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ADRIANA

ADvanced Reactor Initiative And Network Arrangement

Coordination and Support Action

Co-funded by the European Commission under the

Euratom Research and Training Programme on Nuclear Energy

within the Seventh Framework Programme

Grant Agreement no. 249687

Start date: 01/02/2010 Duration: 18 Months

Website: http://adriana.ujv.cz

Roadmap proposal for building knowledge and facilities needed for nuclear energy systems

development

Deliverable D8.4

Author(s): Ladislav VÁLA (CV Řež), Christian LATGÉ (CEA), Pietro AGOSTINI (ENEA), Christian POETTE (CEA), Wolfgang

HERING (KIT), Ludo VERMEEREN (SCK•CEN), Vlastimil JUŘÍČEK (CV Řež), Jakub PRAHL (CV Řež), Nathalie GIRAULT (IRSN),

Martin PEČANKA (LGI), Evžen NOVÁK (CV Řež)

Approval: Ivo Váša (ÚJV Řež a.s.)

D8.4 - ADRIANA - Roadmap proposal for building knowledge and facilities needed for nuclear energy systems development - Final Signed

EURATOM FP7 Grant Agreement no.249687

http://adriana.ujv.cz

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ADRIANA project – Grant Agreement no. 249687

ADvanced Reactor Initiative And Network Arrangement

EC Scientific Officer: Roger Garbil

Document title Roadmap proposal for building knowledge and facilities needed for nuclear energy systems development

Author(s) L. Vála, C. Latgé, P. Agostini, C. Poette, W. Hering, L. Vermeeren, V. Juříček, J. Prahl, N. Girault, M. Pečanka, E. Novák

Number of pages 57

Document type Deliverable

Work Package WP8

Document number D8.4

Date of completion

Dissemination level

_ Confidential (ADRIANA consortium and European Commission)

_ The above + specific dissemination group: …

ü Public

Document status Final version

Summary

The coordinating action ADRIANA (ADvanced Reactor Initiative And Network Arrangement) has been initiated to set up the network dedicated to the construction and operation of research infrastructures in support of developments for the European Industrial Initiative for sustainable nuclear fission. The project sets these objectives for the following reactor systems and related technologies: Sodium-cooled Fast Reactor (SFR) (WP2), Lead-cooled Fast Reactor (LFR) (WP3), Gas-cooled Fast Reactor (GFR, including very high temperature technologies) (WP4), Instrumentation, diagnostics and experimental devices (WP5), Irradiation facilities and hot laboratories (WP6), and Zero power reactors (WP7).

The present document represents a synthesis of the work which has been performed so far by the several groups (Work Packages) involving a number of experts from several European research organisations taking part in the ADRIANA project.

The first part of this report deals with the present situation regarding the existing and planned experimental infrastructure which is necessary for the development of advanced Gen IV nuclear power systems in the EU. The projects of ASTRID, MYRRHA, ALFRED, and ALLEGRO prototypes are presented together with the experimental facilities and experimental platforms (existing, under construction or under design) which were identified by the ADRIANA project participants as needed for the successful development and implementation of the SFR, LFR, and GFR technologies. The situation for crosscutting




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R&D activities in support of these technologies such as instrumentation and ISI&R, material testing reactors, irradiation and fuel testing facilities, laboratories with hot cells, and zero power reactors is also explained.

The second part of the report concern the cost estimation and means of funding of both the major projects (ASTRID, MYRRHA, ALFRED, ALLEGRO) and the needed experimental infrastructure described in the first part.

Finally, the third part presents the proposal of roadmaps for experimental infrastructure which will ensure successful development of SFR, LFR, and GFR technologies in the EU.

