Enhancing nuclear safety

Overview ofGeneration IV (Gen IV)
Reactor Designs// Safety and Radiological Protection
Considerations
Published on September 24, 2012

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Reactor Designs

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2/106Overview ofGeneration IV (Gen IV)
Reactor Designs// Safety and Radiological Protection
Considerations
IRSN Report 2012/158
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IRSN
//in brief
The French Institute for Radiological Protection and Nuclear Safety(IRSN) was founded by Act No.2001-398 of May 9, 2001 enacted
through Order No. 2002-254 of February 22, 2002. This Order wasamended on April 7, 2007 to take into account the Act No.2006686 of June 13, 2006 relative on transparency and nuclear safety.
The IRSN is a public establishment that carries out both industrial
and commercial activities. It is jointly supervised by the Ministers
for Defence, Environment, Industry, Research and Health.
IRSN employs over than 1,700 specialists: engineers, researchers,
doctors, agronomists, veterinarians, technicians, experts in nuclear
and radiation risks.
The Institute performs expert assessments and conducts research in
the following fields:
nuclear safety;safety relative to the transportation of radioactive and fissile
materials;



protection of human health and the environment from ionizing
radiation;

protection and control of nuclear materials;



protection of facilities and transports dealing with radioactive
and fissile materials against malicious acts.

Documents de référence
Editions Property of IRSN
31, avenue de la Division Leclerc
92260 Fontenay-aux-Roses
Tel.: + 33 (0)1 58 35 88 88
Website: www.irsn.fr
ISSN 2117-7791
For further information,please contact:
IRSN
Odile Lefèvre
BP 1792262 Fontenay-aux-Roses cedex
Fax: +33 (0)1 58 35 81 39

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Foreword
The Institute for Radiological Protection and Nuclear Safety
develops research programs and conducts studies on nuclear andradiological risks. It is responsible for public service initiatives aimed
at prevention and provides technical support to the publicauthorities in charge of ensuring nuclear safety and security,
together with radiological protection. In fulfilling these various
duties, the Institute is called upon to define its position on certain
scientific and technical issues.
In line with its policy of transparency and its desire to make quality
information available to all partners and stakeholders for use indeveloping their own views, the IRSN publishes "reference
documents”, which present the Institute's position on specificsubjects.
These documents are drafted by IRSN specialists, with the help ofoutside experts if necessary. They then undergo a quality assurance
validation process.
These texts reflect the Institute's position at the time of publication
on its website. It may revise its position in light of scientificprogress, regulatory changes or the need for more in-depth
discussion to satisfy internal requirements or external requests.
This document may be used and quoted freely on condition thatthe source and publication date are mentioned.
We welcome your comments. These may be sent to the address
given in the margin above and should include the reference to therelevant document.

www.irsn.fr

Documents de référence

Jacques Repussard
Director General

IRSN
B.P.1792262 Fontenay-aux-Roses cedex
Fax: + 33 (0)1 58 35 81 39

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Preface
Any future decision to go ahead with the industrial construction ofGeneration IV nuclear power plants in France will be guided by anumber of strategic considerations, among which safety
performance will obviously be foremost.
A number of countries are currently conducting research to develop
Gen IV reactors using various fuels and cooling systems. France, inparticular, has concentrated its research efforts on the constructionof a new sodium-cooled fast reactor. Within this context, IRSN
considered that it should submit these various reactor designs to athorough review from a purely safety perspective. The review was
based on the premise that future reactors should match, if notoutperform, last-generation pressurized water reactors in terms ofsafety performance, in accordance with general WENRA
specifications and taking into account feedback from the Fukushima
accident.
It is clear from this general review that whatever the reactor system
considered, and notwithstanding the intrinsic advantages of eachone, significant technological progress is required before any claim
can be made that expected safety levels have been met.
I would like to thank Mr Jean Couturier, the principal author of thereport, for this work which is both extremely thorough given theavailable data and a pleasure to read in spite of its complex subject
matter.
I hope you find it pleasant and informative reading.

Jacques Repussard
Director General© IRSN/2012 – All rights reserveddoc référence
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6/106//list ofcontributors
Jean Couturier

General Division

Olivier BaudrandNuclear Facility and
System Security Division
Nuclear Facility and
System Security Division
Strategy, Development andPartnerships Department
Nuclear Facility and
System Security Division
Nuclear Facility and
System Security Division

Daniel Blanc
Thierry Bourgois
Evgeny Ivanov
Hervé Bonneville
Giovanni Bruna
Bernard Clément
Georges Hache
Martin Kissane
Renaud Meignen
Daniel Monhardt
Grégory Nicaise

General Division
Nuclear Facility and
System Security Division
Strategy, Development andPartnerships Department
Nuclear Facility and
System Security Division
Nuclear Facility and
System Security Division
Nuclear Facility and
System Security Division
Nuclear Facility and
System Security Division© IRSN/2012 – All rights reserveddoc référence
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Contents
1/ Introduction

9

1/1 Background9

1/2 Purpose

10

1/3 Possibility of taking into consideration
WENRA’s safety objectives for new nuclear power plants

13

1/4 Initial lessons drawn from the events at Fukushima

15

2/ Overview of concepts

17

2/1 Sodium-cooled Fast Reactors (SFR)

17

2/2 High / Very-High Temperature helium-cooled
Reactors (V/HTR)
2/3 Gas-cooled Fast Reactors (GFR)

4258

2/4 Lead-or lead/bismuth-cooled Fast Reactors (LFR)

66

2/5 Molten Salt Reactors (MSR)
2/6 SuperCritical-Water-cooled Reactors (SCWR)

78
86

3/ Overall conclusion

94

4/ References100

5/ What is the SCWR? // the next logical stepin the LWR path toward simplification

103

6/ WENRA statement on safety objectives for
new nuclear power plants // should be reviewed
no later than 2020 (see Reference Document [2])104

List of figures
Figure 1 The Phénix reactor hall —.................................................................................................18
Figure 2 An "innovative" reactor block for a 1,500 MWe SFR — CEA Project —Source: SNETP-SRA 2009 (www.snetp.eu). ...................................................................................21
Figure 3 The various Superphénix systems that contribute to
the "residual heat removal" function. ............................................................................................32© IRSN/2012 – All rights reserveddoc référence
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Figure 4 TRISO fuel particle; .............................................................................................................43

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Figure 5 Pebble (left), compact forming a prismatic block (right)...........................................43
Figure 6 HTR-PM project – an Institute of Nuclear and New Energy Technology(INET) document. ................................................................................................................................47
Figure 7 Schematic diagram of a (2,400 MWth) GFR – CEA data. ...........................................59
Figure 8 Architecture of GFR systems – CEA data.......................................................................59
Figure 9 Schematic diagram of a lead-cooled reactor. ...............................................................66
Figure 10 Schematic diagram of the ELSY primary system. ......................................................67
Figure 11 Schematic diagram of a reactor where the fuel is dissolved in molten salts. .....79
Figure 12 MSFR primary system and core (schematic view) – Source: CNRS. .....................80
Figure 13 On the left is the phase diagram for water (S for solid, L for liquid,
V for vapor) — the critical point and the operating "points" for BWR, PWR and SCWR
reactors are shown; on the right is the temperature-entropy diagram for water................87
Figure 14 Schematic diagram of a pressure-vessel-type SCWR. ..............................................87
Figure 15 CANDU-style pressure-tube-type SCWR, here paired with a hydrogenproduction facility; to the right: a pressure tube and its fuel bundle......................................88
Figure 16. The SCWR in the context of the various light water reactor designs............... 103© IRSN/2012 – All rights reserveddoc référence
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1/

Introduction
1/1

Background
Many energy-futures studies foresee uranium shortages in thetwenty-first century, including some that assume effective energydemand management, drastic limitation of greenhouse-gas
emissions and strong "renewable energy" policies.
In this context, the US Department of Energy (DOE) launched theGeneration IV International Forum (GIF) in the year 2000. Currently,
the GIF has 13 members including France. Its first action was toselect the six "Gen IV" nuclear reactor technologies (concepts)
which are considered the most promising, both in terms ofconserving uranium resources and with respect to the following
criteria:

For an example (in French)
on these subjects, see the
CEA’s press file on theInternet « Quatrième
generation: vers unnucléaire durable » (Fourth
generation: towards
sustainable nuclear power),
dated March 31, 2010.

 reduction in radioactive waste, especially long-lived high-level
waste;
 safety improvements;
 robustness with respect to non proliferation and maliciousacts; economic competitiveness.
The six chosen reactor concepts are as follows: Sodium-cooled Fast Reactors (SFR);
 Gas-cooled Fast Reactors (GFR); Lead-cooled Fast Reactors (LFR); Molten Salt Reactors (MSR);
 High or Very High Temperature Reactors (V/HTR);
 SuperCritical-Water-cooled Reactors (SCWR).

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According to the studies cited above, industrial deployment of thesenew designs could occur towards the middle of the twenty-first
century, following initial operating experience on experimental
reactors, demonstrators or prototypes.
In parallel, France has demonstrated its commitment to developing
fourth-generation reactors, in particular via the target ofcommissioning a fourth-generation prototype reactor in 2020,
which the French President announced in January 2006, in line with
the French Energy-Policy Act 2005-781. This target is closely linked
with objectives concerning the sustainable management ofradioactive materials and waste associated with French Act 2006739 dated June 28, 2006, which specifies that the industrial
prospects for new generations of reactors (including acceleratordriven systems) be assessed in 2012, regarding their ability toseparate and transmute long-lived radioactive isotopes, so that a
prototype facility can be commissioned before December 31, 2020.
During its session on December 20, 2006, the French Atomic Energy
Commission directed French industry towards sodium- or gascooled fast reactors, in particular with regard to the objectives ofconserving uranium resources and reducing waste (via the ability toburn plutonium or to produce it from uranium-238, and the ability
to transmute minor actinides such as americium and curium). In
2009, the work centered on SFRs. This choice seems to be mainly
based on the maturity of the SFR concept, available know-how andconsistency with the strategic national objectives of closed fuelcycles and long-lived-waste management. A prototype SFR project(the Advanced Sodium Technological Reactor for Industrial
Demonstration, ASTRID) is now being designed under CEA
leadership, in partnership with AREVA and EDF.
The timetable for the first phase of the ASTRID project is as follows: 2008-2012: design and research by the Project regarding
baseline options and alternative options; discussions with the
French Nuclear Safety Authority (ASN) and IRSN;
 2012: submittal of a safety orientations report;
 2014: submittal of a safety options report.
1/2

PurposeThis document is a follow-on to the one that IRSN published on thesame subject in 2007, Reference Document [1]. Its purpose is to© IRSN/2012 – All rights reserveddoc référence
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provide an updated overview of specific safety and radiologicalprotection issues for all the reactor concepts adopted by the GIF,
independent of their advantages or disadvantages in terms ofresource optimization or long-lived-waste reduction. In particular,
this new document attempts to bring out the advantages anddisadvantages of each concept in terms of safety, taking into
account the Western European Nuclear Regulators' Association(WENRA) statement concerning "safety objectives for new nuclear
power plants" issued in November 2010, Reference Document [2].
Although the WENRA objectives target third-generation nuclear
reactors (and should be revised before 2020 as stated in [2]), it still
seems useful to take these objectives into consideration given thelack of specific documents for fourth-generation reactors . Itshould be noted that, at the current (early) stage of considerations
on the safety requirements for fourth-generation reactors, it isgenerally stated that fourth-generation reactors must be at least as
safe as third-generation reactors, and should therefore at minimum
comply with the objectives in [2], where these objectives are
relevant. Using an identical framework for each reactor concept,
this summary report provides some general conclusions regarding
their safety and radiological protection issues, inspired by WENRA’ssafety objectives and on the basis of available information. Initiallessons drawn from the events at the Fukushima-Daiichi nuclear
power plant in March 2011 have also been taken into account in
IRSN’s analysis of each reactor concept.
In general, security aspects have not been covered in this document.
Some general considerations have been mentioned in Reference
Document [1] and other very specific comments have been added inthis document.

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Section 5 of International
Nuclear Safety Advisory
Group Report 10 (INSAG 10),
issued by the IAEA, covers
defense-in-depth for futurenuclear power plant projects,
including "advanced reactors",
which correspond toinnovative reactors, but it
dates from 1996. INSAG 12("75-INSAG-3 Rev. 1"), from1999, refers to "future nuclear
power plants" in Section
3.3.10, but with primaryemphasis being placed onwater-cooled nuclear reactors.
In 2008, the GIF issued the
document cited as Reference
Document [3].

The assessment performed includes all knowledge from IRSN's
international studies and research, in particular in the context ofbilateral collaborations (with Russia, China and Japan) and thefollowing projects funded by the European Commission in thecontext of the Seventh EU Framework Programme for Research andTechnological Development (FP7):
 V/HTR:

RAPHAEL (2005-2010) and ARCHER (2010-2013): studies
supporting the development of the V/HTR,EUROPAIRS (2009-2011): studies concerning the couplingof a V/HTR with an industrial facility;© IRSN/2012 – All rights reserveddoc référence
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 GFR:

GCFR (2005-2009) and GoFastR (2010-2013): studies on
the GFR; SFR:


CP-ESFR (2009-2012): studies on the SFR;

 projects related to multiple designs:


SARGEN IV: this is a new project launched for 2011-2013,
with the aim of producing a safety-analysis framework forthe various fourth-generation designs and identifying the
R&D axes to pursue for SFRs, LFRs and GFRs. This project is
led by IRSN.
ASAMPSA2 (2008-2010): development of a European
methodology for Level 2 Probabilistic Safety Analyses(PSA2) regarding releases into the environment in theevent of severe accidents (the project concerns second-,
third- and fourth-generation reactors). This project is led
by IRSN.
THINS (2010-2014): computer-modeling resources tosupport innovative reactors.

The assessment also includes information that IRSN has acquired
through the various working groups in which it participates at theinternational level, for example, the group that performed a study
for the DOE to assess the "source term" (a characterization of
releases into the environment under accident conditions) for SFRs
and an evaluation of the knowledge required to assess the safety of
V/HTRs.
It should be noted that, in the context of the European Sustainable
Nuclear Energy Technology Platform (SNETP, www.snetp.eu), IRSN
has led a number of discussions aiming to specify R&D axes, inparticular for fourth-generation "systems ". Three documents
should be mentioned:

Reactors and associated fuelcycles.

 regarding systems using fast-breeder reactors ----- SFR, LFR,
GFR and ADS (advanced driven systems, i.e. accelerator-driven
reactors) ----- and V/ HTRs for cogeneration : the document"SNETP ----- Strategic Research Agenda, May 2009" ;Combined heat and powerfor industrial processes.

 regarding SFR, GFR and LFR systems: the document "SNETP----- ENSII ----- The European Sustainable Nuclear Industrial
Initiative ----- A contribution to the EU Low Carbon Energy
Policy ----- The Demonstration Programme for Fast NeutronThis document covers Gen II,
Gen III and Gen IV reactors
combined with their fuelcycles.

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Reactors ----- Concept paper ----- October 2010"; this document
uses SFR as a baseline design, with two alternative concepts:
LFR and GFR;
 regarding MSR systems: the document "SNETP ----- Strategic
Research Agenda ----- Annex: Molten Salt Reactor Systems, final
draft, November 2011".
The safety questions raised in the remainder of this document are
identified as issues requiring further R&D in these SNETP
documents. They also cover the need to develop modeling andinstrumentation resources, subjects which are barely mentioned inthis document.
Furthermore, SNETP has issued a document regarding the use ofthorium (for example in MSRs): "SNETP ----- Strategic Research
Agenda ----- Annex: Thorium cycles and Thorium as a nuclear fuelcomponent, January 2011".
1/3

Possibility of taking into consideration
WENRA’s safety objectives for new nuclear
power plants
As mentioned in Section 1/2, WENRA has recently adopted a set ofgeneral safety objectives for new projects for third-generation
nuclear power plants. These objectives are presented in the
document "WENRA Statement on Safety Objectives for New
Nuclear Power Plants", dated November 2010, Reference Document[2]. The seven objectives, which make explicit reference to thedefense-in-depth safety approach, are related to: normal operation, abnormal events and prevention of
accidents (O1);
 accidents without core melt (O2);
 accidents with core melt (O3); independence between all levels of defense-in-depth (O4); safety and security interfaces (O5); radiation protection and waste management (O6); leadership and management for safety (O7).
These objectives are detailed in the Annex 1 to this document.
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 Ensure that "accidents without core melt induce no off-siteradiological impact or only minor radiological impact (inparticular, no necessity of iodine prophylaxis, sheltering nor
evacuation)".
 "Accidents with core melt which would lead to early or large
releases have to be ‘practically eliminated’ ".
 "For accidents with core melt that have not been practicallyeliminated, design provisions have to be taken so that only
limited protective measures in area and time are needed forthe public (no permanent relocation, no need for emergency
evacuation outside the immediate vicinity of the plant,
limited sheltering, no long term restrictions in foodconsumption) and that sufficient time is available toimplement these measures".
In addition to the comments made in Section 1/2, regarding thelimitations inherent to attempting to position the various fourthgeneration concepts with respect to the WENRA objectives [2], thefollowing three points should be noted:

According to the December
2009 WENRA document
cited in Reference
Document [2], andfollowing IAEA Document
NS-G-1.10, an event can beconsidered as "practicallyeliminated" if it isphysically impossible or if it
can be considered with highdegree of confidence to be
extremely unlikely to arise.

 prevention of incidents and accidents (Objective O1) ----which corresponds to the first two levels of defense-in-depth----- is based on various aspects which, in the current state ofresearch and project design, are not well defined (such assystem and loop architectures, equipment reliability, humanmachine interfaces, operating procedures, incident andaccident procedures). Nevertheless, it is possible to assess theconcepts, in particular in terms of complexity, ease orotherwise of operating the reactor, and intrinsic
characteristics that are helpful or less helpful (such as reactor
thermal inertia or operator grace periods);
 Objectives O2 and O3, pertaining to accidents with orwithout "core melt" are not relevant for all concepts, inparticular V/HTR and MSR (for the latter, the fuel may beliquid under normal operation). The concept of a severe
accident , which is associated with core melt for second-and
third-generation reactors, has yet to be defined for certain
fourth-generation reactors;
 technological mastery of a concept, associated withsignificant experience feedback, is an essential condition forsafety; the six fourth-generation concepts are highly disparate
on this level.

In the sense of the idea of"severe accident" in the
IAEA glossary, which
implies "significant coredegradation". This is does
not necessarily involve core
meltdown.

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With regard to this last point, it should be noted that it was possible
to produce an initial specification of the safety objectives for the
EPR in just a few months, in 1993, in a Franco-German context. This
was the case because PWR concept had been subject to deep
analyses by the various parties involved, including IRSN, since thebeginning of the nuclear power program in 1973 and the EPR
project was positioned as an evolution of the reactors currently inoperation or under construction and as a plant series for the earlytwenty-first century. The situation is significantly different for someof the six fourth-generation concepts chosen by the GIF.
It should be noted that Reference Document [3], produced by the
GIF, suggests particularly the following general objectives: "Generation IV nuclear energy systems operations will excel insafety and reliability"
 "Generation IV nuclear energy systems will have a very low
likelihood and degree of reactor core damage".
In conclusion, a (brief) safety analysis framework has been
consistently applied to each of the fourth-generation concepts,
examining the following aspects one after another: normal operation, abnormal events and prevention of
accidents;
 accidents without core melt;
 accidents with core melt;
and, depending on the design: independence between all levels of defense-in-depth; radiological protection;
 waste management.
1/4

Initial lessons drawn from the events at
Fukushima
As far as possible, (highly preliminary) analyses concerning therobustness of the fourth-generation concepts with respect to theevents of March 2011 at the Fukushima-Daiichi nuclear power plant
are given in this document. With reference to the WENRA
document, cited as Reference Document [4], regarding theperformance of "stress tests" on current facilities, the positioning of© IRSN/2012 – All rights reserveddoc référence
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the fourth-generation designs involves the following hazards anddeteriorated operating conditions: earthquakes;
 flooding; loss of heat sink; loss of electrical power; severe-accident management.
It aims to make an initial identification of sensitive areas andrequirements for the various designs with a view to ensuring thethree basic safety functions (controlling reactivity, cooling the coreand containment).
It should be stressed that robustness to hazards largely relies on thereactor’s specific design and installation provisions, which are
independent of the concepts. Nevertheless, certain concepts call forcomments regarding their greater or lesser vulnerability to suchhazards.

Such as the characteristics
of adopted contingencies
and site selection.

With regard to loss of electrical power and heat sinks, robustness
also largely depends on aspects which, at best, are only specified at
the preliminary-design stage of the reactors (such as systemarchitecture, redundancy and diversifications).
Finally, the accident at the Fukushima-Daiichi nuclear power plant
was marked by hydrogen explosions, originating in the oxidation of
Zircaloy fuel cladding by water vapor. Some fourth-generation
reactors may have a "hydrogen risk", which is identified in thisdocument.

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of concepts

2/

Overview of concepts
2/1

Sodium-cooled Fast Reactors (SFR)
2/1/1

Presentation of the concept
An SFR is a fast neutron reactor, i.e. it operates without amoderator. The core is cooled by a molten metal, sodium.
Compared with thermal-spectrum neutrons, fast-spectrum neutrons
more efficiently convert natural uranium (uranium-238), a fertilematerial, into plutonium, a fissile material, which means that an SFR
could be operated in breeder mode, or conversely in burner mode
for increased plutonium consumption. It can also transmute thevery-long-lived actinides (americium, curium and neptunium).
In an SFR, the power density in the core can be around 300 MW/m3(compared with 100 MW/m3 for the current PWRs in the French
fleet). The maximum temperature of the sodium when the reactor
is operating is approximately 550°C, which provides for highthermodynamic efficiencies (of the order of 40%) and a significant
margin with respect to the boiling point of sodium (approximately
900°C).
It is generally planned that mixed plutonium and uranium oxide fuel
(MOX --- UPuO2) be used, although mixed carbide, mixed nitride andeven metal fuels (such as UPuZr) are also under consideration.
Existing and planned SFRs are of two types: "pool-type" reactors(such as Phénix, Superphénix, PFR and CEFR) where the primarysystem is totally contained in a vessel where the reactor coolant
pumps and heat exchangers are immersed in sodium, or "loop-type"reactors (such as Joyo and Monju) where the primary sodium flows
in loops connecting a main vessel with other vessels where the large
components are located.

