has not undergone
fusion (and that can be reused) and impurities.
Dose rate: the radioactive dose rate determines the intensity ofirradiation (energy absorbed by matter per unit of mass and time).
It is measured in grays per second (Gy/s).

69© IRSN/2018 – All rights reserved

Field coils: systems made up of superconducting magnets which
generate a magnetic field that creates, confines and shapes theplasma. A distinction is generally made between toroidal fieldcoils (confinement), the central solenoid (field generation andcontribution to heating) and poloidal field coils (literally: whichdevelop between magnetic poles). These coils are cooled bypumping liquid helium at a temperature of 4.5 K.
Fusion: the combination of two nuclei of light atoms to form aheavier nucleus, which releases a large amount of energy carriedby the reaction products (nuclei, particles and radiation).
Magnetic confinement: plasma moving under the effect of amagnetic field, held at a distance from the walls of the vacuum
vessel. The magnetic field only acts on charged particles. All
neutral secondary fusion products (impurities, helium-4, etc.) are
extracted by the divertor. Neutrons are absorbed by the wallsof the vacuum vessel, giving up heat which is removed for ITER
and will ultimately be transformed into electricity for a DEMO
reactor.
Plasma: hot low-density gas made up of positive ions andelectrons, which have been stripped off due to the temperature,
produced under the action of a strong magnetic field.
Plasma discharge: the name given to the operating sequence ofthe tokamak with a plasma.
Residual heat: heat still produced by the activated materials
after the shutdown of the fusion reaction.
Solenoid: a device that consists of a coiled electric wire used tocreate a magnetic field.
Stellerator: a magnetic confinement nuclear fusion device thatdiffers from tokamaks, in which plasma confinement is achievedby a helical magnetic field created by a complex arrangementof field coils outside the torus. This configuration is designed tooperate continuously.
Tokamak: a Russian acronym for тороидальная камерас магнитными катушками, meaning “toroidal chamber withmagnetic coils” which uses magnetic fields to create, confine andcontrol a hot plasma inside which the fusion reaction can occur.
Toroidal form or torus: having the shape of an inner tube.
Tritium breeding blanket: a blanket that produces tritium while
also providing neutron protection for the metal walls of thevacuum vessel.

70© IRSN/2018 – All rights reserved

9/ Glossary
Vacuum vessel: a sealed toroidal stainless steel chamber in which
the fusion reactions (plasma) occur. It is the first confinement
barrier for radioactive substances and contributes to plasma
stability. It contains the blanket and divertor. It has a double steel
wall in which cooling water flows to remove the heat given off by
the fusion reactions. This heat will be used to produce electricity infusion power plants.

71© IRSN/2018 – All rights reserved

10/ References

10/

References
1/ R. Stieglitz and al., Safety of Fusion Power Plants in View of
Fission Regulations, 28th Symposium on Fusion Technology (SOFT
2014), San Sebastián, Spain, September 29-October 3, 2014.
2/ A. Vale, R. Ventura, P. Lopes, I. Ribeiro, Assessment of navigation
technologies for automated guided vehicle in nuclear fusion
facilities, Robotics and Autonomous Systems, 97, 153-170,
November 2017.
3/ Y. Wan, Road map of Chinese Fusion Research and the First
Chinese Fusion Reactor CFETR, “3D versus 2D in hot plasma” du
30 avril au 2 mai 2013.
4/ « Chine, le nouveau géant du nucléaire », Revue générale
nucléaire, n°5, septembre-octobre 2017.
5/ T.S. Taylor, A.M. Garofalo and al., An AT-Based Fast-Track Pathto DEMO, 1st IAEA DEMO workshop UCLA, USA, October 15-18,
2012.
6/ M. Abdou and al., Fusion Power Extraction and Tritium Fuel Cycle,
1st IAEA DEMO workshop UCLA, USA, October 15-18, 2012.
7/ G. S. Lee, Korean Roadmap and International Collaboration for
DEMO R&D, 2012.
8/ A. Loving and al., Pre-conceptual design assessment of DEMO
remote maintenance, Fusion Engineering and Design, 89, 22462250, 2014.
9/ H. Yamada and al., Development of strategic establishment oftechnology bases for fusion DEMO reactor in Japan, Journal of
Fusion Energy, Vol. 35, Issue 1, February 2016.

