Generation-four(GEN-IV)
reactors// Summary report
Version dated 21 November 2007

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Generation-four(GEN-IV) reactors// Summary report

First edition 21 November 2007
IRSN report 2007-91
ISRN IRSN 2007/91 FR + ENG© IRSN/2008 --- All rights reserveddoc reference
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IRSN
//Short view
The Institute for Radiological Protection and Nuclear Safety (IRSN),
created by law no 2001-398 of May 9, 2001, then by decree no
2002-254 of February 22, 2002, is a public establishment of anindustrial and commercial nature (EPIC), under the joint authority of
the Ministers of Defense, the Environment, Industry, Research and
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IRSN employs more than 1,500 specialists, including engineers,
researchers, doctors, agronomists, veterinarians and technicians,
experts in nuclear safety and radiological protection and in thecontrol of nuclear and sensitive materials.
The Institute exercises specialist and research assignments in thefollowing fields:•

nuclear safety;•

safety in transporting radioactive and fissile materials;•

protection of man and the environment against ionisingradiation;•

protection and control of nuclear materials;

Doctrine & synthèse•

physical protection of facilities and transport of radioactive
and fissile materials.

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information, please contact:

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FranceFax : +33 1 58 35 79 62

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The Institute for Radiological Protection and Nuclear Safety
conducts research programmes and studies on nuclear andradiological risks; it is responsible for public service activities in theprevention of these risks and provides technical support to thecompetent public authorities in nuclear and radiological protection
safety and security. In this respect, the Institute is called on todevelop a position on a certain number of scientific and technicalissues.
Under its policy of transparency and its desire to make high-quality
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They reflect the IRSN position on the day of their publication on its
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This document may be used and quoted freely on condition thatthe source and publication date are mentioned.
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Doctrine & synthèse
IRSN
BP 17
92262 Fontenay aux Roses
cedex
FranceFax : +33 (0) 1 58 35 79 62

doc.syn@irsn.fr

Jacques Repussard
Director General© IRSN/2008 --- All rights reserveddoc reference
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5/52//List ofauthors
Jean Couturier

Chairman

Olivier Baudrand
Giovanni Bruna
Valérie Bruno
Chantal Casselman
Marie-Paule Elluard
Jacob MarcianoLaurence Martinet
Jean-Marc DormantJoëlle Papin
Véronique Rouyer
Christine Galet
François Besnus© IRSN/2008 --- All rights reserveddoc reference
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FOREWORD
Preparation for the fourth generation (GEN-IV) of nuclear reactors
At the December 20th, 2006 meeting of the Council of Ministers, theministers in charge of research and industry gave a presentation on the
fourth generation (GEN-IV) of nuclear reactors.
In the words of the press release, “France has decided to commit itself
resolutely to participate in the design of the fourth generation of nuclear
reactors, the industrialization of which could start from the year 2040.
The industrial design of these reactors must satisfy several requirements,
fixed by the Government, relative to the third generation (GEN-III)reactors:•

reducing the volume and radiotoxicity of waste produced;•

generating the same quantity of energy while using a lower amountof uranium;•

further improving the safety and security of reactors;•

reducing the proliferation risks.

… (The research to be performed over the coming years) will focus on thesodium - or gas-cooled - fast reactor concepts …
A milestone has been set for 2012 to decide on the technological choicesfrom among the different technological options and to undertake theconstruction in France of a prototype of a GEN-IV reactor, with a view tocommissioning in 2020, in accordance with the decision taken by the
President of the French Republic in January 2006”.
It is part to the IRSN to give technical advice on the protection of
persons, property and the environment and to carry out research
programmes aimed at maintaining and developing the knowledge which
is necessary for the expertise of industrial projects in the fields of nuclear
security in its widest meaning (safety, radiological protection, nonproliferation, resistance to acts of malicious damage).
In this framework, the IRSN initiated internal reflections on safety,
radiological protection and security-related issues affecting the six
reactor concepts selected, at the international level, by the “Generation© IRSN/2008 --- All rights reserveddoc reference
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IV International Forum (GIF)” created in 2000 at the initiative of theAmerican Department Of Energy (DOE). This think-tank, which started
in 2004-2005, was upgraded, following the announcement by the
President of the French Republic in early 2006, to engage the country inthe construction of a GEN-IV prototype by 2020. In addition to powerreactor concepts, the IRSN's analysis has been extended to theassociated fuel cycles.
Taking into account the government orientations, the IRSN decided to
put forward the present state of its thinking on the six reactor concepts
mentioned above, it being understood that the subsequent work of theInstitute would mainly focus on fast reactor safety. The hightemperature and very high temperature reactors, not retained due totheir "open" cycle (where actinides are not recycled), will for their partbe the subject of an active watch in conjunction with industrial partners
interested in their development.
The fast reactors so far selected call on two different concepts:•

The sodium-cooled fast reactors: this concept has already been built
up to an industrial scale in France with Rapsodie and Phénix as well as
the Creys-Malville power plant; subsequent projects (RNR 1500, EFR)
underwent preliminary analysis either within a formal nationalframework (RNR 1500) or within an international informal framework(EFR). It is clear that some improvements of the concept, likely relevant
ones, will be proposed by designers with the aim of improving its
competitiveness, safety, security and waste management performance.
•
The gas-cooled fast reactors: this concept has not yet been worked
on either in France or abroad. Its selection in the framework of GIF islinked to its features for actinide transmutation and its highthermodynamic efficiency (the target temperatures allow considering
the production of heat for industrial processes). Future developments ofthe concept will first require the construction of an experimental reactor(the current REDT project of the French Atomic Energy Commission CEA- is a 50 MWth reactor; the decision for its construction, if any,
would not be taken by 2012).

Concerning the sodium-cooled fast reactors, the text herewith attachedemphasizes a number of specific issues, which require in-depth
examination between now and 2012. The main concerns are presently
the features of the reactor core as well as its behaviour in bothincidental and accidental situations, in relation to the choice of coolingdevices, and the in-service monitoring of the various components(internal structures, steam generators). It is clear that the design
improvements proposed by designers should, in due time, be the subject© IRSN/2008 --- All rights reserveddoc reference
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of in-depth discussions (e.g. giving up the intermediate sodium circuits).
Nevertheless, the actual state of knowledge appears sufficient for abrainstorming to be launched in 2007 on the main safety issues to beaddressed for an industrial size reactor, as it was done for the 3rd PWR
reactor generation at the end of the 1980s (the studies, conductedjointly by IPSN and GRS, led, in mid-1993, to a common statement by
the French and German Safety Authorities on the safety objectives to be
searched for the EPR reactor project). The proposed think-tank will take
into account the results of the analysis carried out for the RNR 1500 and
EPR projects. It will include discussions with other expertise assessment
bodies at the international level. Moreover, it will contribute to defining
the research work in support of the future expertise of a reactor project.
It is worth stressing that, if experiments in “large” research facility were
to be carried out, they should be defined and scheduled in the coming
years in order to match the 2020 milestone.

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The IPSN (Institut de Protection
et de Sûreté Nucléaire) became
the IRSN in 2002.

Concerning the gas-cooled fast reactors, the concept design –even foran experimental reactor- is not achieved enough to adopt a similar
approach in the same time frame. It is essential for designers to perform
preliminary feasibility studies with particular attention to the safety
questions already identified (the main concern being the core cooling in
accidental situations).
In the meantime, the IRSN will also conduct a brainstorming on
security-related issues (non-proliferation and protection against acts ofmalicious damage) for both concepts; these problems must be stressedfrom the very beginning of the design phase so that the overall plantdesign accounts at best for all the nuclear security-related aspects.

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Contents
1/ Introduction

11

2/ Features and industrial maturity of the six Generation-Four(GEN-IV) reactor concepts
13
2/1 Helium-cooled High Temperature and Very High Temperature Reactors(HTR/VHTRs)
13
2/2 Sodium-cooled Fast Reactors (SFRs)
16
2/3 Gas-cooled Fast Reactors (GFRs)
19
2/4 Lead or lead-bismuth cooled Fast Reactors (LFRs)
20
2/5 Supercritical Water Reactors (SCWRs)
22
2/6 Molten Salt Reactors (MSRs)
23
2/7 Summary and generalities
24

3/ Main safety, radiological protection and security issues
identified to date for High Temperature and Very High
Temperature Reactors (HTR/VHTRs)
3/1 Safety objectives and approach
3/2 Reactor-related aspects

26
26
27

3/2/1 Fuel27
3/2/2 Neutronics
28
3/2/3 Materials
29
3/2/4 Passive Systems
30
3/2/5 Safety and reliability of the associated industrial processes30
3/2/6 Other graphite-related risks
30
3/2/7 Radiological protection
31
3/2/8 Fire and explosion risks related to hydrogen production 31
3/2/9 Security concern
31

3/3 Fuel fabrication and processing related aspects

4/ Main safety, radiological protection and security issues
identified to date for Sodium-cooled Fast Reactors (SFRs)
4/1 Safety objectives and approach
4/2 Reactor-related aspects

32

35
35
36

4/2/1 Specific risk inherent to metallic fuel36
4/2/2 Neutronics effects in case of loss of coolant – the so-called“sodium void effect”
36
4/2/3 Risk of core meltdown37
4/2/4 Sodium-related risks
38
4/2/5 Passive systems
39
4/2/6 Structure Inspectability
39
4/2/7 Security concern
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4/3 Fuel fabrication and processing related aspects

5/ Main safety, radiological protection and security issues
identified to date for Gas-cooled Fast Reactors (GFRs)
5/1 Safety objectives and approach
5/2 Reactor related aspects
5/3 Fuel fabrication and processing related aspects

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41

44
44
44
46

6/ Waste production and management in the framework of the
deployment of a Generation-Four reactor fleet
47
7/ Conclusion - Actions undertaken and planned by the IRSN50