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18•2/Overview
In the SFRs that have been built or designed to date, the primarysodium does not directly exchange its heat with the water in the
power-generation system. Rather, it exchanges its heat with sodium
in an intermediate system, comprising several loops (three inPhénix, four in Superphénix) fitted with sodium-sodium heatexchangers, and located in the reactor vessel for the "pool-type"design.

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of concepts

GIF analyses highlight the ability of the SFR to close the fuel cycle
and its qualities with regard to the objectives for conserving
uranium resources and reducing radioactive waste.
Furthermore, GIF analyses consider two options for the SFR: anintermediate size (150 to 500 MWe) and a large size (500 to
1,500 MWe), depending on the desired usage (i.e. on fuel cycle
considerations).
2/1/2

Current state of SFR-concept development and outlook
The SFR design benefits from a certain amount of experience, withseveral power reactors having operated: in France, the Phénix reactor (250 MWe - shut down in2009) and the Superphénix reactor (1,240 MWe --- shut downin 1997),

In this document, MWe is
used as an abbreviation formegawatts of electrical
power and MWth formegawatts of thermal
power.
Figure 1
The Phénix reactor hall —CEA-Marcoule-France).

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 in the Great-Britain, the 250 MWe Prototype Fast Reactor(PFR), which operated from 1974 to 1994,

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of concepts

 in former-Soviet-Union countries, the 250 MWe "BN 350"
reactor, located in Kazakhstan and shut down in 1998, and the
550 MWe "BN 600" reactor in Russia, operating since 1980,
 in Japan, the Joyo experimental reactor, whose power reached
140 MWth, and the 280 MWe Monju reactor which restarted
operation in May 2010 after an interruption following asodium fire in 1995. Currently, both reactors are shut down
for an indefinite period for maintenance operations on theirprimary systems following fuel-handling incidents,
 in India, the Fast Breeder Test Reactor (FBTR), a 13 MWeexperimental SFR, has operated since 1985.
Experience feedback on SFR reactors is the subject of a very detailed
IAEA report (Reference Document [5]).
Two low-power experimental SFR reactors should also bementioned:
 The EBR-I reactor, built on the Idaho National Laboratory site,
in the State of Idaho, USA. This was the first nuclear power
plant to be built, and was operated from 1951 to 1962,
producing an electrical power of 200 kW (for a thermal powerof 1.4 MW). It was cooled by a mixture of sodium andpotassium (NaK). In 1955, it suffered a partial meltdown
during a coolant flow test which increased core temperature.
It appears that this was caused by bending of the fuelassemblies (which were laterally clamped at both ends) underthermal expansion and resultant compaction of the fissile
material.
 The Fermi 1 reactor built in the State of Michigan, USA. This
94 MWe sodium-cooled fast reactor operated from 1963 to
1972. In 1966, two of the core’s 105 fuel-assemblies melted,
due to blockage of a subassembly foot by a primary-system
structural component that had detached.
The significant events that have affected power SFRs had various
causes: inappropriate operator responses, design errors, inhibition ofsafety protection systems, inadequate construction requirements
for some equipment items, and difficulties managing
implementation due to the complexity of the industrial engineering.
In addition, it should be remembered that, in 1989 and 1990, thePhénix reactor suffered reactor scrams due to sudden drops in© IRSN/2012 – All rights reserveddoc référence
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power (i.e. reactivity), whose origins remain unexplained and arestill the subject of investigations.

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In France, in the 1980s, following commissioning of the Superphénix
reactor, studies were performed on a project for a 1,500 MWe SFR
reactor (called the RNR 1500). These studies were then pursued in a
European context, with the Great-Britain, Germany and France
partnering on a project for an SFR reactor design --- the European
Fast Reactor (EFR) project. Ultimately, this project was abandoned.
Given this experience, SFR design seems to be in a state of maturity
which means that production of new industrial prototypes can beenvisaged in the medium term (2020-2030). Several projects areunderway, with varying degrees of advancement: the 1,500 MWe Japan Sodium-cooled Fast Reactor (JSFR)
project (using a loop-type design); preliminary design isplanned for 2015;
 construction of the (800 MWe) BN 800 reactor in Russia,
which was suspended after the Chernobyl accident in 1986,
has been restarted: initial criticality is planned for 2012. Russia
is then planning a commercial 1,200 MWe version (BN 1200).
Construction of a pair of BN 800 reactors in China will begin
in 2013 (following agreements signed in 2009);
 completion of an Indian 500 MWe Prototype Fast Breeder
Reactor (PFBR), for which criticality is planned in 2012, andwhich should be followed by several SFRs;
 the Chinese Demonstration Fast Reactor (CDFR) project, withunit power between 600 and 900 MWe, prior to its
commercial step, the 1,000- 1,500 MWe Chinese Commercial
Fast Reactor (CCFR);
 South Korea’s "Korean Advanced LIquid MEtal Reactor"(KALIMER) project, with unit power of 1,200 MWe; the French ASTRID project, with power of 600 MWe; one of
the specifications for this reactor is that, as far as possible, it
should be possible to extrapolate the technical solutions used
to a more powerful reactor.
Note that in 2005, the USA launched design studies on an SFR
project called the Advanced Burner Reactor (ABS); these were
stopped in 2008.
The year 2011 was marked by the connection of the Chinese
Experimental Fast Reactor to the power grid. This

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25 MWe/60 MWth SFR is the fruit of close collaboration between
China and various Russian institutes.

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of concepts

Figure 2
An "innovative" reactor block for
a 1,500 MWe SFR — CEA Project— Source: SNETP-SRA 2009(www.snetp.eu).

2/1/3

Safety aspects specific to the SFR concept
Risks specific to SFRs are mainly due to sodium’s chemical reactivity
with air and water.
Risks associated with the chemical reaction between sodium and air
Sodium, an alkali metal, burns in air. In the event of a leak ofpulverized sodium, the consequences of its combustion are pressure
increases that can be very rapid (a few seconds) and slower
temperature increases (a few minutes), and the production of highly
toxic sodium-oxide aerosols. These oxides react with water in the
air to form sodium hydroxide, which can react with carbon dioxidein the air to form sodium carbonate.
A sodium fire can be aggravated by the heating of the concrete in
the walls of the affected room, which can lead to water desorption
from the concrete, and a hydrogen fire or even an explosion, in
addition to concrete degradation.
It should also be noted that, given the reactivity of sodium with air,
the parts of the primary and intermediate systems not filled withsodium are filled with an inert gas, generally argon (this is the case

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for the volume located above the sodium in the main vessel, thecover-gas space).

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Risks associated with the chemical reaction between sodium andwater
Water reacts with sodium instantly and exothermically, producing
sodium hydroxide and hydrogen. In particular, this reaction canoccur in the steam generators, which are sodium-water heatexchangers. It could lead to large dynamic loads on the affectedloop of the intermediate system and possibly cause a sodiumwater-air reaction in the event of failure of the steam generator’s
external pressure boundary, which could have extremely significant
consequences (such as hydrogen explosion).
The risk of thermodynamic interactions with sodium
Besides a chemical reaction, contact between water and liquidsodium can provoke a thermodynamic interaction, leading tosudden vaporization of the water (vapor explosion), accompaniedby overpressure effects. Also, and in a similar manner to what could
happen on a PWR in the event of fuel melt, contact between
sodium and molten fuel could also cause a thermodynamic
interaction with sudden sodium vaporization.
Risk of a reaction between sodium and MOX fuelIn the event of fuel cladding failure, sodium can come into contact
with the fuel (once the gaseous fission products have been released
into the sodium). With MOX fuel, this creates sodium uranoplutonate compounds with a high coefficient of thermal expansion,
which interact with the cladding and can split it open over a long
length if the reaction between sodium and the fuel is not stopped
sufficiently quickly by removing the faulty subassembly from thereactor.
Risk of coolant freezing
Sodium freezing can occur when the reactor has been shut down fora long time, when the residual heat is insufficient to compensate forheat losses. Clearly, this situation affects reactors which have been
shut down indefinitely (as is currently the case for the Phénix andSuperphénix reactors), but it can also occur during an SFR reactor’s
service life in the event of prolonged outages, such as forinspections, modifications or replacement of large components. The
associated risk is the cracking of structures due to sodium
contraction or expansion during a phase change (mainly during
subsequent melting, given that sodium expands as it melts). For
Phénix and Superphénix, two complementary provisions were

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adopted to prevent sodium freezing: pump operation to heat thesodium and using heat tracing cables to heat components (such as
vessels, pipes and valves).

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of concepts

Furthermore, the viscosity of sodium may increase, or it may evensolidify (in the cooler parts of the reactor), if impurity
concentrations (oxides, hydrides) increase. In particular, this could
lead to fuel-assembly blockage. For this reason, the Phénix andSuperphénix reactors were fitted with a sodium purification system(in particular, to maintain the oxygen concentration at a few ppm),
associated with detection based on measuring sodium flowrate via
an orifice maintained at low temperature ("plugging indicator").
However, use of this detection procedure posed several difficulties
during the incident where air got into the Superphénix primarysystem in 1990 (difficulties interpreting the measurements).
Risk of embrittlement of steels in the presence of sodium
R&D work, such as that reported in Reference Documents [25], [26]and [27], has shown that some steels can become embrittled incontact with liquid sodium at typical SFR operating temperatures.
While the austenitic stainless steels used in the Phénix andSuperphénix reactors (type 316 and 15-15 Ti steels) do not seem tobe (very) susceptible to embrittlement in the presence of sodium,the influence of the concentration of non-metallic impurities in thesodium (oxygen, hydrogen) would seem to require further study(Reference Document [26]). However, harder steels, such as T91 ,were considered for some EFR structures (steam-generator tubes,
intermediate system and fuel assembly hexagonal tubes). The
studies cited in [25] and [26] show that this steel is susceptible toembrittlement, even in contact with pure sodium. Therefore, forfuture SFRs, the risk of the embrittlement of steels in contact withsodium should be carefully studied so that, by choosing appropriatematerials and metallurgy procedures, the risk of sudden failure ofcomponents under accident loads can be prevented. In any case, therisk of embrittlement of steels would appear to substantiate theusefulness of keeping the sodium in SFRs as pure as possible.

The issue has general scope.
It is covered again laterwith regard to LFRs. Also
embrittlement by liquid
metals was used to explain
two accidents that
occurred in 2004 on natural
gas liquefaction plants (theSkikda accident in Algeria
and the Moomba accident
in Australia): these
accidents were ascribed tothe presence of trace
amounts of mercury in themetal components.

This is a martensitic
stainless steel containing
9% chromium and 1%
molybdenum.

2/1/4

Aspects of the safety analysis
First of all, it is important to note that the SFRs that have been built
and operated clearly have a less complex design that that ofpressurized water reactors in terms of their systems architecture.
Operators who have worked on both types of reactor have alsostressed how much simpler SFRs are to operate. In principle, this

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constitutes a positive aspect for safety, with regard to the risk ofhuman errors. However, while the systems architecture of SFRs is
relatively simple, sodium-based technology is clearly less tried andtested than water-based technology, which benefits from over two
centuries of experience with steam supply systems. While
operations of the French Rapsodie and Phénix reactors weregenerally smooth, in particular thanks to the operator (the CEA)
having significant support resources available (in terms of
consultancy, design and R&D) , the move to industrial operations
with Superphénix was less convincing. Incidents occurred on
Superphénix, in particular air ingress into the primary system (in1990) with resulting contamination of the sodium. This led to the
French Ministers responsible for industry and the prevention ofmajor risks and the French Directorate for Nuclear Safety (DSIN) toquestion "the level of technical expertise made available to the
nuclear power plant" and to note "the difficulty of the diagnostics
that the (operating teams) had to perform", "the excessive delay inthe operating team’s diagnosis due (…) to an erroneous diagnosis
given by experts consulted by the operating team", and to conclude
that "the difficulties associated with the sophisticated design of thisreactor remain" (Reference Documents [28] and [29]).

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Although it was unable toexplain the "reactor scramby negative reactivity"
incidents that occurred on
Phénix reactor in 1989 and1990.

Normal operation, abnormal events and prevention of accidentsa) Anomalies that could cause accident situationsFirst of all, it should be noted that SFRs have characteristics thatmake them especially sensitive to certain events: their high core
power density and neutron feedback that is less favorable than on
PWRs.
In particular, with regard to neutron feedback, it should be notedthat there may be a risk of a significant increase in neutron flux inthe event of loss of sodium by boiling, depending on core size. Thisphenomenon is known as the "sodium void effect" and can also
occur in the event of gas bubbles (such as argon from the cover gas
space) being entrained into and through the core. This constitutes a
specific characteristic in comparison with PWRs where water loss
leads to a significant reduction in neutron flux.
Given these characteristics, special attention must be paid at thedesign stage to reducing reactivity insertion risks as far as possible,
in addition to the risks of sodium boiling and gas passing throughthe core.
The Phénix and Superphénix reactors were fitted with "vent valve
assemblies" outside the core to reduce the risk of gas passing into

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the fuel assemblies. With regard to the risk of sodium boiling,following the accident on the Fermi 1 reactor in the USA, specificprovisions have been made in SFRs to prevent fuel-assembly
blockages (and thus prevent the risk of local boiling). Thus, thecircular-cross-section fuel-assembly feet have several lateral holes
around the circumference to ensure that sodium flow is maintainedeven with migrating bodies present. Furthermore, at least for SFRs
operated in France, operating provisions have been taken to ensuresodium purity (in particular, in terms of oxide concentrations,limited to a few ppm by mass), to prevent particles from
penetrating fuel assemblies and blocking them. As seen above, inthe context of industrial operation (of the Superphénix), compliancewith these provisions was less convincing.

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Managing local boiling risks in the event of fuel-assembly blockage
remains a key area for design and R&D. Research is underway toreliably detect any boiling at fuel-assembly outlets, to installneutron detectors in the primary system to detect power variations
and to measure the sodium flowrate in core assemblies.
The question of risks associated with sodium void effects wasreassessed following "reactor scram by negative reactivity (AU/RN)" incidents on the Phénix reactor, in 1989 and 1990, which arestill unexplained. However, it has been demonstrated by tests
performed on the reactor itself that the reactor scram incidents
were not associated with gas ingress into the core.
The RNR 1500 and EFR projects were developed with the idea ofdesigning reactor cores that minimized positive void effects as far as
possible. This objective has also been adopted by the CEA for its SFR
designs and, more specifically, for the ASTRID project. Given thepotential consequences, special attention must be paid to solutions
used to achieve this objective.
Reactivity insertions can also be caused by inadvertent removal ofabsorber rods, passage of a moderator (such as oil) into the core or
fuel compaction (as occurred on the EBR-I reactor in 1955). In thisrespect, specific provisions were made in the past on the Phénix andSuperphénix reactors to reduce the risk of oil ingress (for example
from reactor-coolant pump motors). Furthermore, it should be
noted that operation of SFR reactors in "iso-generating" mode
requires little reactivity reserve, which means that, in the event of
inadvertent removal of absorber rods, the amplitude of possible
reactivity insertions can be significantly limited. With regard to therisk of fuel compaction, following the EBR-I accident mentioned

This operating mode
produces as muchplutonium as the reactor
consumes.

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above, a fuel-assembly design was adopted which has lateralclamping at only one end (at the bottom).

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of conceptsb) Risks associated with the reaction between sodium and MOX fuelMajor design work has been performed to minimize the risk of
cladding failure. In particular, this work involved the selection ofductile austenitic stainless steels for the cladding and checking their
behavior under irradiation in Rapsodie and in Phénix (in particularregarding swelling due to irradiation by fast neutrons). However, it
should be noted that, for the development of SFR designs, the CEAis studying other stainless steels (such as ferritic-martensitic steels)
for fuel cladding and that these steels are less ductile.
For Phénix, the results were highly satisfactory because there were
only 15 cladding failures among the approximately 150,000 fuelpins that were irradiated during the 35 years of operation, bearing
in mind that the reactor core was often loaded with experimental
fuel assemblies using "prototype" cladding materials.
Furthermore, it should be noted that the reactors operated in
France were fitted with a "cladding failure detection andlocalization" system linked to reactor scram, which aimed to limit
the consequences of cladding failure with respect to the risk of fuelsodium interaction. The localization system identified the failed
assembly so that it could be unloaded and thus limit contaminationof the primary sodium by fuel particles (the "clean reactor"
operating principle). The localization system operated correctly
during the cladding failures that occurred on the Phénix reactor.
c) Risks associated with the reaction between sodium and air
Experience feedback means that criteria are now available forselecting materials for equipment that contributes to sodium
containment. However, as stated above, new steels underconsideration for SFRs could have lower ductility and may also be
susceptible to embrittlement on contact with sodium.
Leaktightness of the main vessel of reactors such as Phénix andSuperphénix was permanently monitored using diversified sodiumleakage detection devices ("candle" system and sodium aerosol
detection systems). It should also be stressed that significant work
was performed regarding in-service inspection of the primarysystem pressure boundary on the Phénix reactor, which also aimed
to check that there was no degradation of core support structures.
In addition to all these provisions which aimed to avoid primarysodium leakage, risk management regarding reactions between
primary sodium and air is based on the presence of a safety vessel,© IRSN/2012 – All rights reserveddoc référence
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surrounding the main vessel, and on nitrogen blanketing of theinter-vessel space.

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On the Phénix and Superphénix intermediate systems, leakdetection systems were placed as close as possible to components(such as pipework and tanks). Detection led to the application of aprocedure to empty the corresponding components. However, thelarge number of erratic detections implies that management ofintermediate-system leaks must be improved.
Furthermore, sodium coming into contact with concrete leads todesorption of water from the concrete and hydrogen production by
the sodium-water reaction. The main preventive measure has been
to cover the concrete walls with a metal liner. Also, sodium canreact with certain constituents of concrete (such as silica, SiO2),
which led to the use of special concretes for Superphénix.
d) Risks associated with the reaction between sodium and waterFirst of all, it should be stressed that minimizing the number ofwelds on the steam-generator tubes constitutes an area ofsignificant progress identified in the context of the EFR project.
Suitable selection of materials for the steam-generator tubes is also
crucial for preventing the risk of a sodium-water reaction. Indeed, in
addition to the usual mechanical loads to be considered, the tube
materials should have good resistance to wastage effects(blowtorch effects induced on a tube by a sodium-water reaction
on an adjacent tube). It should also be noted that a modular steamgenerator design (as used on the Phénix reactor) can significantly
minimize the consequences of a sodium-water reaction andmaintain the cooling capacity of the intermediate system
concerned, via suitable isolation of the affected module.
Furthermore, notable improvements were implemented on Phénix
and Superphénix, in particular in terms of the possibility of early
detection of steam-generator tube defects , followed byautomatic isolation and dryout of the steam generator. However,experience feedback from the PFR reactor in the Great-Britain and
the Phénix reactor in France shows that these detection systems are
difficult to operate (the severity of the event on the PFR reactor in
the UK in 1987, which led to the failure of 40 steam-generator
tubes, was associated with the fact that the early detection system
had been disabled following malfunctions). Finally, identification ofa possible failing tube would have presented a difficulty for
Superphénix, given the design of the steam generators on thisreactor (tube bundle comprising superimposed helicoidal layers).

This involved a very
sensitive procedure formeasuring the hydrogenconcentration in thesecondary sodium, whichmeant that small leaks
from the water tubes could
be detected.

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In the context of R&D performed on SFRs in France, three options
have been explored with a view to preventing any sodium-water
reactions:

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 use of a fluid that does not react with water (lead-bismuth
alloy or nitrates) in the intermediate system and with primarysodium ; use of a gas for the power-generation system (the steam
turbine that drives the alternator would be replaced by a gas
turbine); removal of the intermediate system and use of supercritical
CO2 for the power-generation system.
For this third option, it would clearly be necessary to study the riskof gas ingress into the primary system, an event which would cause
system overpressure, could mechanically stress the fuel assemblies
and possibly prevent absorber rod drop.
e) In-service inspections
In-service inspections play a major role in the prevention ofincidents and accidents. In this respect, they constituted a weak
point for earlier SFRs and require significant design and R&D work.
At an early stage, IRSN brought this subject to the attention ofindustry and of the French Nuclear Safety Authority (ASN).
Given the optical opacity of sodium, in-service inspection of certain
equipment items presents specific difficulties. This is the case forreactor internals involved in supporting the core. In the early 1990s,
this difficulty was highlighted for the Phénix reactor in the context
of a CEA request to extend operation of this reactor (following
unexplained reactor scram by negative reactivity). An innovative,
automated device remotely-controlled from outside the main vessel
was quickly developed by the CEA and provided adequate
assurances for several additional operating cycles.
Developments were also made for the Superphénix reactor:
automated devices to inspect the steam-generator tubes and themain vessel from the outside ("Reactor Inspection Modules").
It is clear that, for SFRs, the inspectability of structures is one of the
major subjects on which progress must be made. This applies toboth pool-type reactors and loop-type reactors (according to
information available regarding the Monju loop-type reactor). In
France, the CEA and its partners have undertaken actions in twoareas: possibilities for improving in-service inspection and repair
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appropriate systems for in-service inspections (such as nondestructive testing in a sodium environment).