Revisions

Rev. Date Short description Main author WP Leader

(name & signature)

Coordinator

(name & signature)

01 15.07.2011 Draft L. Vála J. Prahl I. Váša

02 31.07.2011 Final version L. Vála J. Prahl I. Váša

Distribution list

Name Organisation Comments

Project participants All participants of the ADRIANA project

Mr. Roger Garbil (EC DG RTD)

Mr. Michel Hugon (EC DG RTD)




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Table of contents

List of abreviations ............................................................................................................. 6 1 Introduction............................................................................................................ 8 2 SFR technology ......................................................................................................12

2.1 ASTRID ............................................................................................................................................ 12 2.1.1 Main requirements for ASTRID.................................................................................................. 13

2.2 Support research facilities and projects ........................................................................................... 14 2.2.1 SFR technology development .................................................................................................... 14 2.2.2 Safety research .......................................................................................................................... 17

3 LFR technology ......................................................................................................19

3.1 MYRRHA.......................................................................................................................................... 19

3.2 ALFRED ........................................................................................................................................... 20

3.3 Support research facilities and projects ........................................................................................... 22

3.4 Needed infrastructures ..................................................................................................................... 25

4 GFR technology......................................................................................................27

4.1 ALLEGRO......................................................................................................................................... 27

4.2 Support research facilities and projects ........................................................................................... 28

4.3 Summary of needs for GFR development........................................................................................ 29

5 Crosscutting R&D activities ......................................................................................30

5.1 Instrumentation and ISI&R ............................................................................................................... 30

5.2 Material testing reactors, irradiation and fuel testing facilities.......................................................... 33

5.3 Laboratories with hot cells................................................................................................................ 35

5.4 Zero power reactors ......................................................................................................................... 37 5.4.1 Expressed needs for SFR, LFR and GFR ...................................................................................... 37 5.4.2 Conclusions for zero power reactors ......................................................................................... 39

5.5 Nuclear data measurements for fast reactors [14] ........................................................................... 39

6 Costs and funding of European research infrastructures ..............................................41

6.1 SFR infrastructure cost estimation and funding ............................................................................... 41

6.2 LFR infrastructure cost estimation and funding................................................................................ 42

6.3 GFR infrastructure cost estimation and funding............................................................................... 45

7 Roadmaps for European research infrastructure for fast reactor systems development ....48

7.1 SFR technology roadmap – Milestones for the design and implementation of ASTRID ................. 48




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7.2 LFR technology roadmap................................................................................................................. 49 7.2.1 Roadmaps of MYRRHA and ALFRED LFR.................................................................................... 49

7.3 GFR technology roadmap ................................................................................................................ 50

7.4 Roadmap for instrumentation and diagnostics development ........................................................... 52

8 Conclusions ...........................................................................................................54 9 References ............................................................................................................57

List of figures

Figure 1: ESNII Roadmap for Gen IV fast reactor system development in the EU. ............................................ 9 Figure 2: Scheme of MYRRHA. ......................................................................................................................... 19 Figure 3: Roadmap proposal for qualification of core instrumentation. ......................................................... 32 Figure 4: Main phases of ASTRID projects development. ................................................................................ 48 Figure 5: The SFR technology roadmap with respect to the ASTRID programme. .......................................... 49 Figure 6: Roadmap for experimental infrastructure needed by MYRRHA and LFR projects. .......................... 50 Figure 7: Schedule of the main technology domains to be covered in view of ALLEGRO constraints. ............ 51 Figure 8: Roadmaps for European SFR, LFR and GFR projects......................................................................... 56

List of tables

Table 1: The main parameters of the ALFRED prototype. ............................................................................... 22 Table 2: Lead and LBE facilities and projects – technical goals, status, funding, and availability. ................. 25 Table 3: Main core performance and experimental capability of ALLEGRO.................................................... 27 Table 4: Typical operation and environmental conditions of Gen IV systems................................................. 30 Table 5: Overview of the applicability of available instrumentation............................................................... 31 Table 6: Cost estimation for European experimental platform for SFR development..................................... 42 Table 7: Cost estimation for European experimental facilities for LFR R&D.................................................... 45 Table 8: Cost estimation for the GFR research infrastructure. ........................................................................ 47