List of Figures
Figure 1: Left: a reactor pebble. Right: ‘‘compact’’ and fuel block for hightemperature or very high temperature reactors................................................................. 14
Figure 2: A TRISO particle. Refracting layers are shown in blue.................................. 14
Figure 3: The Chinese HTR10 reactor using pebbles. ....................................................... 15
Figure 4 The Japanese HTTR reactor. ..................................................................................... 15
Figure 5: The Phenix reactor hall is accessible during operation. The reactor inoperation since 1973 will be shut down in 2009................................................................. 17
Figure 6 Schematic diagram of an integrated type sodium-cooled fast reactor. . 18
Figure 7: Schematic diagram of a lead-cooled fast reactor........................................... 20
Figure 8: The BREST project. ...................................................................................................... 21
Figure 9: Schematic diagram of a supercritical water reactor. ..................................... 22
Figure 10: Schematic diagram of a reactor in which the fuel is dissolved in molten
salts...................................................................................................................................................... 23
Figure 11: Heterogeneities in a very high temperature reactor core. ........................ 28
Figure 12: AREVA very high temperature reactor project. ............................................. 29
Figure 13: Sectional view of a TRISO particle. Photo credit: CEA................................ 32
Figure 14: TRISO HTR fuel balls fabricated in the UO2 laboratory’s GAIA facility at
the Cadarache CEA centre. Photo credit: CEA.................................................................... 33
Figure 15: Schematic diagram of the Ganex process studied by the CEA for thegrouped extraction of actinides in the framework of studies on generation-four
reactors closed cycle (FP: fission products; An: actinides; Ln: lanthanides). Photo
credit: CEA ......................................................................................................................................... 42
Figure 16: Example of fuel developed for GFRs.................................................................. 45
Figure 17: Schematic diagram of a gas-cooled fast reactor (reactor vessel,
components vessel, intermediary containment)................................................................ 45© IRSN/2008 --- All rights reserveddoc reference
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1/ Introduction

1/
Introduction
Created in 2000 at the initiative of the American Department Of Energy(DOE), the "Generation-Four International Forum" (GIF) today associates
twelve countries, including France and the community of states which
signed the Euratom Treaty, for cooperation in the development of new
nuclear systems (reactors and fuel cycle facilities) allowing the
production of energy beyond the life of reactors in operation or underconstruction (including EPR).
In this regard, the current plans of Electricité de France (EDF) include thedevelopment of new reactor concepts which could be industriallydeployed around 2040 (Generation-Four, GEN-IV), after extending theoperation of current reactors of GEN-II (current PWRs) andcommissioning the "evolutionary" reactors of GEN-III (EPR); 2040 also
matches the planned schedule for the renewal of the fuel cycle
reprocessing plants. This implies, for EDF, commitment to an industrial
prototype of the Generation-Four around 2015-2020.
The first phase of the work of GIF resulted, in 2002, in the publication ofa development plan for those technologies considered as the mostpromising. Six reactor concepts were selected, which should allow forsignificant advances in economic competitiveness, safety, savings inuranium resources, reduced radioactive waste production, especiallyhigh activity long-lived ones, and resistance to proliferation and acts ofmalicious damage. These reactor concepts are as follows:•
Helium-cooled High Temperature and Very High Temperature
Reactors (HTR/VHTRs),•

Sodium-cooled Fast Reactors (SFRs),•

Gas-cooled Fast Reactors (GFRs),•

Lead or Lead-bismuth cooled Fast Reactors (LFRs),•

Supercritical Water cooled Reactors (SCWRs),•

Molten Salt Reactors (MSRs).

The USA, Japan and France are more involved in the gas-cooled (VHTR& GFR) and sodium-cooled (SFR) reactor concepts. Coordination of GIFled actions is taken on by:© IRSN/2008 --- All rights reserveddoc reference
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1/ Introduction•

France for HTR/VHTRs,•

Japan for SFRs,•

the USA for GFRs.

In France, the company AREVA is carrying out the project of anindustrial VHTR with a thermal power of about 600 MW, named
ANTARES (Areva’s New Technology and Advanced gas-cooled Reactor
for Energy Supply), and the French Atomic Energy Commission (CEA) is
carrying out the GFR and SFR concept studies. AREVA intends to bid fora possible call for tender by the DOE for a VHTR, the "Next Generation
Nuclear Plant" (NGNP), which should be built, between 2015 and 2020,
as a demonstrator, on the site of the Idaho National Laboratory in theUSA. On its side, several years ago, the CEA announced intention tocommission, by 2015, an Experimental Reactor for Technology
Development (REDT in French) of the GFR type; this reactor should
generate a power of 50 MWth.
Moreover, the following activities are carried out with a contribution
from the European Commission:•a VHTR integrated project, named RAPHAEL ((ReActor for Process
heat, Hydrogen And ELectricity generation),•a wide research and development activity in support to the GFR
concept design (known as GCFR project).

In January 2006, the President of the French Republic set the deadline of2020 for commissioning a GEN-IV reactor prototype. Moreover, article
no 3 of the June 28th 2006 law on the radioactive material and waste
sustainable management indicates that "the studies and research (onthe separation and transmutation of long-lived radioactive elements) areconducted in conjunction with those carried out on the new generationof nuclear reactors mentioned in article 5 of programme law no 2005781 of July 13th 2005 establishing the orientations of French energy
policy, as well as on waste transmutation dedicated accelerator-driven
systems, in order to have an assessment of the industrial perspectives ofsuch reactor concepts by 2012 and to put a prototype into operation
before 31 December 2020”.
In this context, this report gives an overview of the main features of thereactor concepts selected by the GIF and presents the main safety,
radiological protection and security challenges raised by concepts which
are most likely to be built in the indicated time frame, due either to theirindustrial maturity, or to a significant interest in their capacity to matchthe above-listed objectives (savings in uranium resources, etc.). Lastly, it
describes the R&D actions undertaken or planned by the IRSN, identified
in terms of safety, radiological protection and security.

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2/ Features andindustrial maturityof GEN-IV reactor
concepts2/
Features and industrial
maturity of the six
Generation-Four (GEN-IV)
reactor concepts
2/1

Helium-cooled High Temperature and VeryHigh Temperature Reactors (HTR/VHTRs)
HTR/VHTRs are graphite-moderated, thermal reactors; the heatgenerated in the core is extracted by a gas (helium) under pressure.
Several thermodynamics cycles for electricity generation are proposed
which would allow very high efficiencies (nearly 50 %, compared toabout 35 % for PWR reactors).
The VHTR concept is an evolutionary stage of HTR reactors, which werebuilt in the 1970s and 1980s, mainly in the USA and Germany. We canmention:•Research reactors from 15 to 40 MWe, such as the Peach Bottom
reactor (USA) and the AVR reactor (Germany),•
Reactors of about 300 MWe, such as the Fort Saint Vrain reactor(USA) and THTR reactor (Germany).

The main significant safety-related events which have affected the HTR
operation are as follows:•

In the Fort Saint Vrain reactor, problems of neutronic instability
induced by the movements of graphite blocks in the core and infiltrationof water in the primary circuit at the level of the moto-blowers,•

In the THTR, ruptures of thermal insulation fixing elements on acore outlet pipe.

The objective for VHTRs is reaching an average coolant temperature ofabout 1000°C at core outlet, compared to 750°C -850°C for HTRs.
VHTRs could thus be a supplier of heat for industries with high energy© IRSN/2008 --- All rights reserveddoc reference
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2/ Features andconsumption, in particular regarding the production of hydrogen throughadvanced processes, some of which, such as the high temperatureelectrolysis, and the iodine/sulphur cycle, are still under development.

industrial maturityof GEN-IV reactor
concepts

Two geometric designs are proposed for the fuel elements: the pebbles
and the rod « compacts ».
Figure 1:
Left: a reactor pebble.
Right: “compact” and fuel blockfor high temperature or very hightemperature reactors.

However, in both cases, the nuclear material is confined in a sphericalmicro-particle named TRISO (TRistructural ISOtropic), made of a central
kernel, coated with three layers of refractory material.
Figure 2:
A TRISO particle.
Refracting layers are shown inblue.

More recently, two experimental reactors have been commissioned: theChinese 10 MWe HTR 10 reactor, built near Beijing, which went criticalin 1998, and the Japanese VHTR 30 MWe HTTR reactor, built on the
Oarai site, which achieved criticality in 2001. The HTTR is equipped witha gas turbine; a hydrogen production process facility was installed in2005 and is at present undergoing testing.

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2/ Features andindustrial maturityof GEN-IV reactor
concepts
Figure 3:
The Chinese HTR10 reactor using
pebbles.

Figure 4:
The Japanese HTTR reactor.

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2/ Features and

The fuel cycle of the HTR/VHTR concept is "open", but preliminary
studies for "closing" it by recycling the fuel in the reactor have been
performed and are now still underway within a European R&D
framework.

industrial maturityof GEN-IV reactor
concepts

The HTR/VHTR concept has reached a stage of maturity allowing theconstruction, by 2020, of an industrial prototype operated through an"open" cycle. South Africa is already developing a pebble bed reactor
project (PBMR) of about 100 Megawatts of electrical power, with ahelium turbine; its announced commissioning objective is around 2010.
2/2

Sodium-cooled Fast Reactors (SFRs)
SFRs are reactors with a fast neutron spectrum, which therefore containno moderator; the heat generated in the core is extracted by a liquid
metal (sodium). The fast neutron spectrum enables the efficienttransformation of natural uranium (U238), the ‘‘fertile’’ material, into‘‘fissile’’ plutonium with a ‘‘conversion ratio’’ (number of plutonium
nuclei generated per fission) close to 1 (that’s the SFR breeding-mode)or, contrariwise, a consumption of plutonium (the SFR incineratingmode); it also enables a transmutation of very long-lived actinides. The
temperature reached by the sodium (in the range of 550°C) allows aquite high thermodynamic efficiency (about 40 %).
The reference fuel is generally mixed uranium and plutonium oxide; theuse of mixed carbide or nitride is also planned. Moreover, SFR concepts
are characterised by a great thermal inertia due to the large amount of
sodium present in the primary circuit, which allows the operators a quite
long response-time to intervene in the event of failure to remove theheat, as well as a margin of more than 300°C before the boiling of
sodium in normal operation. Moreover, sodium has a high capacity fortrapping some fission products in the event of damage to the fuel.
SFRs are operated through a ‘‘closed’’ fuel cycle, which allows in reactor
recycling of uranium, plutonium and minor actinides. The industrial
feasibility of reprocessing the SFR fuel (mixed uranium and plutoniumoxide) with the current PUREX process has already been demonstrated in
the reprocessing plant of La Hague. Several tonnes of fuel unloaded fromthe Phénix reactor diluted with fuel irradiated in the UNGG reactors(natural uranium graphite gas cooled reactors) have been successfully
reprocessed in the UP2-400 factories at La Hague. Moreover, therecycling of plutonium and minor actinides in SFRs has been tested inexperiments carried out in Superphénix (CAPRA programme) and in
Phénix (PAVIX experiment) reactors.