73© IRSN/2018 – Tous droits réservés10/ M. Ni and al., Tritium supply assessment for ITER and
DEMOnstration power plant, Fusion Engineering and Design, 88,
2422-2426, 2013.
11/ T. Tanabe, Tritium fuel cycle in ITER and DEMO: Issues inhandling large amount of fuel, Journal of Nuclear Materials, 438,
S19-S26, 2013.
12/ S. Willms, Tritium handling technology roadmap, 1st IAEA
DEMO workshop UCLA, USA, October 15-18, 2012.
13/ M. Abdou and al., Blanket/first wall challenges and require
R&D on the pathway to DEMO, Fusion Engineering and Design,
100, 2-43, 2015.
14/ J. Pamela and al., ITER tritiated waste management by the
Host state and first lessons learned for fusion development,
Fusion Engineering and Design, 89, 2001-2007, 2014.
15/ N. Taylor and P. Cortes, Lessons learned from ITER safety andlicensing for DEMO and future nuclear fusion facilities, Fusion
Engineering and Design, 89, 1995-2000, 2014.
16/ M. Nakamura and al., Study of safety features and accident
scenarios in a fusion DEMO reactor, Fusion Engineering and
Design, 89, 2028-2032, 2014.
17/ M. Zucchetti and al., Recent advances in fusion radioactive
material studies, Fusion Engineering and Design, 88, 652-656,
2013.
18/ K. Tobita and al., Reconsideration of tokamak DEMO concept
based on the latest design study, 1st IAEA DEMO workshop
UCLA, USA, October 15-18, 2012.
19/ C. Linsmeier, Advanced first wall and heat sink materials, 1st
IAEA DEMO workshop UCLA, USA, October 15-18, 2012.
20/ D. Stork and al., Materials R&D for a timely DEMO: Keyfindings and recommendations of the EU Roadmap Materials
Assessment Group, Fusion Engineering and Design, 89, 15861594, 2014.
21/ Y. Someya and al., Waste management scenario in the hotcell and waste storage for DEMO, Fusion Engineering and Design,
89, 2033-2037, 2014.
22/ J. Thomas and al., DEMO Active Maintenance Facility concept
progress 2012, Fusion Engineering and Design, 89, 2393-2397,
2014.

74
© IRSN/2018 – Tous droits réservés10/ References
23/ B.J. Merrill and al., Normal operation and maintenance safety
lessons from the ITER US PbLi TBM program for a US FNSF and
DEMO, Fusion Engineering and Design, 89, 1989-1994, 2014.
24/ S. Banerjee, Overview of Indian activities on fusion reactor
materials, Journal of Nuclear Materials, 455, 217-224, 2014.
25/ M. Coleman and al., Concept for a vertical maintenance remotehandling system for multi module blanket segments in DEMO,
Fusion Engineering and Design, 89, 2347-2351, 2014.
26/ C.E. Kessel and al., Fusion Nuclear Science Pathways Assessment,
rapport Princeton Plasma Physics Laboratory PPPL-4736, février
2012.
27/ F. Franza, Tritium transport analysis in HCPB DEMO blanket
with the FUS-TPC code, report-Nr. KIT-SR 7642, 2013.
28/ A. Li Puma and al., Design and development of DEMO blanket
concepts in Europe, 1st IAEA DEMO workshop UCLA, USA,
October 15-18, 2012.
29/ D. Carloni and al., Requirements for helium cooled pebble bed
blanket and R&D activities, Fusion Engineering and Design, 89,
1341-1345, 2014.
30/ F. Romanelli, The European Roadmap to fusion electricity.
Présentation au 28e SOFT à San Sebastien du 29 septembre au
3 octobre 2014.
31/ C. Bachmann and al., Initial DEMO tokamak design configuration
studies, Fusion Engineering and Design, 98-99, 1423-1426, 2015.
32/ C.P.C. Wong and B. Merryl, Use of system code to estimate
equilibrium tritium inventory in fusion DT machines, such as ARIESAT and components testing facilities, Fusion Engineering and Design,
89, 1482-1485, 2014.
33/ C. Day and T. Giegerich, The direct internal recycling concept tosimplify the fuel cycle of a fusion power plant, Fusion Engineeringand Design, 88, 616-620, 2013.
34/ M. Nakamura and al., Study of safety features and accident
scenarios in DEMO, Fusion Engineering and Design, 89, 2028-2032,
2014.
35/ T. Chikada and al., Future pass for the development of tritium
permeation barrier in fusion systems, 10th International Conference
on Tritium science and technology, “Tritium 2013”, Nice acropolis,
France, October 21-25, 2013.