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List of abreviations ADRIANA ____ ADvanced Reactor Initiative And Network Arrangement

ADS ________ Accelerator-driven System

ALFRED______ Advanced Lead Fast Reactor European Demonstrator

ASTRID ______ Advanced Sodium Test Reactor for Industrial Demonstration

BWR ________ Boiling Water Reactor

CFD_________ Computational Fluid Dynamics

DHR ________ Decay Heat Removal

DRACS ______ Direct Reactor Active Cooling System

DRC ________ Direct Reactor (active) Cooling system

EFPD________ Effective Full-Power Day

EFR _________ European Fast Reactor project

EPR_________ European Pressurized Reactor

ELSY ________ European Lead-cooled System

EMP ________ Electro-Magnetic Pump

ERIC ________ European Research Infrastructure Consortium

EU__________ European Union

FA__________ Fuel Assembly

FNR_________ Fast Neutron Reactor

GEN IV ______ Generation IV

GFR_________ Gas-cooled Fast Reactor

GIF _________ Generation IV International Forum

HLM ________ Heavy Liquid Metal(s)

HTR ________ High-temperature Reactor

I&C _________ Instrumentation and Control

IHX _________ Intermediate Heat Exchanger

ISI&R _______ In-Service Inspection and Repair

LBB _________ Leak Before Break

LBE _________ Lead – Bismuth Eutectic alloy

LFR _________ Lead-cooled Fast Reactor

LOCA _______ Loss-of-Coolant Accident

LWR ________ Ligh Water Reactor

MA _________ Minor Actinide

MOX________ Mixed Oxide fuel

MTR ________ Material Testing Reactor

NEA ________ Nuclear Energy Agency

ODS ________ Oxide Dispersed Steels

OECD _______ Organisation for Economic Co-operation and Development

PPP_________ Private-Public Partnership

PSAR________ Preliminary Safety Assessment Report




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PWR ________ Pressurised Water Reactor

R&D ________ Research and Development

RVACS ______ Reactor Vessel Auxiliary Cooling System

SC-CO2 ______ Super-Critical CO2

SCWR _______ Super-Critical Water Reactor

SFR _________ Sodium-cooled Fast Reactor

SG__________ Steam Generator

SGU ________ Steam Generator Unit

SGTR________ Steam Generator Tube Rupture

SNETP_______ Sustainable Nuclear Energy Technology Platform

SRA_________ Strategic Research Agenda

TAREF_______ Task on Advanced Reactors Experimental Facilities

TC __________ Thermocouple

US__________ Ultra-sonic

VHTR _______ Very High-temperature Reactor

VVER _______ Vodo-Vodyanoi Energetichesky Reactor = Water-Water Energetic Reactor

WP _________ Work Package




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1 Introduction This fourth generation will focus on fast reactors which are able to convert a large majority of uranium-238 (238U) into plutonium-239 (239Pu) while producing electricity. In this way, it will become possible to exploit more than 90% of natural uranium to generate electricity instead of only 0.5 to 1%. The large quantities of depleted and reprocessed uranium could be used to maintain the current electricity production for several thousand years. The worldwide availability of primary fissile resources can thus be multiplied by approximately 100 times. The construction of fast reactors will also open the door to unlimited plutonium recycling (multirecycling) by taking advantage of its energy potential. Radioactive waste management is yet another challenge facing Generation IV (Gen IV) reactors which involves reducing the volume and the inherent long-term radioactivity of final waste. These reactors may in fact be capable of burning some of the long-lived radioactive elements contained in radioactive waste: minor actinides (americium, neptunium, curium, etc.).

The Generation IV International Forum (GIF) was born out of the desire to create an international R&D framework capable of boosting research in the different countries so that the most efficient technologies can emerge as quickly as possible.