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2/ Features andSFRs have the advantage of a large experience, via the operation ofseveral power reactors:•

industrial maturityof GEN-IV reactor
concepts

In France, the experimental Rapsodie reactor (initially 25 MWth,later upgraded to 40 MWth), then the Phénix (250 MWe) andSuperphénix (1240 MWe) reactors,
Figure 5:
The Phenix reactor hall isaccessible during operation. The
reactor in operation since 1973
will be shut down in 2009.

•

In Britain, the PFR reactor (135 MWe),•In the states of the ex-Soviet Union, the BN 350 (135 MWe) andBN 600 (550 MWe) reactors; the construction of the BN 800 reactor,
which had been postponed for several years has been put again on theagenda,•In Japan, the experimental Joyo reactor, the power of which reached
140 MWth, and the 280 MWe Monju reactor, the resumption ofoperation of which, interrupted after a sodium fire in 1995, is planned forabout ten years from now.

Some of the significant events which have affected SFRs (Phénix,Superphénix, PFR, and Monju) are mentioned below in section 4/2. These
events have diverse origins, such as inappropriate operator reactions,design errors, inhibited security systems, insufficient performancerequirements for some items of equipment, difficulty in mastering
performance due to the complexity of the industrial organization.
Moreover, it must be emphasized that the Phénix reactor underwent, in
1989 and 1990, automatic shutdowns due to a sudden decrease inreactivity, the origin of which remains unexplained and is still underinvestigation.
India is currently building a 500 MWe prototype fast breeder reactor(PFBR) while China is building an experimental reactor, the CEFR, with apower of 25 MWe.

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2/ Features and

In France, in the 1980s and after the commissioning of Superphénix,studies were carried out on a 1500 MWe SFR reactor project (named
RNR 1500), with an increased breeding capacity and lower neutronics
sensitivity to the loss of coolant, including ‘‘fertile’’ or even non fuelzones in the inner core. Such studies were pursued in the European
framework, associating the British, the Germans and the French on an
SFR power reactor fleet project (EFR project).

industrial maturityof GEN-IV reactor
concepts

Two different designs are proposed for existing and future SFRs: the‘‘integrated-type’’ concept (Phénix, Superphénix, PFR, CEFR), the primarycircuit of which is confined in the reactor vessel (which contains boththe primary pumps and the heat exchangers), and the ‘‘loop type’’concept (Joyo, Monju) the primary sodium of which circulates in loops
linking the main reactor vessel to secondary reactor vessels containing
the large components.
Figure 6:
Schematic diagram of an
integrated type sodium-cooled
fast reactor.

The search for cost reductions should lead the designers to suggest
innovative solutions to simplify the systems and their equipment.
The SFR concept has reached a stage of comfortable maturity, whichmakes the construction of a new industrial prototype be likely for 20152020.

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2/ Features and2/3

Gas-cooled Fast Reactors (GFRs)

industrial maturityof GEN-IV reactor
concepts

GFRs are fast neutron spectrum reactors, the heat of which, produced in
the core, is extracted using a gas (helium) under pressure. In principle,
GFRs enable actinides to be transmuted thanks to their fast neutron
spectrum, while remaining by nature insensitive in the event of loss of
coolant and ensuring high thermodynamic efficiencies and, possibly, theproduction of industrial heat (helium temperature of at least 850°C innormal operation).
There is no experience of operating GFR reactors. Nevertheless, theconcept is not totally new: studies have been engaged as early as 1962
in the USA (GCFR project) and since 1968 in Europe (GBR project), in
which the French Atomic Energy Commission (CEA) has been involved.
The GFR concept evaluated in the framework of GIF is being developedby the CEA (project for potentially 300 to 1200 MWe of power). It
includes an entirely refractory core (fuel contained in a ceramicmaterial), able to confine the very high temperature nuclear materials.
The IRSN has no detailed documents describing the design of a GFR andin particular the core design, the barriers and the main equipment. The
project is still in the preliminary definition phase and only generaldescriptions are available. Thus, the nature of the fuel has not alreadybeen precisely defined (oxide, carbide or nitride, with a refractory
coating), but the oxide solution seems to be excluded due to its too low
density in heavy nuclei (for the same power outlet, a GFR needs agreater free volume for the coolant than a SFR, due to the lower heat
extraction capacity of gas compared to sodium, thus a fuel with a higher
fissile kernel density is necessary).
A few experiments have already been performed in the Phénix reactor,
on a nitride based fuel (NIMPHE experiments), and further experiments
were loaded in the reactor in 2007 (FUTURIX-MI experiments on
structure materials, FUTURIX-CONCEPT experiments on the fissile
material and cladding, FTA-nitride experiments).
The GFR fuel cycle is "closed", with the possibility of recycling theuranium, plutonium and minor actinides. It should be emphasized that
GFRs were the reference system selected by the CEA in the framework ofaxis 1 of the 1991 law relative to the research for the management ofhigh activity and long-lived waste (partitioning and transmutation).
The development of the GFR concept would include, in the first phase,the commissioning of a low power experimental reactor. The
construction of such a reactor implies the solution of several
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2/ Features anddevelopment, and the assessment of the safety aspects to be addressed.
Commissioning by 2020 of a demonstration reactor, such as the REDT,
currently studied by the French Atomic Energy Commission, seems quite
optimistic.

industrial maturityof GEN-IV reactor
concepts

2/4

Lead or lead-bismuth cooled Fast Reactors(LFRs)
LFRs are fast reactors cooled by a molten metal, such as lead or a leadbismuth alloy.
Figure 7:
Schematic diagram of a leadcooled fast reactor.

The fuel cycle of the concept is “closed”.
One advantage of lead or of a lead-bismuth alloy compared to sodium isthe absence of any chemical reaction with water and air.
The Russian BREST 300 (300 MWe) is the most industrially advanced
project of a LFR in the world. In this design, the lead is heated up to atemperature of 550°C.

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2/ Features andindustrial maturityof GEN-IV reactor
concepts
Figure 8:
The BREST project.

This concept is derived from Russian Alpha class submarine reactors,
cooled with a lead-bismuth alloy, which were abandoned in favour of
PWR reactors. The last submarine of this class was disarmed in 1995.
The submarines equipped with LFRs suffered from huge maintenanceproblems, especially due to the presence of lead activation products.
According to available information, serious failures occurred in two ofthem.
Several design choices on the BREST 300 project, notably a moderate
power density, allow a satisfactory fuel behaviour, even in the event ofaccident with protection system failure, thanks to the passive cooling(natural convection).
The main difficulty of LFRs is due to the highly corrosive nature of themolten lead on the steel structures. The only known preservation
method consists in creating, then maintaining, a protective layer of ironoxides on the surface of the steel structures in contact with the lead.
That involves injecting oxygen into the lead and scrubbing the leadoxides and the corrosion residues present in it. This method wasdeveloped for the operation of submarine reactors, but deviations in its
application originated severe failures in two vessels (blocking, assembly
melting) and one of them sank.
Moreover, the BREST 300 project energy conversion system has not yet
been defined. In the present project, the reactor, which is of theintegrated type, has no intermediate circuit (to minimize the costs). The© IRSN/2008 --- All rights reserveddoc reference
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2/ Features andsecondary circuit would use supercritical water at a pressure of 250 bar;
this would have the disadvantage of positioning equipment under highpressure near the reactor core, resulting in no appreciable gain, in termsof thermodynamic efficiency, compared to a design with an intermediate
circuit.

industrial maturityof GEN-IV reactor
concepts

Spent fuel processing methods (nitride or metallic type) as well as the
fabrication of fuels containing minor actinides remain to be developed.
Moreover, due to its activation, the lead-bismuth type coolant would
contain long-lived alpha emitting radioactive waste.
Thus, in the light of experience feedback and the present state ofknowledge, a deployment of LFR reactors on an industrial scale is not
likely within the considered time frame. In any case, the construction of
the first experimental reactor would require the solution of several
technological challenges and the solution of some major safetyconcerns.
2/5

Supercritical Water Reactors (SCWRs)
The concept design of supercritical water reactor is derived fromexperience from the thermal electricity production plants which haveadopted this technology. The use of water in supercritical state, at a 250
bar pressure and a temperature of about 550° C, increases the plant
efficiency significantly, compared to a pressurized water reactor: in a
direct thermodynamic cycle, it can be as high as 44%.
Figure 9:
Schematic diagram of asupercritical water reactor.