75© IRSN/2018 – Tous droits réservés36/ S. Konishi, Zero Emission Fusion Tritium Plant, 10th
International Conference on Tritium science and technology,“Tritium 2013”, Nice acropolis, France, October 21-25, 2013.
37/ D. Demange and al., Tritium management and antipermeation strategies for three different breeding blanket options
foreseen for the European Power Plant Physics and Technology
Demonstration reactor study, Fusion Engineering and Design, 89,
1219-1222, 2014.

76© IRSN/2018 – Tous droits réservés11/ Annex 1

11/

Summary of the
IRSN report on theconstruction license
application for the
ITER experimental
facility
1)

Context

In 2010, the ITER Organisation (ITER/O) submitted a construction
license application to the French Nuclear Safety Authority(ASN (8)) to build the ITER experimental facility at Cadarache.
This construction license application was accompanied by a filecontaining a “preliminary safety analysis report” outlining theprovisions to manage risks related to the facility and measures tolimit the likelihood of an accident and its impacts. An “impact study”
was also included in the construction license application. This study
presented the health and environmental impacts caused by releasesof radioactive and chemical effluents produced under normal
operation of the facility. The Institute for Radiological Protection
and Nuclear Safety (IRSN) examined both the report and the study,
and presented its conclusions before the Nuclear Safety Advisory
Committee for Laboratories and Plants (GPU(9)) during meetings
held on 30 November 2011 and 7 December 2011. Several members(8)
Autorité de sûreté nucléaire.

(9)Groupe permanent d’experts
pour les laboratoires etusines.

77© IRSN/2018 – All rights reserved(10)Groupe permanent d’experts
pour les réacteurs nucléaires.
(11)
Groupe permanent d’experts
pour les déchets.
(12)Groupe permanent d’experts
pour les équipements souspression nucléaires.
(13)
The thermal output of a
nuclear reactor is expressedin megawatt thermal(MWth).

from the Advisory Committee for Reactors (GPR(10)), the Advisory
Committee for Waste (GPD(11)) and the Advisory Committee for
Nuclear Pressure Equipment (GPESPN(12)) were also in attendanceat these meetings.
The ITER facility is an experimental installation with the aimof demonstrating the feasibility of controlling nuclear fusionenergy during experiments lasting several hundred to several
thousand seconds, generating an output of around 500 MWth(13).
The initial preliminary design studies for the ITER facility were
led by an international team for a generic site. The studies werecompleted in 2001 and French authorities proposed Cadarache as
a potential location for the new site. In 2002, based on the final
documented report prepared by the aforementioned internationalteam and after adapting it to the Cadarache site (the site’s seismic
spectrum, etc.), the French Alternative Energies and Atomic Energy
Commission (CEA) prepared a safety options report. The Institutefor Nuclear Safety and Protection (IPSN, which became IRSN)
assessed these safety options and presented its conclusions to
the GPU, GPR and GPD on 20 November 2002. The safety options
were deemed acceptable and a number of requests were asked tobe taken into account in the preliminary safety report. Cadarache
was chosen as the site for the new ITER facility on 24 June 2005.

2)

The ITER experimental facility

The basic concept for the ITER facility is an experimental tokamak,or magnetic fusion device. The fusion reaction takes place inside
a torus-shaped plasma made from deuterium and tritium. The
plasma is confined by a magnetic field produced by a set of coils.
This contains the plasma inside a sealed toroidal chamber, or“vacuum vessel” preventing the plasma from touching the vessel
walls.
fusion reactions between the deuterium and tritium produce
α particles that transfer their energy to the plasma, and highenergy neutrons (14 MeV) that are slowed down and absorbed in
the structures surrounding the plasma (blanket, divertor, vacuum
vessel), where a cooling system extracts the energy.
In terms of safety, the facility is characterised by the presence ofhigh quantities of tritium, divided between several large buildings.
Besides the building that houses the tokamak, the facility has
a “tritium building”, used to treat products removed from thevacuum vessel, a “hot cell building”, where equipment frominside the vacuum vessel is overhauled, and a “radwaste building”.