Four main objectives have been defined to characterise the future reactor systems that must be:

• Sustainable: using a minimum natural resources and be environmentally friendly (limiting waste production in terms of radiotoxicity, mass, decay heat, etc.).

• Cost-effective: in terms of investment costs per installed kWe, the price of fuel, the cost of operating the facility, and consequently, the cost of electricity production which must be competitive in comparison with other energy sources.

• Safe and reliable: aiming for improvement compared with reactors currently in operation, while eliminating the need to evacuate the population from the site as much as possible, regardless of the cause and severity of the accident.

• Proliferation resistant and protected against any external hazards.

Six critical systems have been selected by Generation IV, among them SFR, LFR and GFR.

The Strategic Research Agenda (SRA) of the Sustainable Nuclear Energy Technology Platform (SNETP) has selected the fast neutron reactor systems as a key structure in the deployment of sustainable nuclear fission energy. Fast reactors development needs an important technology support to finalise their innovative design and to asses their safety. The ADRIANA project provides a roadmap for European research infrastructure to address this support. The project is focused on the following three fast reactor systems identified in the Strategic Research Agenda (SRA): Sodium-cooled fast reactors (SFR), Lead-cooled fast reactors (LFR), and Gas-cooled fast reactors (GFR). According to the ESNII roadmap for Gen IV fast reactor system development within the EU, the SFR system is a reference technology while LFR and GFR are alternative technologies (see Figure 1).




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Figure 1: ESNII Roadmap for Gen IV fast reactor system development in the EU.

Sodium-cooled fast reactor (SFR) nuclear energy systems are considered, among the six candidate technologies selected in the Generation IV Technology Roadmap, as the most mature technology, thanks to the large operational feedback gathered in several countries since the fifties and for their potential to meet the new technology goals. Currently, six reactors are operated in the world: BN600 and BOR60 in Russia, Joyo and Monju in Japan, FBTR in India, and CEFR in China. Two reactors are being built: BN800 in Russia and PFBR in India. Several industrial projects are in design phase such as BN1200 in Russia, CFBR in India, ASTRID in France, KALIMER in South Korea, and JSFR in Japan. Sodium-cooled Fast Reactor is also considered by Europe as the reference concept within the frame of European strategy. For SFR, the following goals need to be underlined:

• Economic competitiveness: by design simplification, higher burn-up, and improvement of thermal efficiency to attract large-scale investment to commercialise fast neutron reactor technology.

• Inspection and repair: matters, issues and execution of continuous monitoring and periodical inspection (so-called In-Service, Inspection and Repair (ISI&R)) to improve the plant availability and safety. The impact of opacity for liquid metals (sodium as well as lead).has to be taken into account thanks to innovative measurement or inspection technologies.

• Safety: the risks of sodium fire and sodium-water reaction must be eliminated or minimised; the previsions for core disruptive accident included into the plant design.

Sodium-cooled Fast Reactors with a closed fuel cycle and potential for minor actinide burning may allow minimisation of volume and heat load of high level waste and provide improved use of natural resources compared to only 1% energy recovery in the current once-through fuel cycle with thermal reactors.

Among the fast reactor systems, the SFR has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype, and commercial size reactors since the 1940s (mote than 400 years of operation lifetime accumulated, end of 2010). Despite the




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desirable features of an SFR such as a low pressure with excellent heat transfer characteristics, the chemical reactions of sodium with air and water and opaque characteristics of the liquid metals, need to be addressed carefully because they may hamper the deployment of SFRs for electricity generation. One of the important technology areas to be strongly developed is in-service inspection and repair for metal cooled systems. As the first step, an improvement of the respective methods is addressed (i.e. improvement of the inspection methods) by describing specific properties of sodium, basic layout reactor concepts with regard to accessibility and inspectability at the design stage, and by a list of available techniques for inspection. For Gen IV, the state of art has to be improved to develop applicable, reliable, and safe methods and techniques.