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2/ Features and

The concept developed by Westinghouse and sponsored by the DOE is
also minded to the search for lower investment costs than for PWR
reactors (ruling out the secondary circuit).

industrial maturityof GEN-IV reactor
concepts

Supercritical water reactors have not passed the stage of not convincing
feasibility studies (e.g., the first core calculations performed byWestinghouse were based on an assembly with more than tenenrichment zones).
An industrial deployment of supercritical water reactors is unlikely within
the considered time frame and the construction of the first experimental
reactor does not seem foreseeable by 2020.
2/6

Molten Salt Reactors (MSRs)
Two different MSR concepts are proposed: those in which the molten
salt only acts as a coolant, and those the fuel of which is dissolved in themolten salt. In the latter, the molten salt contains a mixture of thorium(a ‘‘fertile’’ material) and uranium 233 (a ‘‘fissile’’ material). Such reactors
allow the fission of uranium 233 produced from thorium, so that an
initial load of uranium 233 or plutonium is required at start-up. The fuelcycle of MSRs is ‘‘closed’’: on line extraction of transuranics elements andfission products, then uranium 233 recycling.
Figure 10:
Schematic diagram of a reactor
in which the fuel is dissolved inmolten salts.

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2/ Features andTwo experimental MSRs, with fuel dissolved in the molten salt, havealready been built and operated in the USA. The first, intended formilitary aeronautic propulsion, was constructed during the 1950s in theframework of the ‘‘Aircraft Reactor Experiment’’ project. The second is
the MSRE (‘‘Molten Salt Reactor Experiment’’), constructed in 1962 at
Oak Ridge, which went critical in June 1965; it did not use any ‘‘fertile’’material (thorium), but uranium 235-based fuel, then uranium 233; thisreactor, which delivered power of 8 MW, was shut down in 1969 after
approximately 13,000 hours of operation.

industrial maturityof GEN-IV reactor
concepts

From an industrial point of view, the designers of molten salt reactors
claim that this concept could be operated at high temperatures without
requiring high pressures, with smaller diameter lines than those ofhelium cooled reactors (which require a much higher coolant flow rate
for the same amount of energy extracted from the core). This should be
advantageous if the reactor is coupled to a hydrogen industrial
production facility.
Molten salt reactors have not passed the experimental stage thresholdand leave complex problems open as for controlling the risk of corrosion,
which should probably push R&D towards non-metallic materials.
Moreover, a specific fuel reprocessing unit should be associated withsuch a reactor.
The construction of a molten salt reactor by 2020 seems unlikely.
2/7

Summary and generalities
As regards the three concepts for which the implementation of anexperimental reactor or an industrial prototype is foreseeable by 2020
(i.e. HTR/VHTR, SFR and possibly GFR), the IRSN reviewed the technicalchallenges identified to date in terms of safety. They are summarized in
the following paragraphs, as well as some elements concerning theradiological protection.
Concerning security, designers have not so far provided precise
information in terms of proliferation resistance or protection against
malicious damage (reactor containment barrier resistance, access rightto different zones in the plant, etc.), nor wondered about specificsensitivities for the concepts under development. Moreover, while for
SFRs (Superphénix, Monju loop reactor, EFR project, etc.) and to a lesser
extent for HTR/VHTRs it is possible to refer to detailed designs, atpresent no GFR project design is sufficiently advanced to allow going
beyond general thoughts in terms of security.
Among the generic aspects which will need to be considered for all the
GEN-IV concepts, those concerning the design and the containment© IRSN/2008 --- All rights reserveddoc reference
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2/ Features andresistance to external aggressions, which could result from possible
malicious actions, are of major importance; they mainly concern thereactor buildings.

industrial maturityof GEN-IV reactor
concepts

Moreover, the physical protection devices and the monitors, whichenable nuclear materials to be constantly located in order to prevent or
detect their theft, loss or diversion need to be adapted depending onwhether they apply to plants operated through an ‘‘open’’ (in the case of
HTR/VHTRs) or a ‘‘closed’’ cycle (fast reactors). In the latter case,especially if advanced fuel separation is planned for recycling actinides,
particular attention should be paid to the design of physical protection
devices.
Some remarks specific to the HTR/VHTR, SFR and GFR concepts are
outlined in the following paragraphs.
Finally, in what follows, the IRSN will not discuss the specific problems
connected to the declared lifetime objectives for GEN-IV reactors;
several documents mention life-times of 60 years and more, which
implies an appropriate certification for materials and equipment which
could not be replaced during service.

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3/ Main issues for
HTR/VHTRs

3/
Main safety, radiologicalprotection and security issues
identified to date for High
Temperature and Very High
Temperature Reactors(HTR/VHTRs)
3/1

Safety objectives and approach
While the major safety and radiological protection objectives for the
GEN-IV reactors must be at least as ambitious as those defined for the
EPR project, their operational application cannot, at present, be specified
in details without better knowledge of the systems considered. It must
be borne in mind that, for the EPR project, the preliminary definition of
the safety objectives was set out as early as 1993, within a few months
because the PWR concept had been investigated in-depth by theconcerned parties, including the IRSN, since the beginning of the Frenchelectro-nuclear programme in 1973, and that the EPR project had beenplaced from its very beginning in the continuity of the current design ofreactors in operation.
Nevertheless, in March 2004, the French Atomic Energy Commission
submitted to the Nuclear Safety Authority a document, drawn up incollaboration with AREVA NP, presenting the ‘‘first ideas’’ on a ‘‘generic’’safety approach for the GEN-IV systems and, more particularly, VHTRs
and GFRs (letter CEA/DEN/DDIN/SF DO 73 of 18 March 2004). This
document mentions the simultaneous use of deterministic andprobabilistic methodologies, e.g. a ‘‘risk informed’’ type approach, as wellas the ‘‘lines of defence’’. An evaluation of this safety approach would
require knowledge of its main applications to VHTR and GFR concepts.

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3/ Main issues for

We can note here the following safety objectives put forward in the
preliminary CEA's document:

HTR/VHTRs•

No need for emergency measures outside the reactor site "for any
accidental situation" (more detail needed on the definition of such asituation),•‘‘Practical elimination’’ of very highly degraded situations with astrong likelihood for radioactivity release, in particular severe generalised
core damage (for example, core meltdown).

Concerning VHTRs, the designer's safety approach extensively exploits
the physical properties of the fuel and core materials, which would allow
controlling both normal operation and cooling system failures, without
requiring intervention of active systems. That would be a significant
advance compared to current pressurized water reactors which haveseveral active systems devoted to control the core loss of coolant. Two
aspects should be analyzed carefully:•

The major part of the first containment barrier robustness (thelayers coating the fuel kernel),•

The wish of designers to adapt the GEN-III third barrier to GEN-IV
system, allowing for the VHTR physical features, especially their highthermal inertia, resulting in slow temperature changes during loss of
coolant transients.

3/2

Reactor-related aspects
3/2/1

Fuel
Designers largely found the safety demonstration of the HTR/VHTR
concepts on the resistance of the TRISO particle, which is assumed to beable to maintain its containment capacity up to a temperature of about
1,600°C, even though it is well known that the refractory layers of thefuel would become permeable to some fission products after a longirradiation period at this temperature. Designers aim therefore at
guaranteeing that this temperature would never be reached whatever
the accidental transient. In this respect, higher temperatures during
normal operation in VHTRs (compared to HTRs) would involve a careful
examination of the demonstration of the resistance of the fuel envelope,
especially in the case of a primary loss of coolant transient. This need
could lead designers to search for other coating materials (zirconium
carbide for example), which would resist higher temperatures. Besides, if
in the near future attaining high burn-up is not necessarily a design
objective for an experimental, a demonstration, or even a prototype© IRSN/2008 --- All rights reserveddoc reference
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3/ Main issues forreactor, such an objective is intrinsically related to the more generalobjective of GEN-IV reactors which is to replace the reactors currently inoperation or planned.

HTR/VHTRs

It would therefore be necessary to examine more specifically:•

Whether the fuel and reactor design possesses sufficient margins
against the stresses to which the fuel is subjected both in normal
operation and in incidental or accidental situations, and those which
would lead to the release of a large quantity of the contaminants
contained in the core fuel,•The amount of radioactive elements which would be transferred tothe primary circuit, to the reactor building and to the environment in theevent of loss of particle robustness, and to deduce the appropriate
measures for containment.
3/2/2

Neutronics
The layout and material composition of HTR/VHTR cores differ
significantly from those of pressurized water reactors and fast reactor
cores already in operation, with many heterogeneities.
Figure 11:
Heterogeneities in a very hightemperature reactor core.

This makes it necessary to develop and qualify specific calculation tools.
The development of such tools and the mock-up experiment
programmes needed to qualify them are considered by designers.
Besides, today there is no evaluation of tritium production by theirradiation of graphite, or even helium, in an HTR/VHTR type reactor.
Tritium transfers in a coupled hydrogen production plant should also be
investigated.

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3/ Main issues for
3/2/3

HTR/VHTRs

Materials
For certain safety-related components including those ensuring the core
support, the concept of the VHTR implies the development of materials
which can withstand very high temperatures (1000°C), much higher thanthose encountered in present power reactors or even in the fast reactors
already in operation (550°C). Composite materials are proposed for theinternal structures and the elements brought up to high temperatures.
They undergo wide research and development programmes.
The design developed by AREVA would enable the reactor vessel and theintermediate heat exchanger, which constitute the second containment
barrier, to be maintained at a moderate temperature in normal operation(400°C), but these components would have to withstand much higher
temperatures in the event of an accident; ferritic steels already used forthe reactor vessels of pressurized water reactors and chromium steels are
the subject of research and development programmes, especially withthe aim of evaluating their weldability.
Figure 12:
AREVA very high temperature
reactor project.

Concerning graphite, it would be necessary to complete the experienceacquired with UNGG reactors in France and AVR concept in GreatBritain: the combined impact of very high temperatures and irradiation
largely remains to be studied, in particular to evaluate the dimensionalvariations of graphite under irradiation and the risks possibly induced on
the core behaviour during normal operation and in the event ofincidental or accidental transient. Research and development
programmes have been undertaken by designers.