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It should be noted that IRSN has already drawn the attention of the
CEA and the ASN to the utility of a specific inspection program
during decommissioning of the Phénix reactor, for characterizing
the state of major structures which were not able to benefit fromdirect, detailed inspections during the 35 years of reactor operation.
The results of a specific inspection program such as this would
confirm and possibly supplement the list of types of damage taken
into account and covered during design and operation of Phénix andSuperphénix, as well as the R&D work that has already beenlaunched by the industrial partners of the ASTRID project. IRSN
plans to assess the usefulness of its involvement in R&D work in thelight of the results of this inspection program.
f) Taking hazards into accountWith regard to taking hazards into account, earthquakes have been
subject to special attention. In particular, these involve risks due toleaks from the sodium systems and reactivity insertions via
movement of fuel assemblies. Two points deserve comment: Given the low pressure in the systems, the sodium vessels andpipework are thinner walled compared with PWRs, so thisequipment is more sensitive to seismic loads. Furthermore,
the large quantities of sodium may produce fluid-structureinteraction phenomena via their significant mass and inertia,
and these should be taken into account. Design anddimensioning of SFRs for earthquake conditions is therefore
complex. The RNR 1500 project adopted an aseismic isolation
system using anti-seismic pads below the reactor.
 To overcome reactivity insertion via movement of fuelassemblies, systems for early reactor scram on detection of anearthquake were installed on French SFRs. In particular, thesesystems used "articulated" absorber rods to shut down thereactor even in the event of movement or distortion of fuelassemblies in the core.
The emergency sodium-air heat exchangers located on the upper
outer part of the Superphénix reactor building could constitute asensitive point for certain external hazards (such as external
explosion or aircraft crash).
Accidents without core melt
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systems. A level-1 probabilistic safety analysis (PSA1), aiming toassess the probability of core melt, was undertaken by EDF and
AREVA for Superphénix, but was not fully completed. It was limited
to internal initiator events and at-power reactor states, and did nottake into account possibilities for recovery of failed equipment. The
main results were that the risk of core melt was low (approximately
3.10-6 per year) and that the situation that contributed the most to
the probability of core melt was delayed failure of the main vessel(by creep) in the event of prolonged unavailability of the residual
heat removal systems, bearing in mind that in such a situation thereleases could be large and unfiltered.

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of conceptsa) Controlling reactivity and reactor shutdown
Absorber rod drop is the only procedure for stopping the chain
reaction in an SFR, given that neutron feedback is weaker than in
PWRs and that this feedback can become positive in the event of
sodium void or fuel compaction.
Absorber rod drop (control rods and scram rods) is initiated bymeasurements mainly taken in the reactor itself (such as power,
flowrate through the core or sodium temperature), when themeasured values deviate excessively from the nominal values. In
this respect, as the power density and neutron feedback of SFRs are
somewhat unfavorable with regard to core-melt risks, they requirevery rapid reactor scram when certain significant parameters
become abnormal. This requires that changes in these parameters
be detected, even when they are local, and therefore requires
particularly extensive and effective instrumentation. Furthermore,
control rod drop must be very reliable, which explains therequirement for a scram system that uses redundant and diversified
means, as was already the case for Phénix and Superphénix (inparticular via the supplementary shutdown system, which on Phénix
used an articulated rod that could be inserted into the core even in
the event of fuel-assembly distortion; Superphénix had several suchrods). Three different reactor scram systems were planned for the
EFR, one of which was passive. Passive systems are planned by thedesigners of new fourth-generation SFRs.
b) Residual heat removalFirstly, some design provisions should be noted which aim to
maintain the sodium inventory. The main vessel of the primarysystem is surrounded by a safety vessel. In pool-type reactors suchas Phénix and Superphénix, the space between the two large
vessels was designed such that a leak from the main vessel would
not lead to uncovery of the intermediate heat exchangers, so that

The diameter ofSuperphénix’s main vessel
was 21 m.

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the residual heat could still be removed via the intermediate
loops .
Furthermore, SFRs have significant thermal inertia, due to the large
quantities of sodium in the primary and intermediate systems (forexample, the Superphénix had 3,300 metric tons of sodium in its
primary system and 1,550 metric tons of sodium in all
intermediate-system loops combined). Typically, in the event of lossof the normal residual heat removal systems , the sodium heats
up slowly and temperatures reach their maximum after 2 to 3 days.
This gives operators significant grace periods, during which they
could restore the cooling systems if necessary.
However, prolonged total loss of the residual heat removal systems
would inevitably lead to vessel failure by creep and collapse of thereactor pit, with extremely significant consequences (such emptyingof the reactor, severe core damage and a large sodium fire in the air
charged with fission products and fissile materials).

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For Superphénix, thereactor pit was designed
such that the fuel pins
would not be uncovered in
the event of leakage fromboth vessels.

In such a situation, residual
heat removal is provided byemergency systems, inparticular systems outside
the vessel in the reactor pit.

Also, a design objective should be to "practically eliminate"prolonged total loss of the residual heat removal systems, or tolimit, via design provisions, the consequences of such an event on
primary sodium containment. On this last point, it should be notedthat on the RNR 1500 it was planned that the safety vessel be"anchored" in concrete. However, this type of solution could pose
problems for safety-vessel inspectability.
Concerning the objective of "practically eliminating" prolonged total
loss of residual heat removal, it should be remembered that
redundant and diversified systems were used on the Phénix andSuperphénix reactors and for the RNR 1500 and EFR projects, butthat achieving this objective would require better assurances on theabilities of such systems to operate passively using naturalconvection. For Phénix and Superphénix, the possibility of naturalconvection was demonstrated for certain systems on a standalone
basis, but was never checked with overall certainty for all required
systems working together. In many cases, this possibility was only
assessed by calculations. Additionally, implementation of naturalconvection requires several local interventions, such as gradual
opening of the sodium-air heat exchanger slide valves, which is acomplex operation. Opening them too quickly could lead to sodium
freezing and blockage of the natural convection, while opening thatis too slow or too late could lead to excessive temperatures in thereactor. Therefore, improvements in SFR design remain paramountin this area.

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Finally, on the Phénix and Superphénix reactors, in addition to
implementing natural convection, passive residual heat removal
involved radiative heat transfer from the vessels to water systems
outside the vessels. However, the emissivity of metal structures canchange significantly over time. Therefore, the maintenance overtime of the possibility of radiative heat transfer should be checked
throughout the service life of an SFR.

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Figure 3
The various Superphénix systems
that contribute to the "residual
heat removal" function.

c) The radiological consequences of accidents without core melt
For accidents that do not lead to core melt, the radiologicalconsequences are relatively minor because, due also to adoption ofthe "clean reactor" operating principle, it is mainly fission noble
gases that are released into the systems, given that the ductility of
the austenitic stainless steel used for the fuel pellet cladding
prevents significant cladding cracking or failure. This may need tobe reconsidered if the fuel cladding on new SFRs were to use newsteels, which are currently the subject of R&D (ferritic-martensitic
steels that are less ductile that austenitic steels). Furthermore,referring to the design of the Phénix and Superphénix reactors, thepresence of buffer volumes in the argon cover gas circuit, gives time
for the radioactive decay of fission products, as does a large holding
chamber located upstream of the stack. This chamber is formed ofchicanes to give time for additional radioactive decay of fission
products by acting as a delay line, deferring releases and providing
an opportunity to trap any volatile fission products.

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The risk of releases of toxic chemical aerosols produced in the eventof sodium fires must be taken into account in SFR design. Also, as
fire aggravates the dispersion of radionuclides, the containment
must be designed to minimize the chemical and radiologicalreleases which could result from a primary-sodium fire, as theprimary sodium contains sodium isotopes, tritium and, possibly,
fission products. The dome fitted to the Superphénix was designed
to "contain" a fire involving a metric ton of primary sodium inpulverized form in the event of a severe accident. For new SFR
designs, special attention should be paid to defense-in-depth withregard to primary-sodium fires, especially under accident
conditions.
Buildings and rooms that house intermediate-system loops mustalso be designed to reduce the duration of a sodium fire as much as
possible by limiting the air supply. Outlets were retrospectively
installed into the wall of the Superphénix containment building, atthe level of the "secondary-system galleries" (rooms containing
secondary sodium loops), to reduce the pressure peak resulting froma pulverized-sodium fire. However, this solution is not totally
satisfactory from a safety perspective as it creates containment
bypass: while the sodium in the intermediate-system loops is only
slightly radioactive, it does contain a small quantity of tritium fromthe reactor that has diffused through the tubes of the intermediate
heat exchangers.
IRSN has activity contributed to increasing knowledge regarding
sodium fires and developing related computer-modeling tools. The
experimental area investigated was sufficiently wide to provide
deep understanding of the combustion phenomena andqualification for the models. In collaboration with the CEA, IRSN
produced a summary report on this subject, which also includes
risks associated with the sodium-water interaction.
It should be noted that the current chemical toxicity thresholds for
sodium aerosols are much stricter than those considered during thesafety analyses for the Superphénix reactor. As an indication, the
IDLH value for sodium has been reduced from 250 mg/m3 to
10 mg/m3.
e) Sodium-air-water reactions
Although the steam-generator pressure boundary is designed beresistant to sodium-water reactions, the consequences of its failuremust be examined as part of defense-in-depth with regard to the

Immediately Dangerous to Lifeand Health: with thisconcentration in theatmosphere, a person canremain for 30 minutes withouta gas mask, a durationconsidered sufficient to flee
the polluted area. There are no
statutory values in France.

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risk of a sodium-water-air reaction. Such an event could haveextremely significant consequences (for example, a hydrogen
explosion which could endanger the reactor building) and should
therefore be "practically eliminated".

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Accidents with core melt
Accidents with core melt were taken into account in the designof the Phénix and Superphénix reactors. In both cases, thisconcerned scenarios involving reactivity insertion into the core. In
particular, taking core melt into account led to certain strength
requirements for the primary system (the main reactor vessel andits upper cover slab). For this reason, the primary system wasdesigned to resist a release of mechanical energy (for example,
800 MJ for Superphénix). Two phenomena could lead to release of
mechanical energy:

This choice is partly based on
experience feedback from the
EBR-I and Fermi 1 reactors in
the USA, mentioned above.

 a thermodynamic interaction between molten fuel andprimary sodium, a phenomenon which would lead to theformation of a bubble of sodium vapor which could distort
structures and cause sodium movements during its expansion
in the reactor;
 fuel vaporization, also leading to a bubble expanding in thereactor.
Furthermore, with the aim of ensuring the containment function of
the Superphénix reactor in the event of core melt, in addition to thesafety vessel installed around the main vessel and a metal dome
above the slab (as slab leaktightness could not be ensured during
the design stage), the reactor was fitted with a "core catcher" inside
the main vessel, an option that was adopted on the RNR 1500 and
EFR projects . Use of an internal "core catcher" aims to overcome
the difficulty that would be caused by a large sodium fire in thereactor pit following piercing of the vessel by corium. The Chinese
CEFR reactor is also fitted with such a device. The BN 600 and
Monju reactors do not have a core catcher. In its current design, the
JSFR project does not seem to have one.

Note that no dome wasplanned on the RNR 1500 and
the EFR projects.

For Superphénix, the risks of core melt were the subject of many
analyses and experimental programs, mainly at the Cabri and
Scarabée facilities on the Cadarache site: this involved analyses andtests regarding the various phenomena from partial melt in a fuelpin to the consequences of fuel-assembly melt and full core
meltdown, focusing particularly on recriticality risks for moltenmaterials containing fuel.

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Several more tests were performed on the Cabri reactor and the
Silène facility following the decision taken in 1997 to definitively
shut down Superphénix. These provided significant lessons, and
IRSN and the CEA produced a joint summary report for all thesetests.

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However, a certain number of complex subjects still require deeper
study with the support of appropriate analyses and experimental
programs. In particular, the following possibilities should be
explored: sufficiently early detection of local melt in the core, toprevent a partial melt becoming a general meltdown; prevention of the risk of recriticality by corium relocation (asthis criticality would lead to a "secondary" power excursion, tobe distinguished from the power excursion that initiated coremelt). For this reason, the fuel assemblies on the Japanese
JSFR project may be fitted with a sodium channel, which thedesigner says would eject the corium towards the top of thereactor in the event of a core melt accident, reducing thepower and preventing the formation of a corium pool at thebottom of the reactor (FAIDUS concept --- see Reference
Document [6]). The Japanese have studied this concept in the
EAGLE experiment program and would have obtained
encouraging results;
 maintaining and cooling the corium in the main vessel, withor without an internal "core catcher". Further demonstrationof the effectiveness of such a device is required; collecting and cooling the corium below the reactor vessel inan external "core catcher" (or corium spreading area). This
assumes that the risks associated with sodium pouring into
the reactor pit have been covered.
Furthermore, the designers may propose less conservative
assumptions regarding "vapor explosion" phenomena and theassociated energy, in particular seeking to "practically eliminate" thepossibility of sudden contact between large quantities of sodium
and fuel. The approach and elements used for this subject will must
be carefully examined (experimental support, such as the use of
powerful simulation tools or additional design provisions).
With regard to radioactive releases, it should be stressed that
sodium is able to trap a significant fraction of the fission products
released in the event of fuel melt, such as iodine, barium andtellurium isotopes (but not cesium isotopes or fission gases).

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However, analyses of possible releases have been less complete for
SFRs than for PWRs. A group of international experts working forthe DOE, to which IRSN has contributed, identified several points
requiring additional work:

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 release of radionuclides from fuel due to a sudden reactivity
insertion leading to high temperatures;
 for metal fuels (not planned on ASTRID), carrying along of fueland the "sodium bond " loaded with fission products during
depressurization of a fuel pin with cladding failure; the rates of radionuclide transfer into the sodium via fuelleaching;

A sodium-filled space between
fuel and cladding.

 high-energy interactions between sodium and molten fuel andassociated radionuclide transfers into the sodium; enrichment of the surface of the sodium contained in themain vessel by dissolved or suspended radionuclides; thermal decomposition of sodium iodide in the containmentbuilding;
 reactions of iodine compounds in the containment building toform volatile organic iodides.
These points are detailed in Reference Document [7].
Furthermore, questions pertaining to the possibilities andconsequences of leakages of primary sodium loaded withradioactive materials via the upper cover and into the containmentbuilding during an accident with core melt, particularly in the eventof "vapor explosion" in the core, shall be examined more deeply for
future SFR projects than they were for the Phénix and Superphénix
reactors. Examination of these questions would need to take into
account issues such as the design of the upper cover, whether or
not it has a dome, and the design of any such dome. Among thesequestions, determination of the type and quantities of radioactive
products from core melt that could enter the primary sodium andthe cover gas (generally argon, which fills the cover-gas space above
the primary sodium), and then be ejected via the upper cover,constitutes a subject that merits deeper consideration, as theassessments made in the past for the Superphénix reactor were
essentially theoretical.

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2/1/5

Assessment of the SFR concept with respect to the
Fukushima accident
Preliminary comments
The accident that occurred on the Fukushima-Daiichi nuclear power
plant caused significant damage not only to the reactors but also to
the spent-fuel pools. In Phénix and Superphénix, spent fuel wasstored in sodium in a component called the "spent-fuel storage
barrel ", with a design similar to that of the main vessel of thesereactors. The following comments could also be applied to this typeof component.
The spent-fuel storage barrel solution was not adopted for the EFR
and IRSN has no information on this issue for ASTRID.
Earthquake
Along with loss of electrical power and the heat sink, whose
consequences are assessed below, a powerful earthquake(significantly stronger than the design-basis earthquake) could havevery significant consequences for the safety of SFRs, as it could leadto core distortion that may impair absorber bar drop, an increase incore power by reactivity insertion (due to fuel compaction) or lossof the second containment barrier (the main vessel and its support
structures).

The Superphénix barrel wasabandoned after a leak fromthe barrel’s corresponding
vessel in 1987, due to a crack
in the 15D3 steel. It was
replaced by a spent-fueltransfer drum with argon (fueltransfer station) that was not
used for storage.

The risk of core distortion should lead to designers examining thepossibility of adopting design provisions to improve core rigidity,
bearing in mind that this can cause problems for fuel-assembly
handling, and of developing effective means to stop the chain
reaction even in the event of significant core distortion (forexample, by using articulated absorber rods as on the Phénix andSuperphénix reactors). Finally, it should be noted that a "seismic"
reactor scram system and diversified means of stopping the chain
reaction (articulated rods or other designs) should have "hardened
safety core" robustness, as per IRSN terminology forcomplementary safety assessments.
With regard to the risk of loss of primary sodium containment, it
should be remembered that in reactors such as Phénix andSuperphénix the presence of a second vessel (safety vessel), or evena third (on Phénix), can reduce the risk of significant losses ofprimary sodium and fuel-assembly uncovery if the spaces between
the vessels are suitably designed. The free space in the reactor pit
could also be reduced to a minimum, given that a "solid" reactor-pit© IRSN/2012 – All rights reserveddoc référence
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design should provide stability in the event of severe earthquakes.
Limitations may arise however, given the need for in-service
inspection of these structures. It should also be noted that the
aforementioned equipment items (vessels and reactor pit) requirethat cooling be ensured to maintain their integrity. This function isprovided by cooling-water systems and air-water heat exchangers:
prolonged absence of water in these systems could lead to vessel orslab failure and to a loss of mechanical strength for the reactor pit
concrete.

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Given these issues, at least one of the vessels (main or safety), thereactor pit, one or more cooling-water systems for thesecomponents and the inert-gas systems (to prevent sodium fires)must have "hardened safety-core" robustness.
Furthermore, a severe earthquake could cause multiple sodium leaksin the intermediate system (where this exists) and sodium fires that
could jeopardize important equipment and rooms. The strategy
with regard to such a situation could be to rely on draining thedamaged systems, providing that robust sodium storage tanks areavailable, and that there is at least one robust means of residual
heat removal (for example, the "RUR" emergency cooldown loops
connected to the primary system in the Superphénix design). Thus,
the sodium storage tanks and the emergency residual heat removal
system(s) would have to be designed with "hardened safety core"robustness.
Flooding
Given the reactivity between sodium and water, which leads to arisk of hydrogen explosion, flooding of the rooms containing sodium
equipment would create a serious risk, given the sodium leaks could
occur in the event of a severe earthquake. Under these conditions,other than drastic choices in terms of site and "elevation" of theplatform, implementation of robust "volumetric protection "would be necessary to "practically eliminate" such a situation. Some
components are particularly sensitive, such as the secondary
sodium storage tanks, which are located in the lower part of the
facility and likely to be filled by sodium during reactor outages or in
the event of sodium leaks.
Total loss of electrical power or the heat sink

This involves provisions toensure that rooms that housesafety-related equipment are
isolated from external flooding
risks. This concept was
implemented in the context ofexperience feedback fromflooding on the Blayais site in
December 1999.

For reactors such as Phénix, Superphénix or the RNR 1500 and EFR
projects, loss of electrical power or the heat sink would haveaffected residual heat removal as follows.

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Total loss of electrical power causes the sodium-circulating pumpsof the primary and intermediate systems to stop. In principle,
adequate cooling could be provided by natural convection, both in
the primary system and the cooling loops such as those of the"RUR" emergency cooldown system on the Superphénix, which justuses atmospheric air as a heat sink (sodium-air heat exchangers).
However, as stated in Section 2/1/4, experimental demonstration ofthis has never been performed and accident operation would not be
straightforward, especially if it had to be deployed in an accident
context like that at Fukushima. Also, the possibility of naturalconvection necessarily involves "deployment" of residual heatremoval systems at high elevations, which could make equipment
such as the sodium-air heat exchangers vulnerable to seismic loads.
The leaktightness and functionality of such equipment in the eventof a severe earthquake would constitute design requirements.

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Also, loss of the heat sink may jeopardize the ability of dedicatedsystems to maintain a sufficiently low temperature for the main
vessel, safety vessel, reactor slab and reactor-pit concrete. Robust
provisions are conceivable for emergency systems, such as theinstallation of water-air heat exchangers on the systems listed
above as on Phénix or the possibility of injecting water directly into
these systems from fire-service type connections.
Severe-accident management
Given the events that occurred at the Fukushima nuclear power
plant, the risk of losing the primary sodium inventory should also be
considered. In this respect, while it is inconceivable that sodium
could be added to the primary system by equipment connected inan emergency manner, it seems possible to foresee at the design
stage devices, as in the case of the Superphénix, to reinject any
sodium that may have leaked into the safety vessel(s) back into theprimary system.
Furthermore, the possibility of vessel piercing by corium (includingin the event of failure of an internal "core catcher") would seem toimpose a design-stage requirement for a suitable inert-gas system(of hardened safety-core robustness) for the reactor pit.
In any case, the possibility of rapid core unloading, in particular in
the event of leakage from the main vessel and safety vessel, should
be studied for SFRs, this subject having already been identified
during the analysis of some safety aspects for Phénix andSuperphénix.

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2/1/6

Conclusion
In summary, the following points should be noted for SFRs:
 Sodium reacts violently both with air, which can cause large
overpressures in rooms and systems, and with water, whichleads to a risk of hydrogen explosion.
 Overpressures can also result from contact between liquidsodium and molten materials (such as fuel or steel cladding)
via thermodynamic interaction (sodium "vapor explosion");this could be the case in the event of fuel melt in the core.
 The risk of embrittlement of steels in contact with liquidsodium, and the possible influence that impurities in thesodium may have on this risk, would seem to require further
research, especially if use of steels other than austenitic
stainless steels were considered.
 SFRs have risks of power increases by reactivity insertion, due
to the possibility of positive neutron feedback in the event ofboiling or gas ingress into the core (leading to a void effect),by fuel compaction in the core (for example, due to assembly
distortion) or by relocation of molten materials (such as steel
cladding or fuel). These risks depend on the core design and, inparticular its size (the positive void effect is greater when thecore is larger and when the fuel is loaded with minor
actinides). Possibilities for reducing the positive void effect are
the subject of research by designers.
 The properties of liquid sodium mean that residual heatremoval by natural convection can be considered for thesodium systems, but this remains to be properly
demonstrated, in particular for the possibility of "generalized"
natural convection in all systems at once. It should bevalidated experimentally on possible future SFRs (for example
during start-up tests).
 The large mass of sodium in an SFR gives it significant thermal
inertia, which provides grace periods for operators undercertain conditions, in particular in the event of failure of theresidual heat removal systems once the chain reaction has
stopped.
 Given the high power density in the core of an SFR (threetimes higher than for PWRs), inadvertent increases in power
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blockage with reactor at power, deserve special attention as
they constitute possible initiator events for fuel melt, or evencore meltdown. These initiator events have been widely
studied for various reactors and reactor projects (in France:
the Phénix and Superphénix reactors, and the RNR 1500 and
EFR projects). However, reduction of the associated risks
requires progress in several areas, such as design of the core
and absorber rods, design and performance of monitoring
devices (such as neutron detectors and power sensors) andthe reactor protection system.