In order to evaluate the potential innovations for SFR, a Europe project (7th Framework Programme 2009-01 to 2013-01), CP-ESFR has been initiated and provides new tracks for improvement for the previously underlined items.

The Lead-cooled Fast Reactor (LFR) is another of the three fast-neutron fission reactor systems studied within the framework of Gen IV. As for other fast neutron reactor systems, the sustainability and waste minimisation are primary reasons for developing an LFR. However, central objectives of the ELSY project and of its successive evolution were to develop a competitive and safe reactor using simple and innovative engineered technical features. The enhanced simplicity is expected to reduce capital cost and construction time.

The choice of lead as primary coolant has a number of positive aspects with regard to safety and in simplifying the design:

• The non-existence of exothermic chemical reactions between lead and water or air provides favourable conditions for elimination of an intermediate circuit which reduces the footprint of the plant even if it is required to investigate the physical consequences of the lead–water interaction. According to the outcomes of ELSY project, most of chemistry control issues related to LBE can be transferred to lead coolant.

• High boiling point of lead (1749°C at 1 bar) reduces the risk of core voiding due to coolant boiling. • High density of lead favours fuel dispersion phenomena when compared to fuel compaction

phenomena in case of core destruction. This reduces the likelihood that fuel collects within the primary system, especially at the reactor vessel bottom in such way that re-criticalities may occur. A core catcher is probably not useful for an LFR, but it is necessary to investigate the behaviour of the fuel at the lead bulk surface, during the severe accident scenario. Nevertheless, the anti-seismic devices need to take into account this high density.

• High heat of vaporisation of lead makes it possible to operate the reactor at low primary system pressure (sub-atmospheric), which allows reduced reactor vessel thickness.

• High thermal capacity allows a significant grace time in case of, for example, loss-of-heat-sink accidents.

• Lead appears to form compounds with iodine and caesium at temperatures up to 600°C. This reduces the source term to the confinement/containment during accidents in which volatile fission products are released from the fuel matrix.

• Lead shields gamma-rays effectively. • Some components like, for example, the steam generator units have an innovative and compact

design. They are placed in the hot leg allowing a simple flow path design. • The low moderation effect of lead permits a greater spacing between fuel pins, resulting in a low

core pressure drop of about 1.2 bar.




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• The combination of a simple flow path and the low core pressure drop enhances the establishment of natural convection of the coolant in the primary system for heat removal from the core thereby reducing the risk of overheating accidents during, for example, a loss-of-flow accident.

• Nevertheless, lead can induce higher corrosion which leads to the necessity to control accurately the oxygen content in the primary vessel, hot and cold plenum in order to stabilize an oxide coating and to prevent from lead oxide precipitation. Nevertheless, alternative dedicated coatings could be developed and qualified in relevant conditions in the near future.

The LFR is a very innovative concept, as no demonstrator has ever been built, and challenging because of the high lead density and the relatively poor behaviour with regards to the structural material, inducing the necessity to foresee specific coatings.

The attractive Gas-cooled Fast Reactor (GFR) concept aims to combine the benefits of fast spectrum and high temperature (up to 850°C), using helium as coolant. The high coolant temperature leads to target high energy conversion efficiency (43–48%) and opens the possibilities for new applications of nuclear energy, such as heat production processes (metallurgy, hydrogen or synthetic hydrocarbon fuel production…). The GFR concept aims to provide these new potentialities by affording a sustainable energy supply in the long term including minor actinides burning capability. In addition, the potential merits of the helium coolant regarding safety, in-service inspection, reparability, operability, and dismantling are acknowledged in spite of the longer lead-time needed by this new reactor line:

• no threshold effect due to phase changing • very limited voiding reactivity effects, providing a quasi-decoupling of the reactor physics from the

state of the coolant, • no chemical reaction, • optical transparency, • temperature measurement capabilities, • inert and non toxic, • not activated.