78© IRSN/2018 – All rights reserved

11/ Annex 1

3) IRSN opinion on safety and radiation
protection for the facility
To assess the safety and radiation protection of the world’s first
fusion facility classified as a regulated nuclear facility, IRSN set up
a specific organisational structure characterised by the following:•• technical support from a plasma physics expert, neutronic
and activation calculations by the European Organization for
Nuclear Research (CERN(14)) in Geneva, and the signature of acollaboration agreement with the Canadian Nuclear Safety
Commission on tritium confinement and the impacts of tritium
releases on the population;(14)
Centre européen pour larecherche nucléaire.

•• the launch of research and development within the Institute on
specific safety problems raised by a nuclear fusion facility.
The main conclusions drawn from the assessment of safety andradioprotection of the future ITER facility are presented below. It
should be noted that this assessment does not take into accountfeedback that could be drawn from analysis of the accident that
occurred at the Fukushima Daichii plant on 11 March 2011 as
ITER/O was asked by the ASN to submit a report on this feedbackfor 15 September 2012.
Progress of preliminary design studies for the facility
IRSN considers that preliminary designs for equipment in the“tokamak building” were on par at this stage of the facility building
process but that they were incomplete for equipment in the“tritium building”, the “hot cell building” and “radwaste building”.
The operator must therefore submit a new, more comprehensive
version of preliminary designs for the equipment of the othernuclear buildings prior to the assembly phase for “tokamak building”
equipment.
IRSN also considers that the preliminary designs for the automatedtransfer casks used to move highly activated equipment between the“tokamak building” and “hot cell building” should also be finalised
within the same timeframe. A demonstration of the effectivenessof emergency measures planned for any potential failure of thesecasks should also be submitted with the designs.
Finally IRSN deemed that the limitations on the facility’s operating
range, which is essential for demonstrating safety and radiation
protection, were clearly defined by the operator.

79© IRSN/2018 – All rights reserved

The inventory of radioactive substances in the facility
Besides tritium, it should be noted that nuclear reactions between
neutrons from fusion reactions and the environment near theplasma (structures, cooling water, air, etc.) produce activatedsubstances and products. The total inventory of activation
products assessed by the operator was considered adequate
by IRSN. However, IRSN considered that the operator should
continue its activation calculations in order to obtain detailed
inventories by radionuclide for the main equipment. This will
make it possible to assess the risks associated with exposure toionising radiation at each work station, and to characterise theproduced waste more accurately in order to ensure that it can bedisposed of through waste management solutions.
Risk management measures for the facility
IRSN assessed the design of containment systems adopted by
the operator to manage risks involving the spread of radioactive
substances such as tritium. The evaluation found that the design
requires that rooms have relatively substantial leaktightness.
IRSN considered that to ensure the stated leaktightness objective,
the operator will need to measure the leak rate of these premises
when the facility is commissioned and take measures to ensurethat these leaktightness levels are maintained throughout
operation. Furthermore, the majority of rooms are equipped withdetritiation systems that clean the air quite effectively in theevent of accidental dispersion of tritium. IRSN considered that theoperator must ensure that no accident situations cause ambientconditions or chemical emissions that could significantly reduce
the expected efficiency of these detritiation systems.
With regard to risks of exposure to ionising radiation, theassessment found that operation of the ITER facility will involve
relatively short experimental campaigns interspersed with long
maintenance periods where a large number of staff will be onsite.
Maintenance operations on high irradiating components canonly be carried out with robotic equipment and access to theareas in question must be prohibited. IRSN considered that therobust design of access control systems for these prohibitedareas needed to be demonstrated. In addition, despite thepresence of relatively large amounts of tritium gas with a highdiffusion level, the operator has set a target as close as possible
to zero for the internal dose level for workers. IRSN consideredthat demonstrating the achievement of this target would requiredetailed predictive dose assessments at each work station.