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 Prolonged total loss of residual heat removal systems would
have severe consequences. It could lead to delayed failure of
the reactor vessel by creep, which could lead to severe coredamage, with large unfiltered releases into the environment.
Design of future SFRs should aim to "practically eliminate"such a situation.
 With regard to fuel melt, or even reactor core meltdown,
which was taken into account in the design of the reactors
and projects mentioned above, some aspects and phenomenarequire further research: molten materials moving and thepossibilities of a recriticality accident, conditions leading to ahigh-energy "vapor explosion", the design and effectiveness of"core catchers", transfers of radionuclides from the degradedcore into the environment (in particular the ability of sodium
to trap radionuclides).
 The inspectability of structures in a sodium environment
remains a key area for progress in safety for future SFRs andinvolves the accessibility of these structures, inspection
procedures and associated devices.
 With regard to sodium-water reactivity, the design of future
SFRs must aim to "practically eliminate" a generalized
sodium-water-air reaction, as such an event could have very
severe consequences (in particular due to possible hydrogen
explosions).
 Given the toxicity of the aerosols produced in the event of a
sodium fire, SFR designs still need to be improved to further
limit releases into the environment. For example, this involves
the choice of where to install sodium systems, andparticularly steam generators, in the various buildings androoms, as well as the resistance-to-sodium-fires requirements
for the rooms.

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 Given the above, achievement of safety at least as good asthat of third-generation PWRs would apparently require
specific demonstrations and significant advances in certain
areas (such as core physics and instrumentation, thepossibility of natural convection in the sodium systems, andin-service inspections).

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 However, nothing has so far been identified that would makes
this impossible.
 With regard to the events that occurred on the FukushimaDaiichi nuclear power plant, the main point to note is thatrobust provisions would be required to "practically eliminate"
flooding of rooms containing sodium systems, which could
include drastic site-selection criteria for SFRs.
 Several SFRs of significant power have been operated (in theUSA, France, the Great-Britain, former-Soviet-Union countries
and Japan) and another industrial-scale reactor (the BN 600)
is currently operating in Russia. Considerable experience
feedback is available on SFRs.
2/2

High / Very-High Temperature helium-cooled
Reactors (V/HTR)
2/2/1

Presentation of the concept
High Temperature Reactors (HTRs) and Very High Temperature
Reactors (VHTRs), grouped as V/HTRs, are thermal-spectrum
reactors, cooled by circulating helium under pressure (50 to
90 bars). Graphite is used both as a moderator and a neutronreflector. V/HTRs are characterized by significant heating of heliumin the core (approximately 500°C) and a planned mean core-outlet
temperature of 750°C to 850°C for HTRs and over 900°C in the
future for VHTRs. These characteristics mean that a thermodynamic
efficiency of at least 50% can be envisaged, compared with 30% to
35% for current PWRs and 40% for SFRs.
The power density in a V/HTR is around 4 to 10 MW/m3, comparedwith 100 MW/m3 for PWRs and 300 MW/m3 for SFRs.
The most advanced fuel developed for these reactors is called
TRISO (TRistructural ISOtropic). It comes in the form of sphericalparticles about a millimeter in diameter, with a kernel of fissile© IRSN/2012 – All rights reserveddoc référence
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material (uranium-, plutonium- or thorium-based carbides, mixturesof oxides and carbides, or oxides) covered by several coatings(Figure 4) which give the TRISO particle good sealing andmechanical strength up to a temperature of at least 1,600°C.

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Developments have been made in the manufacture of this type ofparticle such that low particle failure rates (approximately 105/particle) have been obtained, although it should be borne in mind
that a reactor can contain 109 particles. For this reason, the helium
is purified during reactor operation.
For new fuel, work collated in the IAEA report in Reference
Document [8] gives a value of 2,000°C as the threshold above
which significant breakdown of the particle coatings was observed,
leading to large releases of fission products. However, otherdegradation mechanisms come into play above 1,700°C. For thisreason, the value of 1,600°C has been adopted by V/HTR designers
as the criterion to be met, including under accident conditions. The
melting point of the fuel kernel itself is approximately 2,700°C.
Figure 4
TRISO fuel particle;

The TRISO particles are distributed in a graphite matrix to form thefuel elements. These are in the form of pebbles or compacts(Figure 5). The former are used in pebble-bed reactors.
Figure 5Pebble (left), compact forming a
prismatic block (right).

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The density of TRISO particles in a fuel element can be adjusted
over a wide range, depending on the type of fuel and the power
density sought. This is a key point in understanding the flexibility ofthis design and its robustness with regard to loss of coolant
accidents (LOCAs), as explained in Section 2/2/4 below.

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Finally, it should be noted that the V/HTRs currently underconsideration use enriched uranium fuel, without later reprocessingof irradiated particles. However, several types of fuel kernel havebeen studied and irradiated (plutonium and thorium) and inert
matrices (such as magnesium oxide) could also be used in the placeof uranium-238 to form transmutation targets. Techniques forreprocessing the particles and graphite have also been explored at
the laboratory scale. The V/HTR concept is therefore not necessarily
only associated with a once-through fuel cycle.
2/2/2

Current state of V/HTR-concept development andoutlook
The first HTRs were developed in the 1960s to 1980s, particularly in
the USA, the Great-Britain and Germany.
Two HTRs have been operated in the USA, the first (a 200 MWthreactor) on the Peach Bottom site from 1966 to 1974 and thesecond (a 330 MWe, 842 MWth reactor) at Fort St Vrain from 1977
to 1992.
In Germany, a (15 MWe, 40 MWth) experimental pebble-bed
reactor (the AVR reactor) operated from 1966 to 1988. A 300 MWe
prototype power reactor (the Thorium High-Temperature Reactor,
THTR 300) operated from 1985 to 1988. It should be noted that the
AVR operated using two types of fuel, one made of slightly-enricheduranium and another made of a uranium-thorium mixture. The
THTR used thorium fuel.
Until 1988, Germany was involved in industrial projects: the HTRModul, PNP-200 and PNP-500 projects.

Thorium-232 is a fertile
isotope, which produces thefissile isotope uranium-233 in
the reactor

In Great-Britain, a 20 MWth experimental reactor (the DRAGON
reactor) operated from 1965 to 1976.
Currently, two experimental reactors are operating: in China, the 10 MWth HTR-10 reactor;
 in Japan, the 30 MWth High Temperature Test Reactor(HTTR), which has achieved a reactor-outlet coolant
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this reactor (the cooling was stopped), and the results of thesetests are being analyzed, by IRSN among others.

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Significant events that affected the HTRs which have operated, andthat have been reported in the literature, are as follows: On the Fort Saint Vrain reactor, neutron-power instability
problems were caused by graphite-block movements in thecore and water ingress into the primary system at the motordriven blowers; the NRC report in Reference Document [32]provides a comprehensive summary of the 279 events
catalogued on this reactor.
 On the THTR, there were failures of insulation-component
fastenings on core outlet pipework.
In terms of radiological protection, the available information (inparticular Reference Documents [33], [34] and [35]) generally
reports low collective doses:
 For the AVR, except at the beginning of operations, which wasmarked by collective doses between 1 and 1.25 person-Sv/yr
due to the need for intensive maintenance, the collective dose
gradually declined to approximately 0.2 person-Sv/yr by theend of operations.
 For the THTR, the collective dose was approximately
0.1 person- Sv/yr at the end of operations.
 For the Fort Saint Vrain reactor, Reference Document [35]reports collective doses not exceeding 0.03 person-Sv/yr overthe period 1974 to 1978.
These values can be compared with 0.7 person-Sv/yr for operating
the 900 MWe PWRs in the current French fleet. However, theymust be used with caution, as the volume of maintenance(including in-service inspections) used on these high-temperature
reactors is unknown, and it seems difficult to extrapolate the values
to future V/HTR power reactors.
There have been few industrial projects. Nevertheless, in terms of
R&D, the EU has backed several projects and the IAEA various
Coordinated Research Projects (CRPs).
The withdrawal of Germany has been a significant brake ondevelopment of the concept in Europe, despite the fact that Poland
is examining HTRs with interest (Polish universities and industry
participated in the EU’s EUROPAIRS project), with a view tolaunching a nuclear power industry to replace its coal-fired power
stations.

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In the USA, the Next Generation Nuclear Plant (NGNP ) project
for a combined-heat-and-power reactor paired with an industrial
facility, led by the DOE and currently at the beginning of the preconceptual design phase, is blocked in negotiations with the Federal
Government on the question of public/private funding issues, to thepoint that the NRC has decided to halt certification analyses.
Nevertheless, R&D activities commissioned by the DOE do havefunding for 2012. It should be noted that AREVA developed anindustrial VHTR project, with a thermal power of approximately 600
MW, to be able to respond to a possible DOE invitation to tender(the ANTARES project, which was the subject of some technicalcorrespondence with IRSN).

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http://www.nextgeneration
nuclearplant.com/

In South Africa, construction of the Pebble Bed Modular Reactor(PBMR) planned for Koeberg was finally abandoned in the 2010, due
to a lack of clients and investors.
Finally, China is the only country to have developed a prototype
industrial reactor, the High-Temperature-Reactor Pebble-bed
Modules (HTR-PM) project, comprising two modules each of
approximately 250 MW thermal power, which together can produce
210 MW of electrical power (see Figure 6 and table below). The
option retained is very similar to the most recent German projects
and is based on experience acquired with the HTR-10 reactor overmore than 10 years. According to Reference Document [9], thereactor should be commissioned in 2013, but the main components(reactor vessel and steam generators) are currently still under
manufacture.
With regard to site-specific aspects, it should be noted that, given
the high temperatures aimed for, the V/HTR concept is highly
suitable for combined heat and power generation paired withindustrial facilities. Various safety assessments have been
performed regarding the pairing of a V/HTR with an industrial
facility: for example, the NRC-organized assessment for the NGNP
project (Reference Document [30]) and work performed in thecontext of the EUROPAIRS project (Reference Document [10]). The
NRC-organized assessment highlighted specific risks such as, in theevent of a leak from the industrial facility, the risk of ground-level
dispersion of a layer of dense (cold) gases which could explode.
Work performed in the context of the EUROPAIRS project, led by
IRSN, has shown the need for consistency in combining the safety
approaches adopted for the reactors with those adopted for theindustrial facilities, but did not identify any incompatibilities. It
would seem possible to pair a V/HTR with an industrial facility while
complying with acceptable distancing conditions.

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Figure 6
HTR-PM project – an Institute of
Nuclear and New Energy
Technology (INET) document.

2/2/3

Safety aspects specific to the V/HTR concept
The risks specific to a V/HTR are mainly associated with thepresence of graphite in the reactor.
Air ingress into the reactor could therefore lead to oxidation ofgraphite, or even a graphite fire . Water ingress could lead tocorrosion of graphite and the production of flammable gases,
together with increased core reactivity.

A graphite fire is selfsustaining combustion, i.e.
rapid oxidation at hightemperature which maintains
itself without needing external
heat.

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The risk of a graphite fire due to oxidation following air ingress has
been studied in detail, in particular following the accidents at
Windscale and Chernobyl. It should be noted that in both cases anadditional source of heat (other than oxidation) initiated andmaintained the graphite fire: the Wigner effect at Windscale andthe fuel’s residual heat at Chernobyl which was combined withsignification break-up of the graphite following the reactor’s
explosion. It would seem difficult for these conditions to combine ina V/HTR. Indeed, the energy stored by the Wigner effect isnegligible at the temperatures reached by the irradiated graphite ina V/HTR (> 350°C). Furthermore, in a V/HTR, to maintain graphiteoxidation, there would need to be air circulation strong enough tobring sufficient oxygen but not so strong as to significantly cool thecore. Finally, the oxidation of graphite is strongly influenced by thepresence of impurities, which favor early oxidation. Oxidation ofgraphites like those involved in the Windscale and Chernobyl
accidents becomes significant at considerably lower temperatures
than for the high-purity graphites envisaged for future V/HTRs.

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Graphite stores energy during
irradiation at low temperature(below 115°C). If, after
cooling, the graphite isreheated (above 80°C), it
suddenly releases the stored
energy (Wigner effect), which
can lead to a graphite fire, as
occurred during the Windscale
accident.

In summary, it may be possible to exclude graphite fires for V/HTRs,
although further research is required. It should be noted that, for the
PNP reactor projects, German experts considered that the risk ofgraphite fires could be excluded in the event of a LOCA, even if theblowers continued to be operated (Reference Document [11]).
Furthermore, analyses show that, after several hours, a single-ended
pipe break would lead to a very thin layer of oxidized graphite(various NACOK tests performed by the Jülich research center).
The risk of corrosion of graphite is smaller in the event of water
ingress than in the event of air ingress, in particular because thereaction with water is endothermic.
With regard to the risk associated with gases produced in the eventof water or air ingress, it should be mentioned that, in the context
of the EU’s "Advanced high-temperature Reactors for Cogenerationof Heat and Electricity R&D" (ARCHER) project, IRSN is contributing
to research into fire and explosion risks. According to German
designers, this risk can be excluded if the power of the reactor islimited to 200-250 MWth (which is relevant for pebble-bed core
reactors).
Air and water ingress transients have been studied in detail for thereactors built in Germany, Japan and China. IAEA published several
summary documents on these subjects in the 1980s and 1990s (seethe 1993 Report in Reference Document [11]).

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2/2/4

Aspects of the safety analysis
Normal operation, abnormal events and prevention of accidentsa) Prevention of significant core damage
Clearly, detailed experience feedback would need to be gathered forconsideration of new V/HTRs. IRSN has no specific information on
the solutions envisaged with regard to certain events, in particular
the graphite-block movements in the core that affected the Fort
Saint Vrain reactor.
It should be noted that, in the case of V/HTRs, the favorable
physical characteristics of the fuel and the reactor in general play anessential role in preventing aggravation of incidents or accidents(characteristics such as feedback due to the Doppler Effect andfeedback due to the graphite moderator). Thus, in the event of lossof the cooling systems, the strong thermal feedback due to graphite
virtually stops the nuclear reaction, even without the interventionof a scram system.
Furthermore, adjustment of the moderating ratio (by altering thedensity of fissile matter) can be used to optimize reactor behavior in
the event of water ingress, such that the increase in power could becompensated for by thermal feedback alone.
It should also be noted that the operating mode for pebble-bed
reactors does not require reactivity reserve, which means that theneutronic weight of the control rods can be limited, thus limiting
the amplitude of the reactivity insertion in the event of inadvertent
withdrawal of these rods.
b) Risk of oxidation of graphite due to air ingress
In the event of air ingress into the core, as described in Section
2/2/3, it is the characteristics of the system (limited air circulation)and the graphite (limited impurities) that prevent or limit oxidationof graphite and deterioration of the TRISO particles. Indeed, in
pebble-bed or graphite-block designs, the 5-to-10-mm-thickgraphite layer, which surrounds the area loaded with TRISO
particles, protects these particles which, without this protection,
could be vulnerable to external stresses (such as thermal transients).
c) Risk of corrosion of graphite due to water ingress
Currently, prevention of water ingress remains an issue underdiscussion, but it should be noted that developments in industrial
power generation mean that it is now possible to use entirely gas-

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based systems. Nevertheless, sources of potential water ingress,
such as the blower cooling systems, have yet to be examined.

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of conceptsd) Materials
While industry has materials that are tried and tested fortemperatures characteristic of HTRs, R&D is underway on materials
for structures and components that may be used for VHTRs (inparticular, alloys containing chromium), with regard to issues suchas their high-temperature strength or weldabilty.
e) In-service inspections
IRSN does not have experience feedback on in-service inspectionsfor HTRs.
In terms of radiological protection, primary helium purification
during reactor operation is a favorable aspect of the design, but risks
associated with the deposition of carbon dust loaded with fission
products in the primary system should be taken into account; this
aspect is mention again farther.
Furthermore, in some projects, it is considered possible to exclude"cross vessel failure", i.e. a break in the large-diameter pipe section
linking the reactor to the intermediate heat exchanger, which has
countercurrent flow of "hot" and "cold" helium. However, theprovisions for design and operation of such a component, inparticular with regard to in-service inspections, must be specified
before a position can be taken on this subject.
Accidents without core melta) Prevention of the core-melt risk
According to available studies, the design specifications used in thecontext of GIF work regarding the (low) power density, and theability to remove the residual heat, mean that core melt can be"practically eliminated" for a V/HTR, in particular meltdown of thefuel itself (the kernel of the TRISO particles ----- with a melting pointof approximately 2,700°C) with graphite degradation (as graphite
does not melt).
Thus, as stated above, in the event of loss of cooling (the blowers
stopping), the strong thermal feedback due to the graphite virtuallystops the chain reaction, without the intervention of the reactor
scram system. The residual heat is then removed by conductive andradiative heat transfer through the reactor vessel wall. However, forresidual heat removal by radiative heat transfer to be effective, theconcrete of the reactor pit must be maintained at a sufficiently low
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operate passively using natural convection. According to designers’
analyses, the temperature reached by the fuel should then remain
below the 1,600°C criterion. Such behavior is only possible thanks
to the thermal inertia of the graphite (residual heat storage) and thelow power density in the core, combined with the possibility ofresidual heat removal by natural mechanisms (radiative heattransfer from the vessels).

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However, IRSN considers that failure of the reactor pit coolingsystem could lead to core temperature levels exceeding 1,600°C, atwhich significant damage of TRISO particles and significant releasesof fission products would be possible. Therefore, the reliability of
the reactor pit cooling system of a V/HTR would seem to be a key
issue for the safety demonstration.
It should be noted that, in addition to the favorable neutronic
characteristics of V/HTRs, designers also plan a passive reactor
shutdown system, which would trip in the event of abnormal
heating of the helium (using a Curie temperature magnetic system).
Finally, as stated above, prevention of significant damage to TRISO
particles, via degradation of their protective graphite layer, shall betaken into account for the design of V/HTRs and explicitly coveredin the safety demonstration.
b) Limiting the consequences of accidents without meltdown orsevere core damage
The design-basis accident that could lead to radioactive releases
into the environment is a loss of coolant accident (LOCA) leading todepressurization of the primary system (in a few minutes). Three
specific points should be noted:
 The temperature of the fuel must not exceed the criterion of
1,600°C to ensure a very low level of radioactive releases
from the fuel particles. This should be checked taking into
account air ingress following depressurization, which could
contribute to core heating via oxidation of graphite, in
addition to the residual heat (see Sections 2/2/3 and 2/2/4-cwith regard to preventing a graphite fire).
 Once the primary system is depressurized, there is nothing todrive later dispersal of fission products into the reactor
containment (there is no safety injection of coolant in a
V/HTR).
 As helium is non-condensable, it would seem unrealistic toconfine it in the reactor containment when the primarysystem depressurizes: the gas will be released into the© IRSN/2012 – All rights reserveddoc référence
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environment via filters, as the core’s radioactive content has
not been released from the fuel at this stage.
With regard to this last point, the short-term radioactive content
released into the containment in the event of a LOCA would mainlycontain fission and activation products associated with carbon dust
present in the primary system under normal operation, which will
be partly resuspended. The concentration of contaminants flowing
in the helium under normal operation will be very low, as
determined by the purification system. In the IRSN analysis
performed in the context of its participation in the EU’s "ReActor
for Process heat, Hydrogen And ELectricity generation" (RAPHAEL)
project (Reference Document [12]), it was shown that while theelements associated with dusts are relatively well identified, thequantification of these elements is complex and has relied up until
now on the mainly empirical analyses produced on this subject.
Furthermore, the carbon-dust fraction potentially released into theatmosphere is currently difficult to assess, although the sensitivityof dust transfer mechanisms to the size of the break considered
may be stressed. The very low dose values ----- of the order of ten
microsieverts ----- given for HTR-Modul are associated with a branchline pipe break (see Reference Document [13]). If larger breaks
cannot be excluded, other phenomena could come into play that
could lead to larger releases (possibility of a cliff-edge effect on thequantity of dust resuspended). In any event, "containment
management" (based on the principles of static confinement anddynamic confinement ----- by the ventilation and filtration systems)
is yet to be specified by V/HTR developers.
Accidents with core melt
While it seems that core melt in the strict sense (melting of the fuelkernels of TRISO particles, i.e. reaching a temperature of 2,700°C)
can be "practically eliminated" for V/HTRs for the reasons given
above, it is still necessary to assess the risk of significant fueldamage in the core, specifically deterioration of the sealing of therefractory layers coating the fuel kernels, which could cause large
releases of fission products into the primary system or thecontainment. This risk must be assessed on the basis ofdeterministic and probabilistic safety analyses, taking uncertainties
into account. In other words, there must be consideration of thepossibilities of the fuel particles reaching temperatures above
1,600°C (see Section 2/2/1). It would seem that for V/HTRs, thislarge release of fission products in the core can be considered as asevere accident (as defined in Footnote 6), analogous to core meltfor other designs. It should be noted that it is difficult to assess the

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For this reactor, the bounding
accident scenario corresponds
to a break in a 65-mmdiameter pipe (branch line on
the primary system). This
scenario was also envisaged by
AREVA for its ANTARES
project. The consistency ofthis choice with the types ofbreak used for the safety
demonstrations on SFRs or
EPRs merits reassessment.

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resultant releases into the environment, in the absence of a clear
strategy to limit these consequences on the part of the designers(i.e. a "containment management" strategy). However, it can benoted that, for the example of a LOCA as an initiator event, releaseof radioactive products in the core would not occur until after
system depressurization, such that there would no longer be aneffective drive mechanism to transfer radioactive products into thecontainment.