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3/ Main issues for
3/2/4

HTR/VHTRs

Passive Systems
Exploiting the high thermal inertia and low power densities (less than
10 MW/m3, compared to about 100 MW/m3 for PWRs and 300 MW/m3
for Superphénix) of the HTR/VHTR concept, designers consider, for postaccidental situations, a passive residual power removal by conduction,
radiation and natural water convection in various circuits. Experimental
evidence of the actual capacity of the system to induce simultaneous
natural convection in the circuits, in conditions representative ofaccidental situations could raise some difficulties.
Besides, the experience feedback from Phénix and Superphénix design(see section 4/2/4 below) suggests paying particular attention to thedemonstration that no degradation of possible heat transfers byemissivity of structures would occur during the lifetime of the reactor.
More generally, evaluation of the efficiency of ‘‘lines of defence’’ basedon passive systems will need in-depth consideration; this subject has not
so far been discussed with designers.
3/2/5

Safety and reliability of the associated industrial
processes
The advantage of VHTRs is largely based on the capacity of this type ofreactors to supply high temperatures for industrial applications, chiefly
the production of hydrogen. However, coupling to an industrial hydrogenproduction facility would require a careful analysis of possible
interactions between the reactor and the hydrogen production facility(possible hazards). Though such an installation has been already settled
in Germany and is underway in Japan with the HTTR, no sufficiently
meaningful operation feedback exists. Even without an explosion,
disruptions from the hydrogen production facility could also lead tosignificant load transients to be accounted for in the design basis of thereactor circuits. It would also be necessary to evaluate the consequencesof such disruptions on the behaviour of the fuel in the core.
3/2/6

Other graphite-related risks
The "Wigner effect ", fire risk arising from the liberation of energy
accumulated in graphite during irradiation, should be avoided in
HTR/VHTRs, in view of target temperatures for these reactors (higher
than 400°C).

The Wigner effect played an
important role in the fire which
occurred on 10 October 1957 at
the British Windscale reactor,
Pile no 1.

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3/ Main issues forOn the contrary, for HTR/VHTRs, the risks associated with an accidental
ingress of air into the primary circuit in the event of a break should beclosely investigated. Such an event could indeed activate thedisintegration of the graphite pebbles or ‘‘compacts’’ and thereby
generate a widespread dispersion of fuel particles (TRISO), carbon 14 andchlorine 36. Besides, the graphite oxidation process at target
temperatures (up to 1000°C) should be carefully analyzed; research anddevelopment are in progress on these aspects. Anyway, a supported
demonstration that air ingress into the core of an HTR/VHTR could notlead to a big graphite fire has not yet been supplied.

HTR/VHTRs

The effects on the graphite of ingress of water into the primary circuit ofan HTR/VHTR should also be studied.
3/2/7

Radiological protection
Helium is a gas with a very high spreading capacity, due to its low
atomic weight. This feature must be taken into account in the designand operation of reactors, which should lead to impose a low thresholdvalue on the fission product inventory in the helium of the primarycircuit.
3/2/8

Fire and explosion risks related to hydrogen production
Concerning the fire and explosion risks originated in the hydrogenproduction plants, which could be coupled to VHTRs, the IRSN only
knows functional diagrams showing a geographical separation between
such plants and the reactors, the adoption of hillock separations orburied reactors. This aspect should be particularly examined dependingon the amount of stockpiled hydrogen and the processes adopted for
production. Should these plants be submitted to the safety rules of thechemical industry, the consistency of those rules to nuclear standards
should be checked.
3/2/9

Security concern
A more in-depth examination of the risks mentioned above should
contribute to identifying the sensitivity of HTR/VHTR reactors to acts ofmalicious damage. Designers and operators will have to define theappropriate design and operation provisions accordingly. Nevertheless, it
can be pointed out that, even if the high thermal inertia of HTR/VHTRs,
the existing wide operating margins preventing significant damage to thecore and the fact that the coolant fluid (helium) is chemically idle arefavourable factors, it would be necessary to take account of:© IRSN/2008 --- All rights reserveddoc reference
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3/ Main issues for•

The possible presence of a hydrogen production plant close to thereactor,

HTR/VHTRs•

The risks associated with the consequences of air and/or water
ingress into the primary circuit, which are still to be examined.

Concerning hydrogen, a satisfactory resistance against acts of malicious
damage can only be obtained by implementing appropriate measuresboth for the hydrogen production plant and the reactor itself. While anadequate separation of these two parts obviously would be beneficial,(distance, buried reactor, hillock separation, etc.), the likelihood of adrifting hydrogen cloud must be accounted for.
Limiting the possibilities of massive, provoked air or water ingress into an
HTR/VHTR should be a design demand, systems (main circuits andauxiliary circuits) should be dimensioned and room architecture
provisions defined accordingly: circuit redundancy, diversification ofequipment items, and geographical separation of redundant circuits.
3/3

Fuel fabrication and processing related aspects
First, it should be noted that the consequences of deploying a newreactor generation on fuel cycle management strongly depend on thetransition conditions established from the current NPP fleet to the newone. To date, different scenarios of transition towards a GEN-IV fleet
have been investigated by Electricité de France and the French Atomic
Energy Commission, but they have not yet been submitted to the IRSN.
The TRISO fuel for HTR/VHTRs takes the form of spherical particles ofabout 1 mm in diameter; the fuels studied are mainly uranium oxide or
mixed uranium and plutonium oxide.
Highly porous
pyrocarbon buffer
layer
95 μm

Figure 13:
Sectional view of a TRISO
particle.
Photo credit: CEA.

Inner dense
pyrocarbon IPyC
layer
40 μm
Fuel kernel
500 μmSIC 35 μm
Outer dense
pyrocarbon OPyC
layer
40μm

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3/ Main issues for
HTR/VHTRs

Enrichment in uranium 235 would be less than 20%. Such enrichment
can be achieved with available technologies and it is not challenging.
Nevertheless, the fabrication of such a fuel on an industrial scale would
require the construction of a dedicated facility. In terms of safety, thedesign bases of such a facility should be little different from those
selected for current fuel fabrication plants.
Figure 14:
TRISO HTR fuel balls fabricatedin the UO2 laboratory’s GAIA
facility at the Cadarache CEA
centre.
Photo credit: CEA.

Potential fuel processing procedures after irradiation would include twophases: the first aimed at accessing the TRISO particles contained in thepebbles or "compacts"; the second aimed at separating the actinides
from the fission products. The separation processes developed for GENIV reactors would aim at performing a combined extraction of actinides
to be recycled, more appropriate to limit the risks of diversion of nuclear
materials. The recycling would concern at least the uranium and theplutonium; present studies have developed along two paths:
hydrometallurgical procedures (wet line) and pyrochemical process (dryline).
Concerning production of TRISO particles, mechanical processing(crushing, grinding, cropping) was studied in the USA by the Oak Ridge
National Laboratory, as well as in Germany by FzJ (Forschungzentrum
Jülich GmbH) in the 1960s, 1970s and 1980s; the maximum actinide
recovery rate was quite small. Moreover, grinding can generate pollutionof the graphite by the actinides and fission products. Studies ofseparation processes using combustion in a controlled atmosphere havedemonstrated the impossibility of a total combustion of the graphite.
They raise questions on the production of very fine dust, which could bevery contaminating. A new method for producing fuel particles, at© IRSN/2008 --- All rights reserveddoc reference
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3/ Main issues forpresent proposed by the French Atomic Energy Commission, shouldmake use of a high voltage pulsed-current of 10 to 20 kA (between 200
and 500 kV).

HTR/VHTRs

To remove the fuel nuclei from their refractory layers, a treatment using
nitric acid cannot be proposed as the carbon and silicon carbide layers
are not dissolved by this acid. Mechanical processing by grinding would
involve the use of specific tools, on account of the abrasive nature ofsilicon carbide, and would not lead to a complete release of all thevolatile and other fission products. Theoretically, another possible
method would be the dissolution of the peripheral layers of the particle
using soda or carbonates. This method has not been examined by the
CEA as it presents risks of precipitating the plutonium, which is very
challenging in terms of criticality risk prevention. The methods currently
studied are as follows:•
The pyrochemical method, with elimination of the carbon andsilicon carbide layers by attacking them with gaseous chlorine at 950°C,
followed by pyrolysis in an oxygen atmosphere; the carbon/siliconcarbide coat then volatilises in the form of carbon oxides and siliconchloride,•
The mechanical method with destructuration of the coating layers
using pulsed-currents.

Finally, it must be emphasized that ways of reducing the volume ofgraphite waste from fuel processing are being studied.
In view of the above-mentioned arguments, the demonstration of thefeasibility of a ‘‘closed’’ cycle for the HTR/VHTR concepts has not yet
been completed to date, and important technological breakthroughs willin this respect be necessary.
It should also be noted that, in cycle facilities, the possibility ofadditional moderation from the water should be taken into account, atleast for incidental operation.
Moreover, it will be necessary to evaluate the qualification level of thecomputational tool and chains adopted for criticality risk prevention
studies. The experience available to date cannot be claimed as
sufficiently representative of future calculation configurations.
Nevertheless, it is important to note that this issue highly depends onthe methods implemented, especially the possible need for processing
large masses of ‘‘fissile’’ material. The analysis of processes should
originate new needs in terms of calculation tool and experiments.

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4/ Main issues for
SFRs

4/
Main safety, radiologicalprotection and security issues
identified to date for Sodiumcooled Fast Reactors (SFRs)
For sodium-cooled fast reactors, the IRSN has got a quite wide
knowledge of safety questions since France has started developing theconcept with Rapsodie, Phénix and Superphénix reactors. After launching
the RNR 1500 project, France has been an active participant in the EFR
project, mainly contributing with studies on the improvement ofbreeding capabilities, the reduction of core neutronics sensitivity to core
coolant void (see below, paragraph 4/2/2), and the incineration of minor
actinides. Today, in France, only the Phénix reactor is still in operation(until early 2009).
4/1

Safety objectives and approach
The document submitted by the CEA in March 2004 and prepared jointlywith AREVA NP, does not mention any specific safety studies for SFRs.
Nevertheless, in the straight line with the decision already adopted for
the EFR, designers will be looking for ways to improve the safety of thisconcept to reduce the risk of a generalised core melting, and even, to‘‘practically eliminate it’’. It may however be necessary to plan specificmeasures to contain a possible limited core melting (melting of either
one or several assemblies), accounting for the additional risks which
would result from a reactor vessel failure with an interaction of molten
core materials and sodium; such measures (internal core-catcher for
molten materials) have already been adopted for Superphénix and havebeen planned for both the RNR 1500 and EFR projects.
While awaiting documents from designers on the safety options for a
Generation-Four SFR project, the IRSN planned to examine the safety
orientations selected for the EFR project, established between 1988 and© IRSN/2008 --- All rights reserveddoc reference
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4/ Main issues for

1998, checking in particular their consistency with the general safety
objectives fixed for the SFR project.