80© IRSN/2018 – All rights reserved

11/ Annex 1
IRSN examined risks related to plasma disruptions. Plasma issubject to multiple instabilities, similar to those observed in a
fluid. However, the largest instabilities affect the overall stabilityof the plasma until it collapses. Sudden plasma termination creates
electromagnetic loads in the vacuum vessel, which must be taken
into account in its design. It can also lead to damage on the first wallof the vacuum vessel blanket and divertor. IRSN considered that theoperator needed to continue studies aimed at characterising themost severe instabilities that could affect the plasma of the ITER
reactor.
With regard to internal fire risks, IRSN examined the general fire
protection measures and considered that these risks should befurther analysed in order to demonstrate that the measures are
sufficient. The Institute also considered that the operator should
examine the risk of filtration systems upstream of the detritiation
systems clogging from soot and smoke in the event of a fire.
The risks of internal explosions in the ITER reactor are mainly due
to the presence of hydrogen isotopes (tritium and deuterium) innumerous equipment items. IRSN viewed that these risks cannot
only be managed through prevention measures. Therefore, during
the assessment, the operator supplemented its file by studying
the consequences of an explosion of hydrogen isotopes in rooms
which house equipment used to transfer or store such isotopes.
IRSN considered that for the room housing the cryogenic column
distillation system, the operator should take measures to ensurethat an explosion of the entire inventory of hydrogen isotopes inthese columns does not affect the level of confinement required
for the room.
Accidents in the facility
IRSN examined studies of potential accidents considered by
the operator. The consequences of these events remain limited.
However, IRSN considered that the conservative nature of the
scenarios adopted for some of these accidents, in terms of theirimpacts, should be verified. IRSN also felt that the operator needs
to show that the explosion of hydrogen isotopes or dust, which
could be caused in the event of air or water ingress into the vacuum
vessel, would not affect the confinement of the vacuum vessel andits extensions.

81© IRSN/2018 – All rights reserved

Finally, IRSN viewed that the list of accident situations considered
for the on-site emergency plan design-basis studies needed more
work.
IRSN also identified some unlikely accident situations that cause
loss of the first confinement barrier in the “tokamak building” forwhich the operator needs to demonstrate that the confinementlevel of the second confinement barrier is not affected.
Design of important safety elements
The tokamak is supported by 18 metal columns that are embeddedin the main base slab of the “tokamak building”. The main baseslab is part of the building’s second confinement barrier. IRSN
considered that the operator needs to aim for a more robust
design of the supporting elements than that proposed.
The vacuum vessel is the main element of the first confinement
barrier of the “tokamak building”. The vacuum vessel is a double
walled chamber filled with pressurised water. Due to the elements
around the vacuum vessel (magnetic coils, penetrations, etc.), it
will not be possible to inspect a large portion of the outer wall
during operation. IRSN therefore considered that compensatory
measures needed to be taken for the design and manufacture ofthese uninspectable areas.
Waste and effluents produced by the facility
Radioactive waste from the ITER facility will be activated and/or
contaminated waste, particularly by tritium, for which waste
management solutions have been identified. IRSN felt that theoperator needed to give a precise description of how this waste
would be managed at the facility, of the design of conditioning
for this waste in the facility’s storage buildings, and show thecompatibility of this conditioning with the waste management
solutions. IRSN also considered that work to ensure that thiswaste is integrated into the waste management solutions needed
to be pursued. The Institute also underlined the need to ensurethat these waste management solutions are available whenwaste packages from the ITER facility are ready to be disposed of.
The release of effluents in normal operation was the subject of
optimisation work which IRSN considered acceptable. IRSN and
Canadian Nuclear Safety Commission repeat calculations confirm
that the expected release levels result in very low impacts on theenvironment.

82© IRSN/2018 – All rights reserved

11/ Annex 1

4)

Conclusion

In conclusion, it is IRSN’s opinion that the provisions adopted for the
ITER facility are acceptable on the whole, given the commitments
made by the operator and subject to the response provided forthe points raised by the Advisory Committees. Supplementary
information requested to improve the design of the facility and its
demonstration of safety must be submitted before the start of thetokamak equipment assembly phase.

83© IRSN/2018 – All rights reserved

84© IRSN/2018 – All rights reserved

12/ Annex 2

12/Main past andscheduled milestones
for the ITER facility
Main past milestones:•• 11 and 12 October 1986: decision of the European Atomic
Energy Community (EURATOM), Japan, the Soviet Union andthe United States to jointly pursue design studies for a large
fusion facility, ITER;•• 21 July 1992: signature of the agreement to launch engineering
design of ITER;•• July 2001: finalisation of the ITER conceptual design;•• 28 June 2005: Cadarache chosen as the site for ITER;•• January 2007: work begins on the ITER site;•• 9 November 2012: publication of the construction licence decree
for ITER in the Official Gazette of the French Republic.
Main scheduled milestones:•• 2020: start of tokamak assembly,•• 2026: first plasma with hydrogen,•• 2037: first plasma with deuterium and tritium.

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