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In any event, the absence of any true core meltdown risk for
V/HTRs means that the risks associated with the flow of moltenmaterials are excluded, in particular the risk of basemat piercing andresultant soil contamination.
IRSN has no knowledge of an accident scenario leading tosignificant releases adopted in the safety demonstration for reactorsthat have been built, other than the Japanese HTTR. For this reactor,
a penalizing bounding accident scenario was used at the request
Japanese Safety Authority ("regulatory approach"): this involves
partial oxidation of the fuel with release of a fission-productcontent fraction similar to that used for core melt accidents on
PWRs. Accident management is partly based on fission-productfiltration.
For the HTR-Modul project, which was abandoned, the hypotheticalbounding accidence scenario corresponded to a prolonged loss of
power supplies, leading to the loss of reactor pit cooling. Use of amobile system to inject water into the reactor pit cooling system
was planned.
Radiological protection and waste management
At the current design stage, a few comments can be made regarding
radiological protection and waste management.
The sealing characteristics of the fuel, and helium purification,
should facilitate worker protection. However, attention should begiven to analyzing the issue of areas in which primary-system
contamination may be concentrated (blowers and heat exchangers).
In proportion to the electrical power produced, a V/HTR operatingwith enriched uranium produces similar types and quantities ofwaste to a PWR. Reactor projects operating at high burn-up (over200 gigawatt-days per metric ton of UO2) have been studied. This
type of operation would have the effect of reducing the quantity of
waste produced by V/HTRs and would mean that actinides from
PWRs could be incinerated.

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With regard to graphite, progress in its manufacture (such as
reduced impurities and elimination of chlorine for graphite
purification meaning less chlorine-36 produced under irradiation)means that simpler reprocessing following irradiation can beenvisaged than for Gas Cooled Reactors. In particular, the EU has
funded the CARBOWASTE project for research into the "Treatment
and Disposal of Irradiated Graphite and other Carbonaceous Waste".

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of concepts

IRSN made this suggestion to
AREVA during discussions on
the ANTARES project.

However, the graphite in the various components (graphite
compacts or pebbles, graphite blocks used as moderators or
reflectors, cladding containing silicon carbide and carbon, matrices
containing nitrogen compounds, and carbide or nitride fuels) would
produce carbon-14 (with a half-life of 5,730 years) by neutronic
activation in quantities much greater than those produced byreactors currently in operation. This radionuclide is relatively mobile
in natural environments. This aspect should be taken into
consideration for the safety and radiological protection of operatorsin the upstream fuel cycle, as well as for decommissioning andwaste disposal.
2/2/5

Assessment of the concept with respect to the
Fukushima accident
In the following sections, some qualitative information is given
regarding the presumed behavior of a V/HTR (like HTR-Modul or
ANTARES) in the case of events such as those that occurred on the
Fukushima-Daiichi nuclear power plant. Possible means for the
mitigation and management of accidents are also mentioned.
Earthquakes and floods
It is hard to assess the vulnerability of the V/HTR concept to asevere earthquake (stronger than the design-basis earthquake), as
the general design of such reactor projects (in terms of buildings,
reactor containment and equipment) is not sufficiently advanced.
However, one relatively intrinsic aspect of the V/HTR can be
mentioned: an earthquake would have very little effect on thereactivity of a core using blocks. This type of core is not very
compactable due to the solid nature of the moderator and theprism stacking. For a pebble-bed core, the effects of compaction
should be analyzed to demonstrate that they would becompensated for by neutron feedbacks. Furthermore, a gravity
absorber-ball injection system, whose effectiveness would be little
altered by core distortion, is planned in some V/HTR projects. It© IRSN/2012 – All rights reserveddoc référence
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should also be noted that in the event of loss of normal cooling,
reactor power would be naturally reduced by neutron feedback.

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Residual heat removal would not be affected by changes to thegeometry or by significant core disorder (such as cracks andblockages) as the residual heat is mainly removed by conductive
and radiative heat transfer through the reactor vessel wall.
Finally, flooding of the site and rooms of a V/HTR should not cause
major difficulties for residual heat removal. It should be noted that,in some V/HTR projects, the power-generation system uses anhelium-air turbine.
Nevertheless, the risks of air and water ingress into the primarysystem in the event of site flooding should be analyzed, bearing in
mind that, in the event of a severe earthquake, one or more breaks
could occur on this system.
Total loss of power or the heat sink
Consideration should be given to the aggravating factor of loss ofthe normal means of cooling, leading to loss of the cooling systems
outside the reactor vessel. In such a situation, it is useful toestimate the time available before the maximum acceptabletemperatures for the vessel and the surrounding structures (inparticular the reactor pit) would be reached. In principle, given thethermal inertia of graphite, this would be several hours. Thissituation is to be studied in the context of safety tests planned for2012 on the Japanese HTTR reactor (the OECD’s Loss of Forced
Cooling project, LOFR). It should be remembered that, for somedesigns, natural convection based systems could be used to cool thereactor pit. Furthermore, the ground around the reactor pit could
also play the role of a heat sink (for a buried configuration), but inthis case partial destruction of the reactor vessel is not excluded(depending on the residual heat to be removed). Simple emergency
means could also be planned to ensure reactor pit cooling by direct
spraying or by water injection via fire-service type connections
planned in the design.
Severe-accident management
The question of the definition of a severe accident for a V/HTR has
already been covered.
The results of the available safety analyses show that a LOCA does
not lead to a requirement for coolant injection, as the coretemperature would stabilize at around 1,600°C. However, thissituation would be dependent on effective reactor-pit cooling, as
mentioned above.

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As stated previously, "containment management" in the event of asevere accident is yet to be defined.

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Loss of cooling for the fuel storage areas
Given the low power density of the fuel assemblies used (a few tensof kW/m3), natural convection cooling of stored fuel seems to be arobust solution.
2/2/6

Conclusion
In summary, the following points should be noted for V/HTRs:
 V/HTRs use as coolant a gas such as helium, which does notreact with air or water.
 V/HTRs use a specific fuel: this involves a kernel of fissile
material coated with several layers of refractory materials
forming a millimeter-sized particle. This fuel, called TRISO,
can withstand temperatures of up to 1,600°C. Furthermore,
given the (low but non-zero) rate of particle sealing failure
and the very large numbers of these particles in the core,continuous helium purification is planned. TRISO fuel has
already been tried and tested in past HTRs and is currently
used in two small experimental reactors (HTR-10 in China andHTTR in Japan).
 V/HTRs use large quantities of graphite in the core as
moderators and neutron reflectors. Graphite also plays a rolein fuel-particle protection, for example with regard to thermaltransients. In contact with air, graphite can oxidize at varying
rates depending on its purity. Special attention must be paid
to the possibilities and consequences of air ingress into thereactor, both to prevent degradation of the graphite
protecting TRISO particles and because a graphite fire could
lead to significant radioactive releases. Thus, the possibilities
and consequences of air ingress into the core constitute a key
aspect of the V/HTR safety demonstration. According toavailable analyses, a graphite fire could be made highly
improbable, but clearly this would need to be confirmed on
the basis of such aspects as the detailed design options
adopted on a V/HTR and the quality of the graphite used.
 The large mass of graphite in the core gives the system highthermal inertia such that, under certain accident conditions,operators have grace periods to repair or restore failed
equipment.

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 Furthermore, with suitable reactor and core design (especiallyby limiting the overall power of the reactor and the power
density in the core), the system can present intrinsicallyfavorable characteristics in terms of safety. In particular, thechain reaction can be virtually stopped by neutron feedbacks
and the residual heat removed passively (via conduction
inside the vessel and radiative heat transfer outside the vessel)
without the fuel reaching temperatures that could degrade
the sealing of its cladding layers. These intrinsic characteristics
can also limit the reactivity consequences of water ingress
into the core.

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 However, in the various projects of which the IRSN is aware,
the systems specifications (in terms of aspects such asarchitecture, redundancy, diversification and the use of activeor passive systems) and the safety demonstration seem to beat an early stage and in need of more work. Although fuelmelt as such seems highly improbable, which puts the V/HTR
concept in a favorable position, IRSN has questions regarding
possible scenarios for significant damage to the sealing of alarge number of particles in the core, as well as the possible
releases in the event of a LOCA, taking into account thepossibilities for total failure of the cooling systems.
Furthermore, "containment management" is yet to be
specified. R&D is underway to identify the radioactive
products that could be released from a V/HTR in the event ofan accident, for example in the event of a LOCA leading toreactor depressurization, in particular regarding the carbon
dust that may exist in the form of deposits in the primarysystem and trap fission products under normal operation andwhich would be resuspended from the first moments ofdepressurization.
 The dosimetry consequences of deposition in the primarysystem of carbon dust from the core, which could containradioactive elements, need to be assessed, in particular withregard to in-service maintenance (including in-service
inspection).
 While progress in graphite manufacture means that graphites
containing less chlorine-36 after irradiation than those used in
GCRs are now available, the production of carbon-14 is amajor aspect to be taken into consideration for safety andradiological protection for downstream operations in the
V/HTR fuel cycle, for the decommissioning of these reactors
and for waste disposal.

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 Flooding of the site and rooms of a V/HTR should not lead tomajor difficulties for residual heat removal, but thepossibilities for air and water ingress into the primary system
would need to be analyzed bearing in mind that, during asevere earthquake, one or many breaks could occur on thissystem.

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 In the 1960s to 1990s, three power HTRs were operated(Peach Bottom and Fort Saint Vrain in the USA and the THTR
in Germany), with favorable experience feedback in terms ofradiological protection, although it is difficult to extrapolate
this to future power V/HTRs. Since then, analyses and R&D
have been pursued by designers and research bodies.
2/3

Gas-cooled Fast Reactors (GFR)
2/3/1

Presentation of the concept
The GFR concept aims to combine the advantages of a fast neutron
reactor, with regard to the objectives of uranium conservation andminimizing final waste, with high-temperature operation for
efficient power-generation.
The main design options associated with this concept have been
specified in the context of the EU’s FP6 Gas Cooled Fast Reactor(GCFR) project (see Reference Document [14]):
 a reactor producing thermal power of 2,400 MW;
 a power density in the core of 100 MW/m3;
 helium as a primary-system coolant, at a pressure of 70 bar; mean core inlet and outlet temperatures of 400°C and 850°C
respectively.
The core comprises hexagonal-cross-section assemblies made ofrefractory materials housing fuel pins. The fuel would be in the formof uranium and plutonium carbide pellets (10 metric tons ofplutonium per GWe), with silicon carbide cladding. As experiments
performed in HTRs have shown that rare-earth carbides diffuse
rapidly in carbide pellets and corrode the silicon carbide, thecladding is to be made of fiber-reinforced silicon carbide with a thinrefractory-metal liner for sealing, to protect the silicon carbide fromcorrosion by these rare-earth carbides. At equilibrium, the reactor
would be an iso-generator of plutonium . It would also be

Equilibrium is reached when
the composition of new fuel isconstant at each refueling.

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theoretically possible to transmute minor actinides in the core of a
GFR.

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The reactor design (Figures 7 and 8) comprises a metal vessel
surrounding the core, connected to three primary-system loops
containing intermediate heat exchangers. Three secondary-system
loops, connected to the intermediate heat exchangers, provide
power generation and could use gas turbines instead of steam
turbines. In addition, tertiary-system water/steam loops are planned
to supplement the power-generation system.
Figure 7
Schematic diagram of a(2,400 MWth) GFR – CEA data.

Furthermore, Decay Heat Removal (DHR) systems independent of
the normal systems would be provided for residual heat removal.
IRSN does not have information on the design of the reactor pit andthe systems planned to ensure that the concrete would be kept
sufficiently cool.
Finally, helium purification is planned during reactor operation.
Figure 8
Architecture of GFR systems –CEA data.

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2/3/2

Current state of GFR-concept development and outlook
The EU’s Sustainable Nuclear Energy Technology Platform (SNETP)
currently considers the GFR as an "alternative" solution to SFR
technology. With this in view, an agreement was signed in 2011
between the nuclear research establishments of the Czech Republic,
Hungary and Slovakia to design and build a 70 MWth experimental
reactor (the ALLEGRO project ). The initial cores of this reactor
would use Phénix-type assemblies with steel cladding and would
consequently operate at a limited temperature (core outlet gas
temperature of approximately 500°C).
The EU has funded GFR research work since 2005, with the GCFR
and GoFastR projects.
Various aspects of reactor design were covered in the GoFastR
project. For example, benchmark activities were defined fortransient analyses using computer models. Furthermore, IRSN isleading a working group for Technical Safety Organizations (TSOs)
within this project, covering the safety approach and accidents (inparticular, depressurization and core accidents). The industry hasorganized a "mirror" group on the same subjects. In April 2011, R&D
and safety approach support work for ALLEGRO was proposed to
the EU in the context of FP7 EURATOM projects, but it was not
adopted. A reactor such as ALLEGRO would also be a facility forfast-neutron irradiation, which could serve the development of SFRs
or even ITER.

In principle, the ALLEGRO’s
limited power and operating
temperature would lead todifferent safety characteristics
than an industrial GFR, in
particular for residual heatremoval.

By the consortium formed by
Hungary, the Czech Republicand Slovakia.

Currently, given the objectives combining uranium-resource
conservation, minimization of final waste and high efficiency, a key
technical obstacle is the design of fuel assemblies ----- in particular
the fuel itself and its cladding ----- that are able to withstand thespecified operating conditions (in terms of fast neutron flux, power
density and coolant temperatures). For this reason, the maximum
acceptable temperature for GFR fuel is currently unknown, in
contrast to that of the TRISO fuel for V/HTRs or the fuel pins for
SFRs. Safety analyses, in particular those concerning residual heatremoval in the event of depressurization, assume acceptabletemperatures of 1,600°C to 2,000°C. Significant R&D work has been
performed by the CEA. In particular, in 2007, irradiation capsules
were inserted into the Phénix reactor: the FUTURIX-MI experiment
for structural materials, and the FUTURIX-CONCEPT experiment forfuel and cladding materials. Currently, the Belgian Nuclear ResearchCentre (SCK-CEN) is responsible for the design and scheduling of© IRSN/2012 – All rights reserveddoc référence
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the irradiation of various potential GFR fuels (plates and pins) in theBR2 reactor, under the IrrDemo Program.

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R&D is underway with regard to materials that can be used up to
850°C (covering V/HTRs and GFRs). Thermal shielding is also underconsideration for some GFR structures, which could complicate thein-service inspection of such structures and be a source of loose
parts in the reactor.
The helium purification system has also been the subject ofdevelopment and specific research.
Generally, it can be said that the GFR could only be adopted on
condition that the various innovations envisaged are validated inyet-to-be-built experimental facilities.
2/3/3

Safety aspects specific to the GFRCompared with SFRs (which have risks associated with sodium),
V/HTRs (which have risks associated with graphite) or even LFRs
(which have risks of lead corrosion in particular --- see Section 2/4
below), a GFR has fewer risks associated with the fluids andmaterials used: an inert gas (helium) is used as coolant and there
are no materials susceptible to violent reactions with air or water
under normal operation. However, under accident conditions at very
high temperatures, all fuel-pellet and cladding materials can beoxidized by air or water.
In addition, compared with sodium, the use of helium in a fast
reactor offers some safety advantages:
 a significant reduction in reactivity insertion in the event of a
LOCA ("void effect"); the coolant cannot change phase (it is a non condensablegas).
However, GFRs have some specific issues with regard to safety andthese are discussed in the next section.
2/3/4

Aspects of the safety analysis
Normal operation, abnormal events and prevention of accidents
Prevention of helium leakage would seem to be made easier by thefact that it is an inert gas with no corrosive effects. However, it© IRSN/2012 – All rights reserveddoc référence
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should be noted that at the high temperatures targeted, helium
purity would be an important safety issue.
Prevention of accidents, particularly core damage, requires special
attention on GFRs, due to the use of a gas as coolant and the low
thermal inertia of the system. This concerns helium-leakage (LOCA)
scenarios in particular. In the event of depressurization of a GFR, it
is necessary to maintain coolant circulation to remove the residual
heat, at least during the first few hours following reactor scram. For
this, designers envisage using "close containment" (called "guard
containment" on the GoFastR project) around the primary system,
to provide a fallback pressure of around 10 bar in the event of abreak in the primary-system pressure boundary. In the current
design, this close containment would be a 33 m-diameter metal
sphere, full of air or inert gas, at a pressure slightly above
atmospheric pressure . The quantity of residual heat removal
available via natural convection alone is dependent on the fallbackpressure. The value currently adopted, which is consideredreasonable in terms of manufacturing feasibility, would not provide
adequate residual heat removal in the short term following
depressurization: low-power (battery-powered) forced convection
systems would be required during the first few hours. Coolant
circulation would be provided by DHR systems designed to cover
the whole range of pressures from 1 to 70 bar (normal operation).
Injection of a heavy gas, such as nitrogen or carbon dioxide, is being
studied, with the objective of limiting the effects of depressurizationin addition to the provisions aiming to maintain coolant circulation.

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Another solution would
involve a reinforced-concrete
containment building.

For all cases other than a hypothetical LOCA with "close
containment", the DHR systems are to be designed and sized toprovide core cooling by natural convection, without emergency
power. However, it should be noted that this residual heat removal
mode would not be totally passive, in particular as it would requirevalves to be operated.
Finally, special attention should be paid to the prevention ofsignificant water ingress, an accident which would lead to reactivity
insertion and oxidation of the fuel cladding.
Accidents without core melt
As for other fourth-generation reactor designs, the GFR is designed
using the classic "barriers" approach (fuel, primary system,
containment building) to reduce the risk of radioactive releases as
far as possible, in particular for accidents without core melt.
However, it should be noted that is not envisaged that the "close© IRSN/2012 – All rights reserveddoc référence
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containment" be considered as a barrier for the safety
demonstration.

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For accidents without core melt, the bounding accident scenario
with regard to the potential consequences has not yet been
specified. As for V/HTRs, a LOCA leading to the release of helium
into the "close containment", followed by leakage of this helium
into the reactor hall and ultimately into the environment should beconsidered. IRSN is not aware of any analysis of such a scenario.
However, the primary helium should contain very little
contamination under normal operation, due to its purification andthe strict requirements on cladding sealing (with a liner).
Accidents with core melt
It seems that total or partial core meltdown on a GFR needs to beconsidered. However, the materials that comprise the core havevaried high-temperature behaviors; they may melt, break down or
sublimate. IRSN is not aware of any analyses on GFR core
meltdown, or on means of limiting the consequences of such ameltdown. However, several remarks can be made on accidents
with core melt:
 In contrast to in SFRs, if a molten pool forms, it cannot be
efficiently cooled by the coolant (helium). A spreading area
would be necessary, as in third-generation PWRs such as the
EPR and the AP1000.
 As for SFRs, there is a risk of recriticality.
 Thermal radiation from the high-temperature core could
degrade the surrounding structures without them being incontact with molten materials, a phenomenon which does notexist for SFRs due to the presence of sodium. The thermal
shielding mentioned above should contribute to protecting
such structures under these conditions.
 Unlike SFRs, where only small amounts of the iodine present
the fuel pins are transferred into the primary system’s covergas if the cladding fails, GFRs do not benefit from thisfavorable effect provided by sodium.
These aspects highlight additional difficulties for the managementof accidents with core melt on GFRs.

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2/3/5

Assessment of the GFR-concept with respect to the
Fukushima accident
Earthquakes and floods
In principle, in terms of neutronics, GFRs are as sensitive to seismic
loads as SFRs. In terms of mechanics, the refractory core andinternal structures are lighter than those of SFRs, PWRs or V/HTRs,
because ceramics have about half the density of steels. In the eventof an earthquake, this would provide an advantage by reducing theinertial forces transferred onto the core support structures.
Given the risk of a LOCA in the event of an earthquake, a "hardened
safety core" seismic scram system, with a suitable detection
threshold, would be appropriate for shutting down the reactor assoon as possible. Also, as mentioned above, forced convection in thecore would initially be indispensable to remove the residual heat in
the event of a large break, whence the need for a preinstalled"hardened safety core" power supply.
Total loss of power or the heat sink
All the residual heat removal systems require water (there are no airheat exchangers ----- cooling-water pools are planned for the DHR).
As loss of the heat sink could lead to core melt in the short term, apreinstalled "hardened safety core" heat sink may be necessary.
In the event of total loss of power supplies, residual heat removal
could be performed by emergency systems operating by naturalconvection. However, the operation of active equipment items(such as valves) implies that these receive emergency power frombatteries. All this equipment must be of "hardened safety core"robustness.
Severe-accident management
As mentioned above, the consequences of core overheating anddegradation are not well understood. Therefore, IRSN cannot
currently take a position on this aspect for GFRs.
With respect to a possible LOCA (helium leakage), it would seem
necessary to exclude the possibility of water injection into theprimary system (due to the risk of vapor explosion).
2/3/6

Conclusion
In summary, the following points should be noted for GFRs:© IRSN/2012 – All rights reserveddoc référence
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 GFRs do not have any risk of violent reactions between the
coolant and air or water.

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 Due to the use of a gas (helium) as coolant and the absenceof moderators in the core (in contrast to V/HTRs), GFRs do
not benefit from high thermal inertia.
 Use of a gas as coolant also makes GFRs very sensitive tosituations such as a LOCA. In such situations, the safety
cannot be totally based on passive systems: active systems
are indispensable. Because of this, reactor design and systems
architecture may be complex.
 With regard to the possible consequences of an accident withsevere core damage, GFRs do not benefit from the trapping ofvolatile fission products that occurs in SFRs thanks to thesodium.
 It is currently difficult to have an opinion on whether thesafety level is at least equivalent to that of third-generation
reactors.
 With regard to events such as those that occurred at
Fukushima, the characteristics of GFRs would mean that, fromthe design stage, there would be a need for automatic safetysystems including support systems, such as power supplies
and heat sinks, that are of "hardened safety core" robustness
and able to operate very quickly.
 There is currently no fuel qualified for the temperature levels
planned for the coolant (850°C at the core outlet). This
constitutes a technological obstacle.
 Materials that can withstand the temperatures planned fornormal operations are the subject of R&D work, which is also
being performed for V/HTRs. However, in the event of coremelt in a GFR, important structures could be severely stressedby direct thermal radiation from the damaged core. Thermal
shielding is planned, but this could complicate in-service
inspection and be a source of loose parts in the reactors.
 No GFR has ever been built.