SFRs

4/2

Reactor-related aspects
4/2/1

Specific risk inherent to metallic fuelThe use of metallic fuel (uranium, plutonium and zirconium alloy) is avariant option developed in the USA; at present it undergoes anexperimental campaign for transmutation in the Phénix reactor(METAPHIX experiments). Such a fuel would require a specificreprocessing method (pyrometallurgy) which has not yet been developedin France. It should be emphasized that this fuel presents a risk ofreaction with the steel in the (eutectic) compounds for the cladding,which reduces its melting temperature.
4/2/2

Neutronics effects in case of loss of coolant --- the socalled ‘‘sodium void effect’’
The risk of a sudden and sharp power increase in the event of a sodium
void in some core zones (positive ‘‘sodium void effect’’), by boiling,draining or the uncontrolled passage of a bubble of gas, is one of themain factors which determined adoption in France of several design
solutions for sodium-cooled fast reactors. Measures have been taken by
designers to avoid large-scale boiling of the sodium, gas being sucked
into the core or the core voiding in presence of a main reactor vesselleak.
The problem of the risks associated with the ‘‘sodium void effect’’ hasbeen re-examined after the unexpected automatic reactor shutdowns in
the Phénix reactor, which occurred in 1989 and 1990, the origin of which
remains widely unexplained. Nevertheless, it has been demonstrated
experimentally that these automatic shutdowns could not be originatedfrom insertions of positive reactivity due to any ‘‘sodium void effect’’.
The RNR 1500 and EFR projects were developed with the view ofdesigning reactor cores enabling positive ‘‘void effects’’ to be minimized
as far as possible. This objective, which remained a major aim for the EFR
project, has been adopted by the CEA for the SFR concept.

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4/ Main issues for
4/2/3

SFRs

Risk of core meltdown
Considering the phenomena described above in section 4/2/2, core
meltdown accident likelihood has been accounted for in the Phénix andSuperphénix designs, leading to specific resistance requirements for the
primary circuit (the main reactor vessel and its upper plug). Moreover,with the aim of reinforcing the Superphénix containment resistance inthe event of core meltdown, in addition to settling a safety reactor
vessel surrounding the main reactor vessel and a metallic dome above
the plug, the reactor has been equipped with a core-catcher for moltenmaterials inside the main reactor vessel, an option which has also been
adopted for the RNR 1500 and EFR projects. The Chinese CEFR reactor isalso equipped with such a device. The BN 350, BN 600 and Monju
reactors do not have any core-catcher.
In particular for Superphénix, the risks of core meltdown have been
widely studied and numerous experimental programmes were mainly
conducted in the Cabri and Scarabée facilities, located in the CEA’s
Cadarache Centre. Those programmes included:
Reliability studies for the emergency shutdown system and all thesystems ensuring residual power removal,Studies and tests on phenomena likely to engender partial melting in afuel pin, on the consequences of an assembly melting or a generalised
melting in the core, mainly the risks for criticality of molten materials
containing fuel.
A few tests went on in the Cabri reactor and in the Silène facility after
the decision -taken in 1997- to shut down Superphénix definitively andsupplied useful information. The IRSN has drawn up a summary
document on all these tests.
According to the elements indicated above, designers should
demonstrate that, for future SFRs, a partial melting of the core is not
likely to degenerate into a generalised melting. This could allow reducing
the presently very stringent containment design requirements(mechanical resistance of the primary circuit, capacity of the corecatcher, etc.). In this respect, advanced studies were carried out
retrospectively for the Phénix reactor, on account of the absence of sucha core-catcher for molten materials. Analyses, mainly those concerning
the risks for criticality, are quite complicated and require adoption ofadvanced neutronics, thermal and mechanical computation models andschemes, the development, validation and certification of which shouldbe pursued if necessary. If the above-mentioned orientations wereconfirmed, a qualitative breakthrough would need to be accomplished
for future SFRs on these subjects.

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4/ Main issues forMoreover if, despite those arguments, a generalised core melting was to
be included in the SFR containment design bases, several specific items
would have to be investigated both theoretically and experimentally,
through appropriate R&D programmes, including the possibility for:

SFRs

Avoiding the formation of critical zones within molten materials
containing fuel,
Avoiding significant energy releases in the reactor (by expansion of fuelor sodium vapour bubbles),Post accidental cooling of molten materials and the most suitable design
solutions for an internal core-catcher (in the main reactor vessel).
4/2/4

Sodium-related risks
After the commissioning of Phénix and Superphénix, it was deemednecessary to take into account the risks associated with sodium
pulverisation following a pipe breach for the design of rooms, becausethat could quickly generate quite high overpressures as well as theheating of the walls giving rise to water leaks from the concrete and to ahydrogen combustion or explosion. Appropriate technical solutions havebeen introduced.
The IRSN has been actively involved in acquiring knowledge on sodium
fires and has participated in the development of calculation tools in thisdomain. The experimental area explored was sufficiently wide to allow
obtaining an in-depth phenomenological knowledge of the combustionand supporting the qualification of calculation tools. Nevertheless some
aspects need further investigation, especially the risks of sudden ordeferred ignition of pulverised sodium when at "low" temperature (from100°C to 300°C) and the risks of fire rekindling in rooms should air blowback on aerosols of unburned sodium.
The risks of sodium-water reaction in a steam generator must also be
accounted for: they can lead to significant dynamic loads in thesecondary circuits and, if the accident is not very rapidly brought undercontrol by appropriate automatic systems, it can start an uncontrolled
sodium-water-air reaction up. In this respect, significant improvements
have been implemented in Phénix and Superphénix, mainly aimed atallowing an early detection of defects in steam generator tubes.
Nevertheless, experience feedback from the British PFR reactor and the
Phénix reactor in France shows that these detection systems are quitedifficult to operate efficiently (the seriousness of the event which
occurred in 1987 on the British PFR reactor, which led to the rupture of
40 steam generator tubes, is due to a temporary inhibition of a detection
system, as a consequence of malfunctions). Moreover, on account of the© IRSN/2008 --- All rights reserveddoc reference
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4/ Main issues forsteam generator design, the identification of a possible tube failure
would have been a challenge for Superphénix.

SFRs

Concerning the SFR concepts without intermediate circuits, i.e. in which
the primary sodium which may contain fission products after a cladding
rupture would directly exchange heat with the water of the electricity
production system, the IRSN does not possess enough elements toevaluate the impact of the technological solutions proposed (double wall
steam generator, etc.) on safety. Other solutions could be investigatedby designers (intermediate circuits containing a fluid less reactive withwater than sodium, etc.)
4/2/5

Passive systems
In the straight line of the options already adopted for reactors such as
Phénix and Superphénix, the design of an SFR should call on passive
systems, mainly for residual power removal. However, while thepossibility of a natural convection has been verified for some individual
circuits, no experimental evidence has ever been acquired for all thecircuits as a whole; only a computation-based demonstration has beenobtained. The possibility of performing an overall check by tests insatisfactory safety conditions should be examined.
The likelihood of implementing generalised natural convection in thecase of a "loop" SFR should also be investigated.
Besides, the design experience for Phénix and Superphénix has shownthat determining the residual power level and guaranteeing its removal is
not easy. This has supported either addition of supplementary residual
power removal systems (as it was the case for Superphénix), or
downscaling the operating power (Phénix). Moreover, the emissivity ofthe structures can significantly change in time. The possibility tomaintain a sufficient radiation thermal exchange over time should
therefore be checked throughout the lifetime of SFRs. This would be anecessary condition for the demonstration that active systems are notnecessary and can be avoided.
4/2/6

Structure Inspectability
Due to the presence of sodium in the SFR circuits, in-service inspectionof some components is quite hard; this is especially the case for thereactor internal core support structures. This difficulty was highlighted
for the Phénix reactor in the 1990s, in the framework of the CEA's
request to stretch out the operation of this reactor after the unexplained
emergency shutdowns by negative reactivity. An automated and© IRSN/2008 --- All rights reserveddoc reference
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4/ Main issues forinnovative remote control device, operated from outside the main
reactor vessel, was then developed very quickly by the CEA and provided
sufficient assurance to allow the reactor a few more operation cycles.

SFRs

Developments were also made for Superphénix (automated devices
allowing remote control of the main reactor vessel -MIR device- andsteam generator tubes).
Inspectability of structures is undoubtedly an important field for progressin both ‘‘integrated’’ and ‘‘loop’’ SFRs (according to the information
available on the Monju ‘‘loop’’ reactor, in particular).
Thus, designers and associated research and development organisations
should undertake actions in two fields, namely the possibility for
improving the conditions of in-service inspection and repair by judicious
design choices, and the development of appropriate devices for inservice inspection (non destructive sodium tests, etc.).
Finally, it must be pointed out that the IRSN has already drawn CEA’s
attention to the interest of taking advantage from Phénix dismantlement
to set up a specific inspection programme aimed at characterizing thestate of the major structures and components, which won’t have beenchecked in depth during the 35 years of operation of the reactor.
4/2/7

Security concern
The general arguments put forward in section 3/2/9 concerning
HTR/VHTR security concern can be extended to SFRs.
The very high sensitivity of the SFR concept to acts of malicious damage
is due to the large amounts of sodium stored in the reactor: sodium is
actually a metal which reacts chemically with air and water. The
prevention of acts liable to initiate a generalised sodium-water-air
reaction is thus an important challenge for the liability of the concept.
This should induce designers to take a special care to the systems
involved in fundamental safety functions, mainly those in charge of thechain reaction control and the residual power removal, in terms of
redundancy, separation and diversification.
Moreover, an exhaustive study of the risks engendered by the presenceof sodium as a coolant would be necessary (likelihood of chemical
reactions with other materials than water and air with energy delivery).
Besides, the design of the primary circuits of SFRs provides several
regions free of sodium, filled with gas (generally argon). If future SFR
design options do not allow the power increase effects in case of gas
transit in the core (positive ‘‘void effects’’) to be eliminated, the© IRSN/2008 --- All rights reserveddoc reference
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4/ Main issues forpossibility that the gas could be injected into the core will have to
receive special care.