The GFR concept is clearly innovative, as no demonstrator has ever been built, and challenging because of the poor thermal properties of the coolant. The key feasibility issue is to develop a fuel and clad technology able to withstand high temperature (in the range of 1000°C for the clad and 1300°C for the fuel in normal operating conditions). Innovative refractory materials are considered, leading to a new phenomenology at the very high temperature, regarding the core degradation (> 2000°C).

The GFR development roadmap includes parallel viability studies of:

• a moderate power demonstrator named ALLEGRO (< 100 MWth) without electricity generation as a necessary step towards an electricity generating prototype and further a head of series commercial reactor,

• the target commercial electricity generating reactor (∼2400 MWth) with a designed lifetime of 60 years and its fuel.

Due to the technological goals, Gen IV systems operate on higher temperature like GFR at ∼850°C and at higher system pressure, which may extend up to 10 MPa. Very stringent for instrumentation is the harder neutron flux in the core and in the vicinity. In general, the Gen IV systems cover a quite wide range of temperature, pressure, and velocity.




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2 SFR technology The SFR technology has been selected by ESNII as the reference one. The SFR system uses liquid sodium as the reactor coolant allowing high power density with low coolant volume fraction. While the oxygen-free environment prevents significant corrosion, sodium reacts chemically with air and water and requires a sealed coolant system. The primary system operates at near-atmospheric pressure with typical outlet temperatures of 500–550°C; at these conditions, austenitic and ferritic steel structural materials can be utilized, and a large margin to coolant boiling (about 330°C at 0.1 MPa) is maintained. The reactor unit can be arranged in a pool layout or a compact loop layout. Typical design parameters of the SFR concept are being developed in the framework of the Generation IV System Arrangement. Plant sizes ranging from small modular systems to large monolithic reactors are considered. Here, we will focus on large monolithic reactors.

Despite the desirable features of a SFR such as a low pressure with excellent heat transfer characteristics, the chemical reactions of sodium with air and water, and its opacity may hamper the deployment of SFRs for electricity generation. The characteristics of sodium also make the in-service inspection and repair (ISI&R) more difficult compared to gas option. Each existing SFR design utilizes design measures to increase its reliability in these aspects.

Issues of concern from the viewpoint of SFR safety are the positive sodium void reactivity, except for very small SFR cores, and the possibility of recriticality in a hypothetical core disruptive accident (HCDA). The effects of these issues need to be evaluated carefully in the safety assessment of SFR to confirm that the consequences of the accidents are mitigated and retained by the containment functions of the SFR plant and the release of source terms is restricted adequately.

The Generation IV SFR R&D focuses on a variety of design innovations for enhanced safety performance (in particular, aiming at a decreased risk of core degradation accidents), actinide management, development of recycled fuels, and improved in-service inspection and repair capability. Moreover, a targeted high level of economic performance will be looked for, mainly based on higher burn-up rate, efficient energy conversion system but also simplification of systems.

Currently, in France, a moderate (1500 MWth, ∼600 MWe) power demonstrator named ASTRID (Advanced Sodium Test Reactor for Industrial Demonstration) has been proposed and endorsed by EU.

2.1 ASTRID

The French Act No. 2006-739 dated 28th June 2006 on the sustainable management programme for radioactive materials and waste stipulates the commissioning of a Generation IV reactor by 2020. It was therefore decided by CEA to launch, in cooperation with industrial partners, the conceptual design of a Sodium-cooled Fast Reactor (SFR) – the only Gen IV reactor technology that is sufficiently mature and benefits from considerable feedback both in France and overseas, associated with the potential to attain Gen IV criteria. This project has been called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration).