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2/4

Lead-or lead/bismuth-cooled Fast Reactors
(LFR)
2/4/1

Presentation of the concept
LFRs are fast reactors cooled by a molten metal such as lead, or alead-bismuth alloy often called LBE (for Lead Bismuth Eutectic).
Figure 9
Schematic diagram of a leadcooled reactor.

As for SFRs, LFRs can operate with low-pressure systems due to thehigh boiling point of lead or LBE (1,745°C for lead, 1,670°C for LBE).
In this Section, IRSN uses the LFR conceptual design of theEuropean Lead-cooled SYstem (ELSY) project developed as part of
the EU’s Sixth Framework Programme (FP6) (see Reference
Documents [15], [16] and [17]). Figure 10 gives a schematic diagramof the ELSY primary system.

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Figure 10
Schematic diagram of the ELSY
primary system.

As for the French SFRs Phénix and Superphénix, the ELSY project is"pool-type": the primary lead is confined in the (main) vessel.
In contrast to sodium, one advantage of lead or a lead-bismuth
alloy lies in the absence of any violent chemical reaction with air orwater. Because of this, the designers of ELSY have opted for a design
without an intermediate system, with six in-vessel steam
generators.
The thermal and electrical powers of ELSY are 1,500 MW and 600
MW respectively, with the temperature of the lead coolant ranging
from 400°C to 480°C. Lead circulation is provided by submerged
pumps.
Lead and LBE are weak moderators and neutron absorbers but
excellent reflectors, properties which enable the designers to "spaceout" the fissile material in the core. Compared with SFRs, thisenables a higher proportion of coolant to be used in the core, larger
cross-section coolant paths and a lower coolant speed (2 m/s) tolimit the structural erosion. With respect to SFRs, this also means
lower head loss across the core which promotes natural convectionof the coolant.
The power density in the core is approximately 110 MW/m3.

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MOX fuel with steel cladding is used in the ELSY project. The fuelpin bundles are laid out in a square pitch, forming "open" assemblies
like in PWRs. Another solution is the one used in SFRs (fuel pins
assembled into closed hexagonal housings).

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Nitride- or carbide-based fuels, with higher densities than MOX, canalso be considered for LFRs. They would have the advantage of
operating at lower temperatures (below the boiling point of lead or
LBE), of having no risk of forming oxides with lead or LBE in theevent of cladding failure, and of reducing the loss of reactivity
during a fuel cycle, which means that the neutronic weight of thecontrol rods can be reduced, reducing risks in the event of
inadvertent removal of these rods.
Due to the very large mass of lead involved (6000 metric tons), themechanical design of a reactor such as that of the ELSY project
presents a challenge, especially with regard to seismic loads. In
particular, this has led to reducing the ratio between vessel heightand diameter as much as possible.
The safety vessel, whose internal wall is insulated, is "anchored" inthe reactor pit. The space between the two vessels is maintained ata sufficiently low temperature by the Reactor Vessel Air Cooling
System (RVACS). Such a system may be sufficient to remove
relatively low residual heat, in particular for reactors of low
operational power or, in the case of ELSY, after the first month ofreactor shutdown (see Reference Document [17]). In the case of the
ELSY project, the additional Direct Reactor Cooling System (DRCS)
is therefore needed to remove the residual heat from the reactor in
the first few weeks following shutdown. Reference Document [16]also mentions the Reactor Concrete Cooling System (RCCS) whichcools the reactor pit walls.
2/4/2

Current state of LFR-concept development and outlook
Reactors using LBE coolant were developed and built in the Soviet
Union for submarine propulsion. Seven Alfa-class military
submarines powered by 155 MWth OK-550 reactors, and later BM40A reactors, using highly enriched uranium-235 fuel were operated
between 1967 and 1983. In 2007, the IAEA published an experience
feedback document on fast reactors (Reference Document [5]),which includes LFRs and gives details of notable events that
occurred on three submarine reactors:
 An accident in 1968: the core was blocked by oxides of LBE (inparticular), which had accumulated mainly during outages for© IRSN/2012 – All rights reserveddoc référence
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maintenance on the depressurized primary system, causing
partial core meltdown due to a lack of suitable procedures in
response to the signals received in the control room. There
were deaths due to acute radiation exposure. The mostnotable provision made on submarine reactors after this event
was a "hardening" of the procedures for monitoring oxygen in
the primary system and for purification to remove any oxides
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 An accident in 1971: damage to primary-system pipework
was observed, due to corrosion of the external surface of
these pipes caused by excessive humidity in the reactor
compartment due to a lack of leaktightness on the steam
generator. This led to a leak of radioactive LBE.
 An accident in 1982: widespread corrosion of the steam
generator tube bundle was caused by poor-quality feedwater.
This led to steam ingress into the primary system containing
the LBE. At the end of a fairly complex chain of events,
including human errors, 150 liters of radioactive LBE leaked
into the reactor compartment.
There is no operational experience on an industrial lead- or LBEcooled fast reactor anywhere in the world.
Russia continues to be interested in this type of reactor and isdeveloping two prototypes: a lead-cooled reactor (the 300 MWe
BREST-OD-300) and an LBE-cooled reactor (the 100 MWe SVBR100). According to information recently gathered by IRSN,
commissioning of both reactors is planned around 2020.
The EU’s Sustainable Nuclear Energy Technology Platform (SNETP)
has adopted the LFR, alongside the GFR, as an "alternative" solution
to SFR technology. In consequence, the ELSY project was launchedin 2006, followed by the Lead-cooled European Advanced
DEmonstration Reactor (LEADER) project in 2010, both funded by
the EU as part of FP6 and FP7 respectively. The LEADER project
aims to provide the conceptual design for an industrial-scale LFRdemonstrator called the Advanced Lead Fast Reactor European
Demonstrator (ALFRED), which Romania wishes to build on its
territory by 2025.
Additionally, the Belgian Nuclear Research Centre (SCK-CEN) is
going to build a demonstration reactor, called the Multi-purpose
hYbrid Research Reactor for High-tech Applications (MYRRHA), for
an Accelerator Driven System (ADS) using a particle accelerator todrive a sub-critical nuclear fission reactor. The reactor is to use LBE© IRSN/2012 – All rights reserveddoc référence
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as a coolant (and as a spallation neutron source activated by thebeam of protons). A preliminary design should be the result of theCentral Design Team (CDT) project included in FP7. Submission of
the MYRRHA safety options file to the Belgian Nuclear Safety
Authority is planned for 2014. Reactor commissioning is planned for
2024.

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Finally, there are projects for small-scale reactors in the USA, withthe 45 MWth "Small, Sealed, Transportable, Autonomous Reactor"(SSTAR), and in South Korea, with the "Proliferation-resistant,Accident-tolerant, Self-supported, Capsular and Assured Reactor"(PASCAR), and for experimental reactors in Sweden, such as the
0.5 MWth European Lead Cooled Training Reactor (ELECTRA).
2/4/3

Safety aspects specific to the LFRRisks of structural erosion and corrosion
This seems to be a key issue for LFR-type reactors.
Lead is highly eroding and for this reason its speed is limited toapproximately 2 m/s. Lead is also highly corrosive for steel
structures (in particular, for the cladding and vessels). Furthermore,lead oxides can lead to blockages in fuel assemblies. The method
used on the Alfa-class submarines, which remains the baseline
solution, was to create a layer of iron oxides on the surface of steels
in contact with lead, by injecting oxygen into the lead, combinedwith continuous purification to remove the lead oxides created. The
solubility of iron oxides in the lead depends on oxygen
concentration and temperature, which leads to different behaviors
between fuel-pin cladding and the main vessel for example. For the
ELSY project, the temperature of the lead at the core outlet has
been set at 480°C to limit the risk of structural corrosion on the
primary system . According to Reference Document [16],operation at typical SFR temperatures (550°C at the core outlet)
would create a corrosion risk, at least at the current state ofknowledge regarding the compatibility of the steels qualified for
SFRs with lead or LBE. With regard to the fuel cladding, whereoperating temperatures would reach approximately 560°C,
Reference Document [15] states that the corrosion-riskmanagement method mentioned above would not be adequate andthat the cladding would require coatings.

Although not explicitly
mentioned in the reference
documents consulted, it would
seem that a system for
injecting oxygen into theprimary system would also be
indispensable.

Given the diversity of materials and operating temperatures in an
LFR, corrosion-risk management, which requires even distribution of© IRSN/2012 – All rights reserveddoc référence
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the injected oxygen in the whole primary system, could be complex
for large-scale reactors like ELSY.

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Finally, given the risk of lead corrosion, it is difficult to imagine thatthe components could remain in the reactor for several decades.
The possibility of replacing the in-vessel primary coolant pumps andsteam generators is an objective that has been adopted for the ELSY
project (see Reference Document [16]).
Risk of embrittlement of steels in the presence of lead
The report in Reference Document [25] shows that, independent ofcorrosion risks, certain steels could become embrittled in contact
with lead or LBE, which could be exacerbated by irradiation. The
report mentioned previously (in Footnote 8) concerns T91 steel,which is a candidate for ADSs. It would therefore seem necessarythat designers and associated R&D bodies examine this subject indepth, so that the risk of sudden failure of components such as areactor vessel under accident loads can be excluded via a suitablechoice of materials and metallurgy processes.
Risks associated with slow chemical reactions between lead or LBE
and air or water
In the event of air ingress into a lead or LBE system, or in the eventof a small water leak from a steam-generator tube, compounds
such as oxides, hydroxides and hydrides would form in the lead or
LBE. These would increase its viscosity, reduce its heat-exchange
properties and lead to overheating or even fuel meltdown, as
occurred during the 1968 accident on a Soviet submarine reactor.
Risk of coolant freezing
Lead freezes at 327°C, which is only 73°C lower than the core inlet
temperature during LFR operation, while for SFRs the margin isapproximately 300°C. This requires heat devices for all the systems
that contain lead, to prevent lead freezing during reactor outages inparticular. As with sodium, lead expands when it melts. IRSN does
not have information on the consequences of lead freezing or
melting in a LFR.
The main advantage of LBE is that it freezes at 125°C, which ismuch lower than the freezing point of lead. This means that thefreezing risks for an LFR using LBE are in principle much easier tomanage. However, it should be mentioned that LBE froze in thereactors of three Soviet submarines (other than those mentioned in
Section 2/4/2, which had accidents), including one when it wasunder the icecap. This led to their permanent shutdown.

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Chemical and radiological risks

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Lead is highly toxic. Its Occupational Exposure Limit (OEL) is 20
times lower than that for sodium hydroxide (a compoundconsidered with regard to chemical releases from a sodium fire in an
SFR).
Furthermore, the use of LBE would lead to significant production ofpolonium-210, which is highly toxic. However, it should be noted
that, even if lead were used, polonium-210 would still be produced
due to the presence of bismuth impurities, albeit in significantly
lower quantities. The choice of lead as the coolant for the ELSY
project was mainly aimed at reducing production of polonium-210.
Risk of thermodynamic interactions
While lead and LBE do not react violently with air or water, contact
between these molten metals and liquid water could lead to "vapor
explosion" thermodynamic interactions, shockwaves and thedisplacement of structures and fluids. In particular, this led thedesigners of the ELSY project to adopt several lines of defense,
especially via design provisions for the in-vessel steam generators,
to "practically eliminate" not only the possibility of serious
consequences on the reactor’s internal structures but also gas
ingress into the core in the event of a steam generator tube rupture(see Reference Document [16]). Furthermore, outlets with blowout
diaphragms are connected to the cover-gas space over the lead,which acts as an expansion volume to dampen the overpressure .
It should be noted that the risk of overpressure in the reactor could
be more radically excluded by adopting an intermediate systemthat uses a low-pressure fluid that cannot interact
thermodynamically with the primary lead.

However, it is not stated in the
document consulted whether
these outlets release directly
into the reactor hall or into
special containers.

Initial analysis suggests that any contact between molten fuel andliquid lead would probably not cause a high-energy "vapor
explosion". However, this would need to be confirmed.
2/4/4

Aspects of the safety analysis
Normal operation, abnormal events and prevention of accidents
Experience feedback on LFRs is limited to a small amount of
information available about submarine nuclear reactors from theformer Soviet Union.
As stated above, the management of corrosion risks for metal
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pressure boundary, constitutes a key issue and could be complex fora large-scale reactor like ELSY. For future LFRs, this issue must be
resolved and it must be possible to perform in-service inspections
on the structures, in particular to check that there is no harmfulcorrosion. With regard to this, lead is opaque (like sodium) andspecific techniques need to be developed for inspecting the internal
structures of the primary system. However, it should be noted thatin-service inspections on the ELSY project should be facilitated by
the relative simplicity of the internal structures, in particular theabsence of a diagrid supporting the core (as the fuel assemblies are
suspended). The "sizing" of the space between the two vessels on
the ELSY project should be a compromise between the need to
introduce devices for inspecting the two vessel walls and the needto avoid too large a drop in the lead level in the main vessel in theevent of leakage from it.

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Furthermore, it has been stated above that the possibility of
replacing the major in-vessel components (coolant pumps andsteam generators) has been adopted as a design objective for thereactor.
Accidents without core melt
As for SFRs, LFR cores could experience reactivity insertion due tothe "void effect", for example in the event of lead overheating (dueto fuel assembly blockage) or uncovery. Although this effect isweaker on LFRs than on SFRs due to the coolant density, IRSN has
no information on the management of such an accident for the
projects currently underway. Nevertheless, it can be noted, withregard to the coolant boiling risk, that the margin is greater on LFRs
than on SFRs as the boiling point of sodium is 900°C compared with
1,750°C for lead (which is above the melting point of the fuel-pin
cladding and steel structures).
LOCAs have been analyzed (see Reference Documents [13] and[16]). In this respect, the combination of a core with moderate
power density, high thermal inertia and use of a coolant with avery high boiling point makes the LFR design tolerant with regard to
LOCAs, including in the event of scram failure. Even under suchconditions, the melting point of the cladding would not be reachedaccording to Reference Document [16].

Essentially just the same as forthe sodium in an SFR, for thesame coolant volume.

Some comments are made below on the safety functions "Controlof reactivity", "Removal of heat from the core", and "Confinementof radioactive materials".

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There is no information available on the design of the control rods
but the design should be such that they can be inserted quickly into
the core despite the high density of lead (twelve times that of
sodium).
b) Removal of heat from the core
As stated above, the ELSY design facilitates core cooling by naturalconvection in the event of primary-coolant pump shutdown, forexample in the event of loss of normal and emergency power
supplies. Low head loss across the core and lead’s high density arefactors that promote natural convection in the primary system.
The sources consulted (Reference Documents [15] and [17]) givedifferent architectures for the ELSY project’s residual heat removal
systems. According to Reference Document [15], in the event thatthe normal residual heat removal systems are unavailable, two
redundant and diversified systems, which are said to be able tooperate passively, remove the residual heat. One of these systems
comprises two loops, each having a lead-water heat exchanger
immersed in the primary system, into which water is fed by gravity
from a tank, which then evaporates and is released into theenvironment via a stack. The second system also has two loops,
each using a steam generator as a heat exchanger. IRSN has no
specific information on the design requirements and possible
operating modes for the two RVACS and RCCS cooling systems forthe space between the two vessels and the reactor-pit concrete
respectively.
The high thermal inertia of an LFR strongly limits coolant
temperature-rise kinetics in the event of total loss of coolingsystems, whence providing significant grace periods (several hours)
for installing means of residual heat removal before the reactor
reaches temperatures that could lead to core damage.
c) Confinement of radioactive materials
Given the high toxicity of lead, its containment requires special
attention to prevent releases into the environment as far as possible(design provisions, in-service inspections, etc.).
Accidents with core melt
Available analyses on accident sequences performed for the ELSY
project (see above) do not reveal any scenarios that may lead tocore melt. However, it should be noted that, as for SFRs, prolonged
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failure by creep and possible failure of the reactor pit, an event
whose consequences have not been analyzed. However, the highthermal inertia of an LFR’s primary system provides a significant
grace period to repair some of these systems, or install others,
before damage occurs.

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IRSN does not have information on the behavior of molten fuel andits relocation in an LFR. Molten MOX fuel should float on the
primary lead and, for this reason, a "core catcher" at the bottom of
the reactor would be of no use. IRSN has no information on the riskof recriticality of corium floating on the primary lead.
Furthermore, as stated in Section 2/4/3, special attention has been
paid in the ELSY project to the "practical elimination" of reactor
damage in the event of a steam generator tube rupture. The
associated design provisions have been supported by simulations
performed using the SIMMER simulation tool (Reference Document[16]). However, with respect to experience feedback concerning
steam generators, use of the rupture of only one steam generator
tube in the analysis would clearly require substantiation.
Finally, in principle, lead (like sodium) can trap fission products(such as iodine), which are significant contributors to theradiological consequences of an accident.
IRSN does not have information on the design of the thirdcontainment barrier (in particular, the reactor building andventilation systems) for the ELSY project.
2/4/5

Assessment of the concept with respect to the
Fukushima accident
Earthquake
Earthquake is a hazard which requires design precautions on LFRs.
Movements of fuel assemblies could cause reactivity insertions andthe considerable mass of lead contained in the reactor vessel could
impose large loads on structures, especially given that an
earthquake could cause lead displacements with wave effects. In
the ELSY project, an aseismic isolation system using anti-seismic
pads under the reactor is planned and an EU project on the subject
was launched in 2012: "Seismic-Initiated events risk mitigation in
Lead-cooled Reactors" (SILER).
As for SFRs, a "hardened safety core" seismic scram system shouldbe planned.

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Flooding

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Unlike for SFRs, there is no risk of chemical interactions between
lead and water. However, simply because of the highly toxic natureof lead, robust prevention (with "hardened safety core" construction
provisions) of the risk of contact between lead and water should beplanned, such as that described in Section 2/1/5 for SFRs.
Total loss of power supplies
As stated above, the ELSY project seems to have passive residual
heat removal systems (until the water in the tanks is exhausted)
and to have the ability to circulate lead by natural convection in thecore, which, in terms of safety, positions it favorably with respect toother designs (especially GFRs) in the event of total loss of power
supplies. However, as also stated above, IRSN lacks information on
the requirements adopted by the designer concerning the cooling of
the vessel walls and the reactor-pit concrete.
A difficult situation for LFRs could result from loss of the means ofheating the lead contained in the primary system when the reactor
has been shut down for sufficiently long that its residual heat is
insufficient to prevent the lead from freezing. Maintaining the
integrity of a vessel containing solidified lead is a subject that needs
analysis. Use of LBE would have fewer risks in this respect.
Loss of the heat sinkIn the ELSY project, there are two heat sinks: sea or river water forthe turbine (the normal means of heat removal) and theatmosphere for the residual heat removal systems. For these lattersystems, the stacks to discharge water vapor could constitute asensitive area with respect to hazards. The possibility of water topup into the tanks associated with one of the two residual heatremoval systems would seem to be necessary ("hardened safetycore" provision).
Severe-accident management
For situations similar to those that affected reactors at FukushimaDaiichi, LOCAs should be considered. The issue is quite similar tothat for SFRs. It is hard to envisage supplementing the lead in thereactor vessel using equipment connected in an emergency,
although the risks of such an operation would be lower than for
SFRs as lead does not react violently with air or water. Preinstalled,"hardened safety core" provisions would be desirable, in particular
for reinjection into the main vessel of any lead that may leak into
the safety vessel.

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IRSN has no information on the possibility of rapid core unloading.

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2/4/6

Conclusion
In summary, the following points should be noted for LFRs:
 Compared with sodium, lead has the advantage of notreacting violently with air or water.
 However, even without a violent reaction, contact between air
or water and lead or LBE would cause an increase in theirviscosity, which could cause overheating of fuel in the core.
 As for sodium, liquid lead can cause thermodynamic
interactions when in contact with easily vaporizable fluids(such as water), which would cause overpressures. In thisrespect, the LFR design with in-vessel steam generators would
require special precautions.
 Depending on the materials envisaged, detailed research may
be required on the risk of lead or LBE causing embrittlementof metal structures resulting in their sudden failure.
 Lead is highly toxic. The risk of releases in the event of leaks
from systems or components containing lead therefore
requires special attention, and should be "practicallyeliminated" in the event of flooding.
 Given the specific properties of lead, it is possible to design
cores with lower power densities than for SFRs, which are lesssensitive than SFRs to loss of cooling and which promotenatural convection.
 The large mass of lead gives LFRs high thermal inertia andavailable analyses regarding the ELSY project do not reveal
any core-melt initiators: in the event of total loss of residual
heat removal systems and scram failure, the melting point of
the cladding should still not be reached. However, IRSN finds
it hard to pronounce on the exhaustiveness of the analyses
performed, and if core melt cannot be excluded, the accident
phenomena would be quite original as the corium would floaton the surface of the lead. No analyses are available on thissubject.
 The risk of structural corrosion by the lead is a crucial aspect
which, at the current state of knowledge, limits the operating
temperatures (460°C at the core outlet). While the riskmanagement provisions made on Soviet submarine reactors© IRSN/2012 – All rights reserveddoc référence
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would seem to have been validated by experience, it may be
more complex to design analogous provisions for large-scale
LFRs like the ELSY project.

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 Experience feedback is limited to Soviet submarine reactors.
Three of these reactors suffered "serious damage".
2/5

Molten Salt Reactors (MSR)
2/5/1

Presentation of the concept / Current state ofdevelopment
The MSRs studied fall into two groups: reactors in which the molten
salt acts only as a coolant and reactors whose fuel is dissolved in alithium-fluoride-based eutectic molten salt (FLi, FLiNaK or FLiBe ).
For this second type of MSR, the molten salt contains a mixture ofnatural thorium(thorium-232) and uranium-233. The reactor
operates by fission of the uranium-233 produced from the thorium232 (the fertile material) and therefore requires an initial loading of
uranium-233 or plutonium for start up.
Dissolved-fuel MSRs are a completely different concept than otherfourth-generation reactors, in particular because the fuel andcoolant are mixed together. Furthermore, a special fuel reprocessing
unit must be associated with such a reactor to continuously
eliminate neutron-absorbing elements, to prevent the chain
reaction from being stopped by the generation of neutronic poisonsin the coolant.