SFRs

Finally, according to the future SFR design's sensitivity to a total andprolonged loss of electrical sources which could lead to a temperature
increase in the core support structures, appropriate resistance of theelectrical sources to acts of malicious damage will need investigation.
4/3

Fuel fabrication and processing related aspects
The mixed fuel (uranium and plutonium) of SFRs can assume an oxide,
carbide, nitride or metallic form. The cladding materials are mainly
stainless steels.
The fabrication of fuels in mixed uranium and plutonium oxide form
benefits from the experience of manufacturing the fuel for the Rapsodie,
Phénix and Superphénix reactors at the CEA’s Cadarache Centre, as wellas the manufacturing of fuel for PWR reactors, mainly in the MELOX
plant of Marcoule. This fabrication is carried out in glove-box dedicatedunits ensuring sufficient, but limited, protection for workers against
ionizing radiation (neutrons and gamma). The predictable change of theisotropic characteristics of the plutonium, unfavourable in terms ofradiological protection of operators, as well as the recycling of very
active minor actinides (americium, curium), could force equipment
remote maintenance, which would have a significant impact on thedesign of manufacturing plants and on the doses received by operators.
Concerning the fabrication of fuel in the form of carbide or nitride, noindustrial experience has been achieved. Apart from the risks inherent to
the processes adopted for handling and manufacturing the fuel, it isworth mentioning the high reactivity of carbides with water, oxygen and,to a lesser extent, nitrogen; that would require these compounds to be
manipulated in an inert environment (when they are in solid form).
Processing of the fuels in mixed oxide form can be based on theexperience feedback of processing similar fuel for the Rapsodie andPhénix reactors, performed at the La Hague plant or in the pilot
workshop at Marcoule (APM). Although the industrial feasibility of thewhole process has been demonstrated, it remains that a new factorywould be necessary to process the irradiated fuel of a whole fleet of fast
reactors. In terms of safety, the design bases of such a factory should
show only few differences from those of the present factories at La
Hague.

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4/ Main issues for
SFRs
Figure 15:
Schematic diagram of the Ganex
process studied by the CEA forthe grouped extraction ofactinides in the framework ofstudies on generation-four
reactors closed cycle (FP: fission
products; An: actinides; Ln:lanthanides).
Photo credit: CEA

Spent fuelRecycling

Dissolution
Preliminary
separation
Extraction

An backextractionFP

Ln backextraction
Ln

Waste

Concerning the post irradiation processing of the fuel in carbide or
nitride form, the numerous laboratory experiments performed
throughout the world and chiefly the Indian pilot experiments(processing several hundred kilograms of materials) apparently have notrevealed any major solubility problems during fuel dissolution phases. It
seems therefore possible to rule out the risks of criticality engendered by
the presence of undissolved components in the processing equipment.
Nevertheless, concerning carbides, the formation of carboxylic acid, due
to the presence of residual carbon in the solution, could activate thepiling-up of complex compounds, quite difficult to extract. These
compounds could combine with a fraction of the plutonium and thus
limit the efficiency of the whole process. Concerning the nitrides, theproduction of ammonium ions as by-products of the process should beallowed for as it could generate an explosion risk. Moreover, suitabletrapping devices could have to be developed to trap Carbon 14.
Concerning the prevention for criticality risks, it should be noted thatsome minor actinides have much more constraining neutronics features
than plutonium or uranium, e.g. the minimum critical masses of curium
245 and americium 242 are a few tens of grams, as compared to 510
grams for plutonium 239. The minimum critical masses could be evenlower for some minor transuranics piling-up at very high burn-ups;
moreover, at present, significant uncertainties exist on the minimum
critical masses of these actinides. Therefore, it would be necessary tovalidate calculation tools for minor actinides environments, knowing that
todate few critical experiments are available. Finally, care should betaken to ensure that the composition of the fuel after irradiation can becorrectly predicted in order to determine the most reactive© IRSN/2008 --- All rights reserveddoc reference
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4/ Main issues forenvironments to be selected for studies relative to the various stages offuel cycle.

SFRs

Some useful data on carbide and nitride type environments (crystalline
densities, etc.) are available, as well as results from criticality calculations
performed for laboratories, assuming that the characteristics of the
carbide and nitride type and oxide type environments are quite similar.
Nevertheless, additional studies would be necessary to cover the whole
fuel cycle in the framework of an industrial deployment of reactors using
fuel based on carbide and nitride. The CEA experimental facilities at
Valduc could be used when the time comes to obtain complementary
data to qualify the calculation diagrams.

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5/ Main issues for
GFRs

5/
Main safety, radiologicalprotection and security issues
identified to date for Gascooled Fast Reactors (GFRs)
In view of the few indications available on the GFR design, thisdocument only provides a preliminary evaluation of the safety-related
issues which should be addressed, in the present state of knowledge. It isclear that significant differences could result from the choices made bydesigners as regards the purposes of the reactor (production of electricity
or heat, combined cycle, coupling to a chemical production unit, etc.), its
overall design and the fuel selected.
5/1

Safety objectives and approach
An overall safety approach for GFRs is presented in the CEA’s March
2004 letter mentioned above (section 3/1), together with the main‘‘generic’’ safety options adopted. In view of the very high power densityof a GFR core compared to a VHTR core, recourse to active safetysystems to ensure core cooling in accidental situations seems a priori
unavoidable. Besides, in the CEA’s document, it appears that a total core
meltdown might not be retained by designers as a design basis for thecontainment.
5/2

Reactor related aspects
The neutronics calculations for GFR reactor cores imply the development
or improvement of nuclear data relative to current fast reactors, notably
as regards the refractory materials, graphite and minor actinides. These
data should be acquired through development and experimental© IRSN/2008 --- All rights reserveddoc reference
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5/ Main issues for

programmes planned in particular in the CEA Masurca critical mock-up
located at Cadarache.

GFRs

Concerning GFR fuel, as indicated in section 2/, several options are being
developed by the CEA, with the objective of reaching a resistanceequivalent to that of TRISO particle fuel and limiting at about 1,600°C
the maximum temperature reached in case of depressurization, an
accident situation particularly severe for this system.
Figure 16:
Example of fuel developed for
GFRs.

It should be noted that the power density of a GFR core (50 to 100
MW/m3) would be more than ten times higher than a VHTR reactor core,with a much smaller reactor thermal inertia.
Therefore, several specific measures are studied by the CEA, especiallysetting up an intermediate containment shell under helium pressurearound the reactor vessel (between 5 and 20 bar depending on theoptions), completed by active systems for core cooling. Setting up acontainment shell under pressure around a reactor would constitute aninnovation requiring a very detailed safety analysis.
Figure 17:
Schematic diagram of a gascooled fast reactor (reactor
vessel, components vessel,intermediary containment).

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5/ Main issues forOn the other hand, by design (use of a gas for heat removal from thereactor core), GFRs should be slightly affected by the coolant ‘‘void
effect’’, thus eliminating several total core meltdown initiators retainedfor SFRs. Thus, the main initiators would be related to breaches, as wellas to failures of the residual power removal systems, which could
generate a very sharp temperature increase, damageable to the fuel. The
reliability of the residual power removal systems (active and passiveones) and the temperature resistance of the fuel are, at a first sight,
major safety issues for GFRs, mainly if the designer aims at ‘‘practicallyeliminating’’ a generalised core meltdown. The qualification of active
systems functioning at high temperatures seems very challenging.

GFRs

The risks associated with water ingress into the primary circuit, inparticular the risk for a power increase due to a neutron moderation
effect, with a sharp increase of the power, are still to be assessed.
As regards the materials, according to the design choices adopted, aspecific aspect could concern the risk of embrittlement for components
subjected to a considerable flux of fast neutrons and the risk of a sudden
rupture of equipment also subjected to a significant pressure (70 bar).
Concerning the resistance of GFR reactors to acts of malicious damage,on account of the specific safety risks of this concept due to the highpower densities in the core associated with a primary fluid of low
calorific capacity, design measures (redundancy, diversification,
geographical separation) should be implemented to limit as far as
possible the risks for the core coolant being affected at the level of themain components (main reactor vessel, intermediate containment shell,
piping) and the primary circuit pumps.
5/3

Fuel fabrication and processing related aspects
The mixed (uranium and plutonium) fuel of GFR reactors could be in theform of carbide or nitride oxide. The cladding materials considered areceramic. Concerning the studies conducted by the CEA, the reference
fuel is today in the form of carbide and the reference ceramics is siliconcarbide.
Concerning fuel fabrication and processing, several of the considerations
developed in sections 3/3 and 4/3 also apply here.