Today, the reference configuration of ASTRID [1] is a pool concept like for the previous French SFR PHÉNIX and SUPERPHÉNIX operated by NERSA (European consortium: EDF-ENEL-SBK), then European project EFR, BN600 in Russia or PFBR, being built in India. Other main features of this SFR demonstrator are: an intermediate sodium circuit, oxide fuel for starting cores, transmutation capability, and in-sodium fuel handling. Thanks to it design the reactor should meet high level expectations in terms of safety demonstration. ASTRID sets out to demonstrate the progress made in SFR technology on an industrial scale by qualifying innovative options, some of which still remain open in the areas requiring improvements,




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especially safety and operability. The open questions of ASTRID design are, for instance, type of energy conversion system (the reference is water/steam, an alternative is N2), number of intermediate loops, devices to eliminate severe accidents, corium catcher technology, steam generator materials and technology, innovative technologies for sodium fires detection and extinction. There are also innovative options to be tested during the R&D phase like carbide fuels or SiC-SiC materials. Many of these options will be decided during the pre-conceptual design before the end of 2012, however some of them will remain open and the decision will be taken during the conceptual design phase of ASTRID. The conceptual design will be delivered by the end of 2014.

ASTRID will also be used as a test bench for advanced inspection and repair techniques, aiming for a level of safety that is at least equivalent to that of the EPR (third generation of nuclear reactor systems), with progress made in SFR-specific fields. Moreover, the reactor will be designed to investigate and to demonstrate the feasibility of transmutation of radioactive waste. It will be replacing the PHÉNIX plant in terms of an irradiation reactor so as to test homogenous and heterogeneous minor actinide recycling modes, whose industrial interest are currently being assessed.

2.1.1 Main requirements for ASTRID

As the prototype of this reactor technology, ASTRID has the main objective of demonstrating advances on an industrial scale by qualifying innovative options in the above-mentioned fields. It must be possible to extrapolate its characteristics to future industrial high-power SFRs, particularly in terms of safety and operability.

ASTRID will nevertheless differ from future commercial reactors for the following reasons:

• ASTRID will be a 1500 MWth reactor, i.e. generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and main components. This level will also compensate for the operational costs by generating a significant amount of electricity. A sensitivity study will be conducted on this power level.

• It will be equipped for experiments. Its design must therefore be flexible enough to be able to eventually test innovative options that were not chosen for the initial design. Novel instrumentation technologies, new fuels and even new system components will be tested in ASTRID.

• It will be commissioned at approximately the same time as Generation III power plants, which means that its level of safety must be at least equivalent to these reactors. Focus will nevertheless be placed on validating safety measures enabling the future reactors to ensure an even more robust safety. This means taking into account core meltdown accident conditions from the design phase, by improved preventive measures, the ultimate objective being its ‘practical elimination’. The impact of a mechanical energy release accident will be integrated at the design level if the demonstration of the core meltdown accident elimination is not sufficiently robust.

• The capacity to inspect structures in sodium will be improved, with efforts especially focusing on structures ensuring a safety function.

• The risks associated with the affinity between sodium and oxygen will be reduced by improved preventive measures (detection systems, innovative components, improved operating procedures…): sodium fires and sodium/water reactions.

• ASTRID’s availability objective is below that of a commercial power plant to its experimental character. However, the options chosen must demonstrate that a higher level of availability can be reached when extrapolated.




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• Without being a material testing reactor (MTR), ASTRID will be available for irradiation experiments like those conducted in PHÉNIX in the past. These experiments will help to improve the performance of the core and absorbers, as well as to test new fuels and structural materials, such as carbide fuel and oxide dispersion steel (ODS) cladding. ASTRID should be equipped with a hot cell for examining irradiation objects, built either in the plant or nearby.

• ASTRID will be able to transmute radioactive waste so as to complete the industrial demonstration of this technique for reducing the volume and lifespan of final radwaste.

• Though future fast reactor plants intend to be breeders, ASTRID will be a self-breeder considering the current nuclear material situation, while being able to demonstrate its breeding potential.