FLiNaK is a LiF-NaF-KF
mixture, with a melting pointof 454°C; FLiBe is a LiF-BeF2
mixture with a melting pointof 459°C.

Thorium is four time moreabundant than uranium. Thereis a major thorium deposit inBrittany.

MSR designs can include both thermal and fast neutron reactors.
The size of the reprocessing unit and its coupling with the reactor
strongly depends on the neutron spectrum used. Reprocessing-unit
and reactor operations are less correlated for fast neutron reactors
than for thermal ones, as the neutronic poisons have more effect on
the latter.
As for SFRs and LFRs, MSRs use a low-pressure coolant.
Two experimental MSRs were built and operated in the USA. The
first was a reactor designed for military aircraft propulsion, built
during the 1950s as part of the Aircraft Reactor Experiment (ARE)project. The second was the Molten Salt Reactor Experiment(MSRE), which was built at the Oak Ridge National Laboratory(ORNL) in 1962 and reached criticality in June 1965. It did not use a© IRSN/2012 – All rights reserveddoc référence
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fertile material, but rather fuels based on uranium-235, plutonium
and then uranium-233. This 8 MWth reactor was shut down in1969 after approximately 13,000 hours of operation.

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Figure 11
Schematic diagram of a reactor
where the fuel is dissolved inmolten salts.

The ARE project encountered difficulties associated with structural
corrosion by the salt. Experience feedback led the designers todevelop a procedure to manage this phenomenon by controlling thesalt’s redox potential. This procedure was implemented on the
MSRE, which does not seem to have encountered corrosion
problems.
Studies were then performed until 1976 at ORNL as part of the
Molten Salt Breeder Reactor (MSBR) project, a planned prototype2,500 MWth reactor.
Five countries are currently taking an interest in MSRs:
 France, where the National Centre for Scientific Research(CNRS) has been involved since 1997. The CNRS started byproducing a reassessment of the MSBR design and is currently
studying the Molten Salt Fast Reactor (MSFR) design, a fast
breeder reactor (which is also able to incinerate thetransuranium elements produced in current reactors).

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 Russia, where the Rosatom Nuclear Energy State Corporation
has developed the "Minor Actinide Recycling in molten Salt
reactor" (MARS) project, which is also a fast reactor (with aview to incinerating minor actinides).

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 Japan, where Mitsubishi is studying a small-scale (350 MWth)MSR thermal reactor, the FUJI-12, which is very similar indesign to the MSBR.
 The USA, where the Oak Ridge National Laboratory produced
a catalogue of technology options for Fast-Spectrum Molten
Salt Reactors (FS-MSRs) in 2010.
 China, which committed itself to an MSR project in 2011
(with funding of $250m).
Furthermore, the EU’s "Evaluation and Viability Of Liquid fuel fast
reactor" (EVOL) project was launched in early 2011 as part of theFP7-EURATOM-FISSION program. In particular, it aims to produce
the preliminary design for the MSFR and an initial safety approach.
The discussions below are based on information obtained from the
CNRS concerning the MSFR. The 3,000 MWth MSFR is to use
uranium-233 and thorium-232 as fuel, containing 18 m3 of (LiFThF4-UF4) salt fuel, with 9 m3 in the core area. Due to its highboiling point (1,800°C), this salt can reach temperatures of 700°C
to 750°C without requiring high pressures, which means that
thermodynamic efficiencies of around 50% can be envisaged. The
core comprises a space filled with salt fuel inside the first barrier(the "core chamber"), which is itself located inside an inerted vessel
containing the primary-system components.
Figure 12
MSFR primary system and core(schematic view) –Source: CNRS.

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The MSFR has a high power density: 330 MW/m3. The power is
produced in the coolant itself with no structural components in thehigh neutron flux areas.

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2/5/2

Safety aspects specific to the MSR
Risks of structural corrosion
The corrosive effects of molten-salt fuels are a major risk that must
be considered, especially with regard to the design of thecontainment barriers. The procedure for managing redox potential
used in the MSRE seems to provide an answer to the problem, andwas validated by MSRE operations. According to the CNRS, such a
procedure could be applied to a power reactor like the MSFR, bymeasuring the redox potential of the extracted salt and correcting
this potential in the injected salt. Clearly, this would need to bevalidated in a prototype or demonstration reactor.
In any event, structural corrosion by salts has been the subject of
R&D work since the 1950s, in particular in Japan and the USA,
which are also supporting the development of fusion reactors using
molten salt technology. In summary, Reference Document [18]essentially states that understanding corrosion mechanisms stillrequires significant R&D work to better determine the corrosion
resistance of various metals that could be used in an MSR,
depending on factors such as the type and composition of the salts,
their purity, their oxygen concentration and various possible
electrochemical effects. For the MSFR project, CNRS has adopted(nickel-, chromium-, molybdenum- or tungsten-based) Hastelloy®
alloys.
Salt toxicity
Risks associated with the chemical toxicity of the salts used should
also be examined. FLiBe seems to merit special attention, given the
toxicity of beryllium . The salt itself is not toxic but it would benecessary to consider its reduction or decomposition by overheating(above about 1,800°C), leading to the release of beryllium oxide or
beryllium metal. This question clearly requires deeper analysis.
However, it should be noted that the CNRS has not adopted FLiBefor the MSFR, as they consider that beryllium, as a neutron
multiplier, could lead to risks of recriticality in the reprocessing unit.

Beryllium is carcinogenic, evenin very low doses. It can also
cause an allergic reaction
called berylliosis.

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Chemical risks and risks of thermodynamic interactions with molten
salts

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According to the sources consulted, none of the salts mentioned in
Section 2/5/1 react violently with air or water, including thesodium- and potassium-based salt (FLiNaK). Clearly, this would
need to be confirmed.
IRSN has no information regarding the thermodynamic interaction
between salts and water. This issue also needs further analysis.
Control of reactivity
MSRs do not have absorber rods or burnable poisons to control thechain reaction. Shutdown is performed by draining the fuel salt.
Power is controlled by "demand", via acting on the intermediatesystem pump. A decrease in flowrate leads to a reduction in reactor
power. The feedback coefficients are sufficiently negative to provide
this control in a few tens of seconds for a 50% power variation,
without a significant variation in temperature (just a few °C). It
should also be noted that the void coefficient is strongly negative.
Removal of heat from the core
For the MSFR, it is envisaged that after the fuel salt has been
drained into the dump tank(s) located below the reactor, residual
heat removal can be provided by passive systems, using naturalconvection in the tank(s) and surrounding heat pipes. Residual heatremoval by active systems under normal conditions is also planned.
Confinement of radioactive materials
MSRs are unusual in terms of containment barriers as there is no
cladding to contain the fission products in the fuel. The firstcontainment barrier is provided by all the pipework andcomponents in which the fuel salt circulates (such as the "corechamber", pumps, heat exchangers, upper rings and dump tank(s)).
The second containment barrier is the inerted vessel containing the
primary system and the dump tank(s) for the fuel salt loop.
No information is available on the reactor building which should
provide the third barrier. According to the CNRS, this should be
similar to a PWR or SFR reactor building.
2/5/3

Aspects of the safety analysis
Obviously, the WENRA objectives in their current form cannot beapplied to MSRs, in particular due to the fact that the fuel is molten© IRSN/2012 – All rights reserveddoc référence
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during normal operation and due to the absence of cladding.
Furthermore, an appropriate application of defense-in-depth andthe safety demonstration would be required for MSRs. An
unavoidable initial step, especially important for the safety analysisof an MSR design, would be to determine the radioactive contents(in terms of aspects such as the locations, quantities and spectra ofradionuclides) in the system as a whole, including the reactor itself
and the reprocessing unit, which would contain fuel salt and fission
products. This analysis should cover normal operation and various
possible failures (such as leakage), to systematically substantiate
aspects such as the architecture and design of the barriers and to
identify any risks of criticality.

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Nevertheless, some comments on various safety aspects can bemade here.
Firstly, it should be stressed that molten-salt technology has beenused for several decades in industrial heat-transfer processes (inparticular in aluminum manufacture). However, these processes use
different salts than those involved here. With regard to nuclear
reactors, as stated above, corrosion risks seem to have been wellmanaged on the MSRE using a procedure for controlling the redox
potential of the salts, which clearly would need to be validated forlarger reactors.
With regard to controlling reactivity, the MSFR has favorable
neutronic characteristics, in particular a negative void coefficient. As
stated above, "reactor scram" is performed by fuel salt drainage.
Thus, drainage is a safety function which must be very highly
reliable, like the redundant and diversified shutdown systems onother designs. A passive drainage system using a "freeze plug" isplanned for the MSFR. An important aspect is that the fuel salt
recovery device is designed to avoid any risk of criticality.
The design should also reduce as much as possible risks of water
ingress into this salt recovery device, as the introduction of amoderator could make the fuel salt critical again (addition of amoderator into fast-neutron fuel may make it go critical and lead toa power excursion).
The risk of criticality can be reduced by adopting a design using
several tanks installed in separate cells.
The MSFR has low thermal inertia. For example, in the event of lossof intermediate-system cooling, fuel salt drainage should occur
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which are unacceptable for the structures of the first containment
barrier, and within 30 minutes to prevent salt boiling.

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In-service inspection is an important aspect, both with regard to
structural corrosion risks and to check that thermal shielding
remains in place, as this could be a source of loose parts. IRSN has
no data on this topic, particularly from the R&D point of view.
However, reactor emptying will be a ‘routine’ task and as suchshould open up prospects in this area.
Furthermore, advantages of MSRs highlighted by designers arecontinuous (or small batch) fuel reprocessing, the absence of highlyirradiated structures (no cladding or fuel assemblies) and therecycling of actinides (whether minor or otherwise) in the reactor.
Finally, an MSR fleet would minimize the transportation ofirradiated fuel due to on-site fuel reprocessing. However, each MSR
itself presents a major proliferation risk.
2/5/4

Robustness with regard to the events that occurred at
Fukushima
Earthquake
IRSN has no analysis available regarding the seismic resistance of
this type of reactor and it is difficult to take a position on this issue.
However, the small mass of the combined fuel and coolant
compared with SFRsand LFRs should be a favorable aspect for MSRs.
The fuel salt is homogenous and totally fills the "core chamber". The
overflow tank is located above the reactor and has too small a
volume to present a risk of criticality. Waves in the tank may posemechanical problems. The general design of this overflow tank has
not yet been fixed for the MSFR (Figure 12 shows a possible
solution with a ring located above the reactor).
Flooding
An important issue to be examined is the risk of water ingress into
the dump tank(s) containing the fuel salt in the event of flooding of
the facility or in the event of an earthquake leading to leakages
from the tank(s).
Site selection and robust ("hardened safety core") design provisions
should aim to "practically eliminate" such risks.

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Total loss of power supplies

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The MSFR seems to be tolerant of total loss of power supplies, inparticular due to the fact that the dump tank(s) can be cooled
passively. It should also be noted that the "reactor scram" system,
based on gravitational drainage of the fuel salt, can be activatedpassively by a "freeze plug" in the event of excessive salt
temperature.
Loss of the heat sink
The MSFR seems to use only the atmosphere as a heat sink under
accident conditions. As stated above for SFRs and LFRs, air heatexchangers located high up on the facility could be a sensitive pointin the event of an external hazard (such as aircraft crash or
explosion).
Severe-accident management
The spread of fuel salt into the environment would have severe
consequences. However, the approach to preventing and limiting
the consequences of such an event is yet to be specified. It would
need to identify possible scenarios (such as leakage from the dumptank(s) containing fuel salt combined with leakage from the thirdcontainment barrier).
2/5/5Conclusion
In summary, the following points should be noted for MSRs:
 Compared with sodium, the salts envisaged for MSRs do notreact violently with air or water.
 Among the salts envisaged, those that contain beryllium may
have to be excluded due to criticality risks in the reprocessing
unit and to prevent any spread of beryllium, which is highly
toxic, in the event of reduction or decomposition of the salt.
 The fact that the fuel is dissolved in the salt is a particularity
which means that the defense-in-depth approach must be
adapted, paying special attention to the characterization of
the radioactive contents (in terms of aspects such as thelocations, quantities and spectra of radionuclides) in thesystem considered as a whole, including the reactor itself andthe reprocessing unit, which will contain fuel salt and fission
products. This analysis should systematically substantiate thearchitecture and design of the barriers and identify risks ofcriticality, in particular in the event of flooding.

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 It seems possible to use passive systems to stop the chain
reaction and remove the residual heat on MSRs.

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 MSRs have low thermal inertia, which means that operators
only have a limited grace period (of the order of ten minutes)
to intervene in the event of failure of the "reactor scram"drainage system.
 There are currently no analyses available regarding
conceivable accident scenarios and their consequences for theenvironment.
 The risks of structural corrosion by the salts are a significant
issue that has been the subject of R&D work since the 1950s.
With regard to these risks, there is a procedure for controlling
the redox potential of the salts, which seems to have beenvalidated during operation of the American MSRE reactor.
Clearly, this procedure would need to be validated for a larger
reactor. Alloys are available that are compatible with the useof salts at the temperatures targeted for MSRs (700°C):
nickel-, chromium-, molybdenum- or tungsten-based alloys.
 IRSN has no information on MSR in-service inspection.
2/6

SuperCritical-Water-cooled Reactors (SCWR)
2/6/1

Presentation of the concept
Supercritical water-cooled reactors are the only design selected by
the GIF that uses water as coolant. In this concept, water ismaintained under "supercritical" thermodynamic conditions, inpractice above 221 bar and 374°C (see Figure 13 below), whichmeans that an efficiency approaching 45% can be envisaged,
compared with 33%-35% for pressurized water reactors.
In the reference design, the reactor would operate at a pressure of
250 bar with core-inlet and core-outlet water temperatures of
280°C and 500°C respectively (possible core-outlet temperatures
ranging from 500°C to 625°C are mentioned). Figure 13 abovesummarizes the thermodynamic conditions for water in pressurizedwater reactors (PWRs), boiling water reactors (BWRs) and SCWRs.
As for boiling water reactors, the turbine is directly fed by thesupercritical reactor coolant water. However, a version with anintermediate system has been studied, with a view to avoiding the© IRSN/2012 – All rights reserveddoc référence
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risk of contamination of the turbine and auxiliary systems. See
Reference Document [19].

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Figure 13
On the left is the phase diagram
for water (S for solid, L for liquid,
V for vapor) — the critical point
and the operating "points" for
BWR, PWR and SCWR reactors
are shown; on the right is thetemperature-entropy diagram forwater.

Two design options are envisaged, one using a pressure vessel (seeFigure 14) and one using pressure tubes as in CANDU reactors (seeFigure 15).
Figure 14
Schematic diagram of a pressurevessel-type SCWR.

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Figure 15
CANDU-style pressure-tube-type
SCWR, here paired with ahydrogen production facility; tothe right: a pressure tube and its
fuel bundle.

Two fuel cycles are also envisaged: an open cycle using UO2-based
fuel in a thermal reactor and a closed cycle using MOX fuel in a fast
reactor.
Given the low density of supercritical water, thermal reactors would
require moderator elements in the core. For this reason, several
projects use "water rods", analogous to control rods, in which thesupercritical water flows from top to bottom. The loss of reactivity
as the fuel is irradiated can be compensated for by controlling theflow rate in these "water rods", which avoids the need for highreactivity reserve.
For fast SCWRs, preliminary analyses show that a negative void
coefficient can be envisaged.
In both cases, the electrical powers cited are between 1,000 and
1,700 MW.
The power density in the core of an SCWR is 100-110 MW/m3.
With regard to fuel cladding, the zirconium-based materials
generally used in light water reactors could not be used withsupercritical water. Nickel-based alloys are planned.

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Current state of SCWR-concept development andoutlook
The SCWR concept is an optimized version of current light water
reactors, in particular with regard to economics. The very high heat
capacity of supercritical water means that the mass flowrate in thecore can be limited (it would be eight times less than in a PWR),
reducing the pumping power required. In addition, the use of
supercritical water as a coolant avoids the problems associated withliquid-vapor phase changes, such as departure from nucleate boiling(DNB) or dryout, which is one of the limiting factors for PWRs andBWRs. In principle, the SCWR concept, with or without anintermediate system, eliminates the need for pressurizers and steam
generators (as needed in PWRs) or recirculation pumps , steam
separators and driers (as needed in BWRs).
Furthermore, supercritical water has been used in coal-fired power
stations for many years, which provides significant experience
feedback for the conventional island, along with advances in termsof turbine technology and corrosion-prevention provisions (seeReference Document [23]).
Nevertheless, the use of supercritical water in a nuclear reactor
raises many questions, especially concerning materials (cladding
materials and materials for reactor structures), which are frequently
considered to constitute the major difficulty for this design withregard to the characteristics of supercritical water (chemistry, hightemperatures and high pressures). In addition, the very particularbehavior of water in the "pseudo-critical " region ----- withsignificant variations in its thermodynamic properties depending on
the thermal flux in the fuel and its mass flowrate ----- would require
much research, including with regard to the normal transients ofreactor start up and shutdown. These considerations mean that four
subjects can be highlighted that would require major developments
so that the feasibility of an SCWR could be assessed:

In a BWR, the unvaporizedwater is redirected to the core
inlet by recirculation pumps.

On the left-hand diagram in
Figure 13, "pseudo-critical"
temperature and pressure
conditions correspond to theextrapolation of the liquidvapor curve.

 materials;
 the chemistry of supercritical water;
 the thermo-hydraulics of supercritical water combined withneutronics;
 the loss of coolant accident (LOCA) scenario, during which thewater would change from the supercritical state to a "normal"
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large variations in heat exchange depending on thecomposition of the mixture.

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The main countries involved in developing the SCWR are as follows: Canada, via Atomic Energy of Canada Limited (AECL) andvarious universities, which is probably the country mostcommitted to the concept and has developed the CANDUSCWR design, a pressure-tube-type reactor using a thorium
cycle. The reference design would provide a thermal power of
2,550 MW (1,200 MWe), with core inlet and outlet
temperatures of 350°C and 625°C respectively and an
efficiency of 45%.
 Europe, with the High Performance Light Water Reactor(HPLWR), funded by the EU under the Framework
Programmes for Research and Technological Development.
The reference design would provide a thermal power of
2,300 MW (1,000 MWe), with core inlet and outlet
temperatures of 280°C and 500°C respectively and an
efficiency of 43.5%. The main stakeholders are AREVA NP and
the Karlsruhe Institute of Technology (KIT). The CEA is listed
as a key stakeholder in the program but its contribution ismodest.
 Japan, in particular the University of Tokyo, which has been
pursuing work on a thermal SCWR (the "Super LWR") since
1989 and on a fast SCWR (the "Super Fast LWR") since 2005.
 South Korea (via KAERI), which is developing the SCWR-SM
design using a solid moderator (ZrH2).
 Russia, which is developing various designs, including the SKD
with primary and secondary systems, using natural convectionin the primary system.
Work has also been performed in the USA with the Nuclear Energy
Research Initiative (NERI) project funded by the DOE (see Reference
Document [21]). The reference design would provide a thermal
power of 3,600 MW (1,600 MWe), with core inlet and outlet
temperatures of 280°C and 500°C respectively, 250 bar pressure
and an efficiency of 45%.
In addition, it should be noted that the IAEA has launched two
Coordinated Research Projects (CRPs), one on the "Benchmarking ofStructural Materials Pre-selected for Advanced Nuclear Reactors"(2010-2014) and the other on "Heat Transfer Behavior and Thermohydraulics Code Testing for SCWRs" (2008-2012). All stakeholders
mentioned above are involved.

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However, the increasing number of projects could give a false
impression. In fact, the financial and human resources involved arevery limited.

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2/6/3

Safety aspects specific to the SCWR
As mentioned in the last section, the specific safety issues
associated with the SCWR concern the unusual properties ofsupercritical water.
Heat transfer
Reference Document [22] gives a fairly clear summary of the
problems associated with heat transfer when using supercriticalwater, due to the significant variations in its thermodynamic
properties around the so-called pseudo-critical temperature. Aroundpseudo-critical conditions, a deterioration of heat transfer between
the fuel and water can occur under certain specific thermal-flux andmass-flowrate conditions. Much research exists on the subject,which shows that these phenomena appear for high thermal fluxes
and low mass flowrates. Thus, it can be assumed that theconsequences of an abnormal event leading to an increase inthermal flux or a decrease in mass flowrate could be amplified if thevalues that trigger the heat transfer deterioration phenomenon
were reached. This phenomenon could occur abruptly and lead toan equally sudden rise in cladding temperature. While the extensive
research has established a certain number of laws that can be used
to predict the appearance of this phenomenon and its
consequences, the complexity of reactor-core design means thatdevelopments are being pursued on this subject to explore in detailthe geometrical configurations and mass flowrate distributions that
could be encountered.
Furthermore, even without considering the incident transients
mentioned above, the temperature ranges associated with SCWR
operation (well below the pseudo-critical temperature at core inlet
and well above at core outlet), and sudden changes in thethermodynamic properties in the pseudo-critical region, mean thatthe reactor’s normal start-up and shutdown phases must be studied
in detail. Study of these phases is not easy, given the physicalcomplexity of an SCWR core: several zones with different levels of
uranium-235 enrichment, strong coupling between neutronics andthermo-hydraulics, water flowing up the fuel assemblies and down
the "water rods" and multiple passes through the core to heat the© IRSN/2012 – All rights reserveddoc référence
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water from 280°C to 500°C (the EU's HPLWR project has a threepass core).