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6/ Waste production
and management

6/
Waste production andmanagement in theframework of the deploymentof a Generation-Four reactor
fleet
The question whether deploying a fleet of GEN-IV reactors may have asignificant impact on the production and management of radioactive
waste includes several aspects. The two criteria put forward today are
reducing the harmfulness of waste and minimizing their volume.
Concerning harmfulness, the main expected gains are generally
expressed by designers in terms of potential ‘‘radiotoxicity ” decrease
and are related to the reduction by transmutation of the spent fuelcontent of plutonium and minor actinides or of residual waste, after
reprocessing. The GEN-IV reactors which are the subject of thisdocument have advantageous features in this regard since they are allable to burn plutonium and/or actinides out, but at a variable extent. It
should nevertheless be emphasised that a significant reduction in thedosimetric impact for the storage of waste and/or spent fuels in deep
geological formation cannot be expected. All the evaluations performed
to date (European EVEREST and SPA exercises, ‘‘dossier 2005 Clay’’by ANDRA) show that the impact is mainly due to long-lived fission
and/or activation products (mainly, iodine 129 and chlorine 36), as theactinides remain confined over very long periods in a small volume ofrock surrounding the stored waste packages. Even adopting very
unfavourable assumptions (transfers through fractures, low retention
properties of the rock), the dosimetric impact of actinides remains of thesame order of magnitude as for fission products, and it is attributable tothe uranium 238 in secular equilibrium concentration with its daughters,
and not to plutonium and minor actinides.

Radiotoxicity: Sum of the
products of the activity ofeach radionuclide containedin the waste by the dosecoefficient (Sv/Bq) of theradionuclide which indicatesthe effects on man.
EVEREST (EValuation of
Elements Responsible forthe effective Engaged dose
rates associated with thefinal STorage of radioactive
waste): European project formodelling storage site forreprocessing waste (vitrified
wastes, enclosures and endpieces, etc.) in three typesof geological formation:clay, granite and salt.
SPA (Spent fuel disposal
Performance Assessment):
European project devoted tomodelling a storage site for
spent
fuels
in
deep
geological formation.

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Reducing the harmfulness of waste and spent fuels (if the latter were to
be stored), would require the new reactors to produce less iodine 129(and, to a lesser extent, less chlorine 36). Concerning iodine 129, theGEN-IV reactors do not appear as significantly advantageous comparedto the current reactor fleet for the same installed electrical power.
Moreover, chlorine 36, the generation by activation of which depends onthe graphite component fabrication mode (impurity content, addition ofchlorine through the purification process, diverse pollutions during
fabrication) and on the irradiation history (radial or axial position in thecore, neutron spectrum and fluxes, radioactive decay, incidents), shows atrend toward an increased production compared to current reactors if
the graphite component fabrication specifications do not impose drastic
thresholds on the content of relevant impurities. In this respect,industries are searching for graphite purification processes which avoid
the use of chlorine (fluorine based processes). As regards the graphite in
the components (”compacts” or pebbles, blocks used as moderators or
reflectors, cladding containing silicon carbide and carbon, matrix
containing nitrogen component, carbide or nitride based fuels), it will beable to generate, by neutron activation, carbon 14 (the radioactive
period of which is 5,730 years) in much larger quantities than for theoperating reactors. This radionuclide is quite mobile in naturalenvironments and the capacity of the storage site selected for thecorresponding waste to ensure a satisfactory containment over several
tens of thousands of years should be examined.

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6/ Waste production
and management

Therefore, the capacity of GEN-IV reactors to significantly reduce thelong-term dosimetric impact of waste and fuels to be stored in deep
geological formations, does not supply convincing arguments for thedevelopment of such reactors. Moreover, it is known that the operationsof separation and fabrication of the fuel from separated products would
be complex. From a strictly technical point of view, a real assessment
should be carried out, taking into account the doses received byoperators and the waste produced by the above-mentioned operations.
Concerning the reduction of the volume of spent fuels and waste
produced, the very high burn-ups of the new reactors should, in principle,
reduce the quantities of spent fuel to be unloaded and, consequently, theamount of waste from reprocessing. However, this conclusion assumes
implicitly that the adaptations of the processing procedures necessary toaccommodate the characteristics of the unloaded elements (newcladdings, carbide matrix) would not increase the amount of processwaste and waste generated per unit mass of processed fuel. It would also
be necessary to check that the intrinsic containment properties of
packages and fuels remain sufficiently high to obtain an overall gain interms of short and long-term waste management, as a result of thesmaller volumes produced. Today the state of research does not allow© IRSN/2008 --- All rights reserveddoc reference
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formulating a judgment on all these issues. Besides, the reduction in thevolume of spent fuels and waste to be stored does not necessarily
generate a proportional reduction of the depot area of a geologicalstorage site, which is strongly dependent on the heat production
associated with waste. In a qualitative approach, the reduction in theamount of minor actinides to be stored would not allow lower thermal
emissions after a few centuries than the fleet of reactors in operation.
Nevertheless, it will be necessary to make sure that the short andmedium-lived fission products contained in the new waste can decrease
within a sufficiently short time to reach, while they are stored, criteria ofthermal emissions compatible with their introduction into the geologicalformation.

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6/ Waste production
and management

In conclusion, at this stage, it is hard to evaluate whether GEN-IV
reactors could provide any significant profit to waste management.
Nevertheless, it is worth emphasizing that the actual advantage can beprovided by the management of nuclear materials in a ‘‘closed’’ cycle,
rather than by the features of the GEN-IV reactors considered. Such amanagement could indeed avoid spent fuel storage, which would
facilitate the long-term safety demonstration of a geological storage
installation, even if such a demonstration does not seem impossible
should these fuels need to be stored. These considerations also apply tothe existing stocks of depleted uranium. These stocks, of low activity
today, will become active and highly productive of radon once thesecular equilibrium between uranium 238 and its daughters isestablished, just like very rich deposits which cannot be exploited
without extensive radiation protection measures; hence the interest in
using these stocks.
These considerations show that the analysis of the potential of GEN-IV
reactors to improve the management of radioactive waste cannot be
dissociated from all the elements of decision relative to the exploitationof the energy potential of nuclear materials.

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7/ Conclusion Actions by the IRSN

7/
Conclusion - Actions
undertaken and planned by the
IRSN
In conclusion, only the helium-cooled HTR/VHTR (High or Very High
Temperature Reactor) and the SFR (Sodium-cooled Fast Reactor)
concepts appear to be at a sufficient level of maturity to consider theconstruction of a prototype in France by 2020. Concerning GFRs (Gascooled Fast Reactors), in the framework of a progressive approach, only
the construction of a great flexibility experimental reactor is conceivable,
which would, for example, allow individual testing of the performancesof some sub-assemblies and components (fuel and core, components,
gas turbine, etc.)
A precise definition of the safety approach for the different types of
GEN-IV reactors would require the design studies to be further
advanced, especially in the identification of the incidents and accidents
to be dealt with. In this respect, the IRSN has already undertaken
discussions with designers and partner organisations on a number ofgeneral aspects. In this framework, the role which could be assigned to
the Probabilistic Safety Analysis (PSA) and its reliability, especially incase of recourse to passive systems, will be discussed.
Concerning the helium-cooled HTR/VHTR concept, technical discussions
have been undertaken between the IRSN and AREVA NP on the
ANTARES project, on the basis of a protocol signed on February 2005,
detailing the framework of the discussions and their non-committing
character. Several specific actions cover the technical aspects shared by
HTR/VHTRs and GFRs.
Concerning SFRs, discussions were undertaken in 2006 between the IRSN
and the French Consultative Group for Safety (GCSF), including the
Nuclear Energy Directorate (DEN) of the CEA, AREVA and Electricité deFrance (a GIF “mirror” group). The IRSN is reactivating its own internal
main competencies, in order to have fundamental knowledge and tools
at its disposal, especially in the fields of accidents leading to fuel damage© IRSN/2008 --- All rights reserveddoc reference
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and the risks related to sodium. The neutronics of future SFR cores may
cause problems, as SFR cores are significantly different from cores ofreactors in operation (“heterogeneous cores” with a large amount ofminor actinides); on these open subjects, the IRSN plans to rely on a
specialist competency group.

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7/ Conclusion Actions by the IRSN

Concerning GFRs, the CEA is pursuing developments in the perspective of
the REDT project, a low power experimental reactor. The IRSN and the
GCFS also undertook consultations on this subject in 2006.
Apart from the discussions, participations and involvements mentioned
above, the IRSN has selected a few privileged strategic axes for shortterm actions of study, research and development corresponding to GENIV safety items it considers as very important:•

The behaviour, in normal operating, incidental and accidental
conditions, of refractory fuels for HTR/VHTR and GFR reactors, predictive
calculation models, uncertainties and associated statistical analysis,•

The neutronics of the HTR/VHTR, SFR and GFR reactor cores, withthe progressive setting-up of the competency group so as to acquire thenecessary knowledge and calculation tools, and to be able to implement
them as soon as sufficient data are available (for example the reductionof the ‘‘sodium void effect’’ in the SFR cores),•
The risks associated with water ingress into a GFR, and theimplementation of the associated calculation models,•

The transfers of radioactive products out of refractory fuels and into
the circuits in the event of an accident, for gas reactors,•

The risks of propagation of a local melting in SFR and possibly GFR
cores, the risks of a return to criticality of a corium (molten corematerial) in the reactor vessel, the propagation modes of the corium
inside and outside a reactor vessel.
The studies undertaken will allow the future identification of the needs
for experimental programmes specific to the IRSN, to be performed
either in mock-up facilities or in test reactors.
Concerning the aspects related to the fuel cycle of GEN-IV reactors, the
IRSN will keep a technological watch on the development of the fuelfabrication and processing procedures. The possibility of participating ininternational working groups concerned with the safety andmanagement of the cycles proposed for GEN-IV reactors will be
examined. Experimentation and development of models cannot however
be usefully defined as long as the cycle and material fabrication andreprocessing procedures have not been precisely described. On the other
hand, the IRSN can already benefit from the existing knowledge
concerning the scenarios of GEN-IV reactor deployment in order todetermine the impact of different strategies on the use of nuclear© IRSN/2008 --- All rights reserveddoc reference
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materials and on the waste elimination methods and to go deeper into
the considerations presented in this document.

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7/ Conclusion Actions by the IRSN

Lastly, it is clear that, in the development of GEN-IV systems, securityrelated issues will have to be investigated at the design stage.

© IRSN/2008 --- All rights reserved