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The chemistry of supercritical water
Like its thermo-hydraulic properties, the chemical properties ofsupercritical water are very different to those of the water in PWRs
or BWRs. Much research is underway to better understand thischemistry, using the experience acquired in conventional power
stations, to determine the parameters to be managed to limit its
aggressiveness towards the reactor's structural materials andcladding. Furthermore, it should be noted that the chemistry ofsupercritical water under irradiation, especially with regard toradiolysis, is currently unknown.
In particular, research is aiming to assess the resistance of various
alloys to generalized corrosion, stress corrosion and irradiationassisted stress corrosion cracking (IASCC) phenomena insupercritical water. Data is very sparse in the literature and there isa major need for experiments. Given the operating conditions (interms of temperature and irradiation), assessment of alloys mustalso cover their resistance to heat creep and irradiation creep.
2/6/4

Aspects of the safety analysis
The issues discussed above show that the most important current
research is based on the objective of obtaining "robust" normal
operating conditions and adequate accident prevention. At thisstage, much R&D is still required simply to become convinced that
the SCWR concept is well mastered and can be well managed in allphases of normal operation.
The information that is currently available is barely adequate totake a detailed position on the issues associated with preventing
accidents and limiting their consequences. Nevertheless, given theanalysis performed by the IAEA in particular (see Reference
Document [31]), it is reasonable to assume that the SCWR design
could benefit from the safety advances obtained from developing
third-generation light water reactors, especially regarding thepossibility of implementing passive systems. The safety systems
outlined in developing the SCWR concept reuse the major principlesof BWRs (using systems such as borated water tanks and especiallya safety injection system), and core melt is taken into account at
the design phase. For example, hydrogen recombiners and a "corecatcher" are planned in Reference Document [24].

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2/6/5

Robustness with regard to the events that occurred at
Fukushima
It can simply be said that the SCWR concept is in a similar position
to current light water reactors with regard to sensitivity to theissues raised by the Fukushima accident (earthquake, flooding, lossof power supplies and heat sinks, and severe-accident
management). Naturally, design of an SCWR should involve taking
into account developments on current light water reactors to
improve their robustness with regard to such events.
2/6/6

Conclusion
On paper, the SCWR concept offers very economically attractive
characteristics, mainly in terms of being an extrapolation of current
light water reactors, in particular BWRs. However, while muchresearch exists on SCWRs (including studies of incident andaccident transients), a certain number of issues require specific R&D
to enable the industrial viability of the design to be confirmed, i.e.
the possibility of well-managed normal operations. Currently, the
SCWR would seem to be relatively complex and ongoing work isaiming, in particular, to simplify the design. In principle, the SCWR
should be able to comply with the major safety principles adopted
for third-generation light water reactors, and the lessons drawn
from the Fukushima accident.

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conclusion

3/

Overall conclusion
In summary, the overview of fourth-generation reactor concepts, as
given in Section 2 above, highlights the following major points.
Sodium-cooled Fast Reactors (SFR)
SFRs benefit from significant industrial experience on theinternational level (in France, the Great-Britain, the USA, formerSoviet-Union countries, Japan, China and India).
While there is no technological obstacle as such for this reactor
type, significant advances are nevertheless required to allow for inservice inspection of safety-related structures (in particular thosethat support the core), which is made difficult by the fact that
sodium is opaque.
Accidents involving fuel damage, or even total core meltdown, havebeen taken into account in the past in the design of a certain
number of SFRs, in particular Phénix and Superphénix, the "RNR
1500" project and the EFR project. For current projects, in particular
ASTRID, how core-melt accidents are to be taken into account has
not yet been discussed, but some aspects would require further
analysis, concerning issues such as the flow of molten materials, thepossibility of keeping it within the reactor vessel, the triggering of a
sodium "vapor explosion" and its extent, and the transfer ofradionuclides from the corium into the primary sodium, thecontainment and the environment.
In any case, the prevention of accidents with severe core damage, inparticular meltdown, should be strengthened, in particular withregard to: scenarios involving reactivity insertion in the core, taking into
account the possibility of unfavorable neutron feedback (apositive "void effect") depending on the core design, as well as© IRSN/2012 – All rights reserveddoc référence
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fuel assembly blockage (an accident which can become
serious very quickly given the high power density in the coreof an SFR); these scenarios should only require limited
countermeasures to be taken, with an adequate grace period
for their implementation

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conclusion

 the scenario of prolonged total loss of the residual heatremoval systems (despite the high thermal inertia provided by
the large quantities of sodium in the reactor); provisions
should be taken to "practically eliminate" this scenario.
The actual possibilities for removing the residual heat by naturalconvection in all or some of the sodium loops (as claimed by thedesigners) will partly depend on the architecture of the systems.
Risks associated with the violent reactions between sodium and air(a sodium fire) or between sodium and water (leading to hydrogenproduction) remain major safety issues for SFRs.
High or Very High Temperature Reactors (V/HTR)
HTRs benefit from significant experience feedback, with power andexperimental reactors having operated in Germany, the GreatBritain and the USA in the past, and currently in Japan and China.
V/HTRs use the most advanced fuel called TRISO: this comprises
spherical fissile particles coated with layers of refractory materials
to ensure containment up to a temperature of 1,600°C. V/HTR
designs plan to use core power densities that are ten times lower
than in PWRs and thirty times lower than in SFRs.
There are no technological obstacles for HTRs. However, to reachhigher temperatures (for VHTRs), more robust materials must be
developed for structures and fuels.
One of the potential risks specific to V/HTRs is a graphite fire, as
large quantities of graphite are used in the reactor as moderatorsand neutron reflectors. This has been the subject of extensive R&D
work for several decades, in particular following the major graphite
fires that occurred at Windscale in 1957 and Chernobyl in 1986.
According to available analyses, a graphite fire could be "practicallyeliminated" on V/HTRs, but clearly this would need to be confirmedon the basis of aspects such as the detailed design options of a
V/HTR and the quality of the graphite used. Furthermore, water
ingress into a V/HTR could cause an increase in core reactivity andproduction of flammable gases via the reaction of water vapor withgraphite. There is past and ongoing R&D on this subject and design
provisions are planned to avoid both risks.

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In terms of the safety demonstration, the available analyses
produced by the designers mean that, given a low power density inthe core (as mentioned above), it is possible to envisage residual
heat removal by radiative heat transfer from the reactor core and avery low risk of fuel cladding damage.

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conclusion

Furthermore, fuel meltdown as such can be excluded, which is afavorable aspect of V/HTRs. However, this observation would need
to be supplemented by a comprehensive assessment of the risk ofreaching core temperatures above 1,600°C, which could lead toextensive damage to the fuel particle coatings, taking into accountthe risks of total cooling system failure, including the reactor pit
cooling system (which may operate passively in some designs).
The dosimetry consequences of deposition in the primary system ofcarbon dust from the core, which could contain radioactive
elements, need to be assessed, in particular with regard to in-service
maintenance (including in-service inspection).
Finally, with reference to the events that occurred on the
Fukushima- Daiichi nuclear power plant, the risk of air or water
ingress into the reactor, in the event of a LOCA or flooding, would
need to be covered.
Gas-cooled Fast Reactors (GFR)
GFRs do not benefit from any experience feedback (as no power orexperimental GFR has ever been built or operated).
High temperatures (over 800°C at the core outlet) are targeted inorder to obtain high power-generation efficiencies. However, there
is currently no fuel that is compatible with such conditions, whichconstitutes a technological obstacle. R&D work is underway.
Compared with SFRs, favorable aspects include the use of an inert
gas (helium) for core cooling, a much weaker "void effect", and thegeneral absence of materials that may react violently with air orwater.
However, in terms of the safety demonstration, prevention of coremelt is a major subject for GFRs. In particular, managing theconsequences of a LOCA is a major safety issue for the design,
which could lead to complex architecture for protection andsafeguard systems, necessarily involving active cooling systems.
Furthermore, these systems would have to be highly robust withrespect to events such as those that occurred on the FukushimaDaiichi nuclear power plant. In addition, water ingress into a GFR
could lead to a reactivity accident and appropriate prevention
would be required.

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Lead-or lead/bismuth-cooled Fast Reactors (LFR)

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conclusion

For LFRs, relatively limited experience feedback has been acquired,uniquely for the reactors on Soviet Alfa-class submarines. Three ofthese reactors suffered serious damage, including one case of partial
core meltdown.
LFRs use molten lead or a lead-bismuth eutectic alloy (called LBE)
as coolant, which have the advantage of not reacting violently with
air or water. There do not seem to be technological obstacles toovercome for LFRs. Management of the risk of corrosion by lead or
LBE is nonetheless a key safety issue, but a procedure exists tomanage this risk, based on maintaining an appropriate oxygen
concentration in the lead and then purifying the lead to remove anyoxides formed. This procedure seems to have been validated during
operations of the submarine reactors mentioned above, although its
suitability for the operation of a large reactor would need to be
confirmed. It should also be stressed that the oxygen concentrationin lead or LBE would need to be carefully adjusted, as the presenceof oxygen (or hydrogen) in lead or LBE increases their viscosity,
which can cause overheating in the core.
Special attention would need to be paid to preventing the lead or
LBE from freezing on an LFR.
In addition, the risk of embrittlement of metal structures by lead or
LBE and their sudden failure is a major issue that has been thesubject of research.
In terms of the safety demonstration, the concerns mentioned
above for SFRs also seem to apply to LFRs, albeit with the favorable
aspect of a lower power density in the core than for an SFR(approximately three times less).
With regard to the events which occurred on the Fukushima-Daiichi
nuclear power plant, although lead has the advantage of notreacting with water, it is toxic and robust design provisions would
need to be taken to prevent releases of lead into the environment,
for example in the event of a break in a loop containing lead or LBE
in a flooded room.
Ultimately, the operation of an LFR seems complex, with the needto manage the risk of reactor overheating due to excessive oxygenor hydrogen (increasing the viscosity of lead or LBE) combined with
the need to manage corrosion risks by using a technique based on
injecting oxygen into the lead or LBE. While this is conceivable foran "overmanaged" operational context or an R&D context© IRSN/2012 – All rights reserveddoc référence
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problematic.

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conclusion

Molten Salt Reactors (MSR)
Experience feedback on MSRs is limited to two experimental
reactors built and operated in the USA in the 1960s: the Aircraft
Reactor Experiment (ARE) and the Molten Salt Reactor Experiment(MSRE), which operated for 13,000 hours.
The CNRS is currently developing the Molten Salt Fast Reactor(MSFR) project, using uranium-233 and thorium-232 dissolved in alithium-based fluoride salt. At the current state of the project, theMSFR seems to benefit from favorable neutronics characteristics.
Reactor shutdown would be provided by fuel salt drainage andpassive residual heat removal from the dump tanks is being
examined.
There does not seem to be a technological obstacle, even thoughcorrosion by the salt is a significant issue that has been the subjectof R&D since the 1950s. There is a procedure for managing this risk,
which seems to have been validated during operations on theAmerican MSRE reactor. Clearly, this procedure would need to bevalidated for larger reactors. Alloys that are compatible with the useof salts at the temperatures targeted for MSRs (700°C) areavailable.
However, given the originality of the MSR design as described by
the MSFR project, associated with the use of a fluid that is both fueland coolant, there would need to be a revision of defense-in-depth
and the safety approaches that have been produced andimplemented for current reactors. Special attention would need tobe paid to characterization of the radioactive contents (in terms ofaspects such as the locations, quantities and spectra ofradionuclides) in the system considered as a whole, including thereactor itself and the reprocessing unit, which will contain fuel salt
and fission products. This analysis should systematically
substantiate the architecture and design of the barriers and identify
risks of criticality, in particular in the event of flooding for example.
Finally, use of beryllium salts would be discouraged, given the
toxicity of beryllium and its characteristics in terms of neutron
moderation (and the consequent risk of criticality).
SuperCritical-Water-cooled Reactors (SCWR)
While supercritical water technology has been used in conventional
industry, no SCWR has ever been built and operated. The SCWR
concept has some very economically attractive characteristics,© IRSN/2012 – All rights reserveddoc référence
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mainly as it can be considered as an innovative version of lightwater reactors, in particular BWRs. SCWRs reuse a certain numberof the general characteristics of these reactors (such as control ofreactivity using absorber rods and emergency borated water, and asafety injection system). However, a certain number of subjects
merit specific R&D before a position can be taken on its industrial
viability. Furthermore, the engineering associated with this design
appears relatively complex and current research is focusing
primarily on subjects that deserve development and is also aiming
to simplify the design. In principle, the SCWR should also be able
integrate the main safety principles that are being developed on
third-generation light water reactors, and experience feedback fromthe accident that occurred on the Fukushima-Daiichi nuclear power
plant, in particular the improvements that may be made to BWRs.

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4/

References

1/ IRSN summary report on GEN IV reactors, 21 November 2007.
www.irsn.fr
2/ "WENRA statement on safety objectives for new power plants",
November 2010 (document faisant suite au rapport Safety
objectives for new power reactors – WENRA Reactor Harmonization
Group, December 2009). www.wenra.org
3/ "Basis for the safety approach for design and assessment ofgeneration IV nuclear systems", GEN IV International Forum, GIF/
RSWG/2007002, revision 1, November 24, 2008. www.gen-4.org
4/ "Stress tests specifications", Proposal by the WENRA Task Force,
21 April 2011. www.wenra.org
5/ "Liquid Metal Cooled Reactors: Experience in Design andOperation", IAEA-TECDOC-1569, December 2007. www.iaea.org
6/ Y. Tobita, H. Yamano, I. Satto, "Analytical study on elimination ofsevere recriticalities in large scale LMFBRs with enhancement of fueldischarge", Nuclear Engineering and Design, Vol. 238, Issue 1,
January 2008, pp. 57–65.
7/ D.A. Powers, B. Clément, R. Denning, S. Ohno, R. Zeyen,"Advanced Sodium Fast Reactor Accident Source Terms: Research
Needs", SAND Report SAND2010-5506, September, 2010.
www.sandia.gov
8/ "Fuel performance and fission product behaviour in gas cooled
reactors", IAEA-TECDOC-978, November 1997. www.iaea.org
9/ Zuoyi Zhang and al., "Current status and technical description ofChinese 2x250 MWth HTR-PM", Nuclear Engineering and Design,Vol. 239, Issue 7, July 2009, pp. 1212–1219.

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10/ "Feasibility of the safety assessment of the coupled system",Rapport IRSN/DSR/SEGRE n° 16 de mars 2011. www.irsn.fr
11/ "Response of fuel, fuel elements and gas cooled reactor cores
under accidental air or water ingress conditions", IAEA-TECDOC784, October 1993.
12/ M.P Kissane, "A review of radionuclides behaviour in the primarysystem of a very high temperature reactor", Nuclear Engineering and
Design, Vol. 239, Issue 12, December 2009, pp. 3076–3091.
13/ "Safety assessment of the design of the modular HTR-2 nuclear
power plant" – Summary – TÜV Hannover/Sachsen-Anhalt, June
1998.
14/ Gas Cooled Fast Reactor – Specific Targeted Research Project –Work Package 1.1 and 4: GFR Design and Euratom – Gen IV –GCFR-DEL-037.
15/ A. Alemberti and al., "European lead fast reactor – ELSY",
Nuclear Engineering and Design, Vol. 214, Issue 9, 2011, pp. 34703480.
16/ L. Cinotti and al., "Lead-cooled system design and challenges in
the frame of Generation IV International Forum – ELSY". Journal of
Nuclear Materials, Vol. 415, 2011.
17/ L. Cinotti and al., "The ELSY Project", http://www.empre.es/20i.pdf
18/ P. Sabharwall and al., "Molten Salts for High Temperature
Reactors: University of Wisconsin Molten Salt Corrosion and Flow
Loop Experiments – Issued identified and Path Forward", INL/EXT10-18090, March 2010. www.inl.gouv
19/ W. Van Howe, "Supercritical Light Water Reactor (SCLWR) with
Intermediate Heat Exchanger – Tractebel Engineering", ICAPP 2007.
20/ GEN IV International Forum – GIF R&D Outlook for Generation
IV Nuclear Energy Systems (2009). www.gen-4.org
21/ "Feasibility study of supercritical light water cooled reactors forelectric power production", Final Report, Idaho National Engineeringand Environmental Laboratory, INEEL/EXT-04-02530, January 2005.
22/ "Supercritical Water Reactor – Survey of materials experienceand R&D needs to assess viability", INEEL/EXT-03-00693,
September 2003.

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23/ X. Cheng, T. Schulenberg, "Heat transfer at supercriticalpressures – Literature review an application to an HPLWR”,
Forschungszentrum Karlsruhe, 2001.
24/ "High Performance Light Water Reactor Phase 2", Public Final
Report, December 2010.
25/ P. Skeldon and al., "The liquid metal embrittlement of 9Cr steel
in sodium environments and the role of non-metallic impurities",
Corrosion Science, Vol. 36, Issue 4, April 1994, pp.593–610.
26/ O. Hamdane, J.-L. Courouau and al., « Caractérisation
mécanique et endommagement des matériaux de structure en
présence de sodium liquide », CNRS and CEA, Paris 12 and 13
January 2011.
27/ « Fragilisation d’aciers par les métaux liquides : phénomènes etmécanismes potentiels, apport de calculs aux éléments finis,
déclinaison aux problématiques de sûreté nucléaire », Report
IRSN/PSN-RES/SAG/2012-00086, publication pending. www.irsn.fr
28/ French Ministers of Industry and Major Risk Prevention, letter to
NERSA, CAB 478 MZ dated 26 October 1990.
29/ Report to the French Ministers of the Environment and Industry– M. Lavérie (DSIN) – DSIN Paris 978/92, 16 June 1992.
30/ "Next Generation Nuclear Plant Phenomena Identification andRanking Tables (PIRTs), Volume 1: Main Report", NUREG/CR- 6944,
vol. 1.
31/ "Passive safety systems and natural circulation in water cooled
nuclear power plants", IAEA-TECDOC-1624, November, 2009.
www.iaea.org
32/ "Fort Saint Vrain Gas Cooled Reactor Operational Experience",
NUREG/CR-6839, ORNL/TM-2002/223, 2003.
33/ K. Krüger and al., "Operational experience and safety
experiments with the AVR power station" Nuclear Engineering anddesign, Vol. 109, Issues 1–2, September–October 1988, pp. 233–
238.
34/ R. Bäumer and al., "Construction and operating experience withthe 300-MW THTR nuclear power plant", Nuclear Engineering anddesign, Vol. 121, Issue 2, 2 July 1990, pp. 155–166.
35/ "Occupational Radiation Exposure at Commercial Nuclear
Power Reactors", NUREG-0594 – 1978.

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5/

What is the SCWR?
// the next logical step in the LWR path
toward simplification
Figure 16.
The SCWR in the context of thevarious light water reactor
designs.

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6/

WENRA statementon safety objectives
for new nuclear
power plants// should be reviewed no later than 2020(see Reference Document [2])Objective O1

Normal operation, abnormal events and prevention of
accidents
 Reducing the frequencies of abnormal events by enhancing
plant capability to stay within normal operation.
 Reducing the potential for escalation to accident situations byenhancing plant capability to control abnormal events.
Objective O2

Accidents without core melt
 Ensuring that accidents without core melt induce no offsite radiological impact or only minor radiological impact (inparticular, no necessity of iodine prophylaxis, sheltering nor
evacuation ).

In a deterministic andconservative approach withrespect to the evaluation ofradiological consequences.
However, restriction of foodconsumption could be neededin some scenarios.

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 Reducing, as far as reasonably achievable:

the core damage frequency taking into account all typesof credible hazards and failures and credible combinationsof events;
the releases of radioactive material from all sources.

 Providing due consideration to siting and design to reduce theimpact of external hazards and malevolent acts.
Objective O3

Accidents with core melt Reducing potential radioactive releases to the environment
from accidents with core melt , also in the long term , byfollowing the qualitative criteria below:


accidents with core melt which would lead to earlylarge releases have to be "practically eliminated" ;orfor accidents with core melt that have not been practicallyeliminated, design provisions have to be taken so that only
limited protective measures in area and time are needed
for the public (no permanent relocation, no need for
emergency evacuation outside the immediate vicinity ofthe plant, limited sheltering, no long term restrictions infood consumption) and that sufficient time is available toimplement these measures.

Objective O4

For new plants, the scope ofthe safety demonstration has
to cover all risks induced by
the nuclear fuel, even whenstored in the fuel pool. Hence,core melt accidents (severe
accidents) have to beconsidered when the core is in
the reactor, but also when thewhole core or a large part ofthe core is unloaded andstored in the fuel pool. It has
to be shown that such
accident scenarios are either
practically eliminated or
prevented and mitigated.
Long term: considering thetime over which the safetyfunctions need to be
maintained. It could bemonths or years, depending on
the accident scenario.
Early releases: situations thatwould require off-site
emergency measures but withinsufficient time to implement
them.

Independence between all levels of defence-in-depth
 Enhancing the effectiveness of the independence between alllevels of defence-in-depth, in particular through diversity
provisions (in addition to the strengthening of each of theselevels separately as addressed in the previous threeobjectives), to provide as far as reasonably achievable an
overall reinforcement of defence-in-depth.
Objective O5

Safety and security interfaces
 Ensuring that safety measures and security measures aredesigned and implemented in an integrated manner. Synergies
between safety and security enhancements should be sought.

Large releases: situations thatwould require protective
measures for the public that
could not be limited in area ortime.
In this context, the possibilityof certain conditions occurring
is considered to have beenpractically eliminated if it isphysically impossible for theconditions to occur or if theconditions can be consideredwith a high degree ofconfidence to be extremely
unlikely to arise (from IAEA
NSG1.10).

© IRSN/2012 – All rights reservedchronologique.
Programmes»:
Les tranches
CP0,
CP1 et CP2,
de
900
le
numéroMWe sont
regroupées dans
respectant
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Objective O6Radiological protection and waste management
 Reducing as far as reasonably achievable by design provisions,for all operating states, decommissioning and dismantling
activities:

individual and collective doses for workers;radioactive discharges to the environment;quantity and activity of radioactive waste.

Objective O7

Leadership and management for safety
 Ensuring effective management for safety from the design
stage. This implies that the licenses:

establishes effective leaderships and management forsafety over the entire new plant project and has sufficientin house technical and financial resources to fulfil its prime
responsibility in safety;
ensures that all other organizations involved in siting,design, construction, commissioning, operation anddecommissioning of news plants demonstrate awarenessamong the staff of the nuclear safety issues associatedwith their work and their role in ensuring safety.

© IRSN/2012 – All rights reserved