• ASTRID must also integrate feedback from past reactors, especially PHÉNIX and SUPERPHÉNIX, while being clearly improved and belonging to Generation IV, take into account current safety requirements, especially in terms of protection against both internal and external acts of malevolence, as well as the protection of nuclear materials, meet the latest requirements in terms of proliferation resistance, and maintain its costs by following a value analysis approach from design.

2.2 Support research facilities and projects

2.2.1 SFR technology development

The ADRIANA work package WP2 has drawn the following priorities for European R&D experimental facilities to support development of SFR in the EU [1], [2]. They are presented here as technological platforms implemented in various European research organisations. For each platform, the facilities to be built are given together with their goals, current status, and availability.

1) SUSEN Platform (CV Řež)

SUSEN–Na loop (2015)

Goals: • Study of heat and mass transfer in cover gas, aerosols behaviour and deposits in narrow gaps

• Instrumentation, with relevant operative conditions and slab geometry

Status: Conceptual design by CV Řež, with support of CEA. Partial funding from European structural funds foreseen.

SUSEN–SC-CO2 loop (2015)

Goals: • study of alternative SC-CO2 Brayton cycle energy conversion system • SC-CO2 technology development Status: Detailed design by CV Řež and manufacturing on-going. Partial funding from European structural funds foreseen.

2) KIT Platform (KIT Karlsruhe)

KASOLA + ALINA facility (2012, then 2014 (bundle))

Goals: • 2D slab pool model of SFR, represents heat exchanger, DHR-system, contains flow sensors; detailed flow measurements




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• test of full length 19 rod bundle; scaled heat transfer • tests of natural convection • possibility to test FA, dummy assemblies, control rods / shutdown system Status: Detailed design by KIT and manufacturing going-on. Funding provided for Pool configuration, but to be funded for bundle configuration.

3) Liquid Metal Platform (ENEA Brasimone)

ISA facility (2014)

Goals: • tests to characterize physical effects induced by sodium-water interaction in the intermediate loop including Steam Generator Unit. Status: Conceptual design by ENEA to be launched, with support of CEA. Funding to be identified.

4) AMPERE Platform (IPUL)

STL-300 facility (2014)

Goals: • Study of materials in extreme conditions. (possibility to work under vacuum or intertization) • Qualification of innovative heat-exchangers,… (under vacuum) Status: Refurbishment and up-dating to be carried-out. Partial funding from European structural funds assumed.

D100 facility (2014)

Goals: • Test of EM pumps and their calibration Status: Conceptual design by IPUL to be launched. Partial funding from European structural funds assumed.

FCS-100 facility (2014)

Goals: • Tests and calibration of instrumentation (flow-meters…) Status: Conceptual design by IPUL to be launched. Partial funding from European structural funds assumed.

5) Na School (ESML) (CEA)

SUPERFENNEC facility (2014) (refurbishment)

Goals: • Training to Na facilities Status: Refurbishment to be performed. Partial Funding.

Mock-ups for Na

Goals: • Na fire, • Na-water interaction, • Na properties… Status: Detailed design of mock-ups to be carried-out. Funding to be identified.

Simulator for SFR operation

Goals: • training (steady-state, transients, start-up & shut-down…) Status: Requirements to be defined, Funding to be identified.




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6) DRESDYN Platform (HZDR)

DRESDYN facility (for Dynamo and Nuclear systems) (2015)

Goals: • Basic studies on Na boiling and gas entrainment • Tests for new instrumentations and measurement techniques (flow rate sensors, local velocity measurements by ultrasonics, gas bubble detection, CIFT) • Tests for new instrumentations and measurement techniques (visualizations by ultrasonic technologies and magnetic field tomography) Status: Detailed design by HZDR carried-out and manufacturing going-on. Funding provided for loop, but to be funded for mock-ups, instrumentation…

7) CHEOPS Platform (CEA)

In addition to the contribution to the qualification of the main SFR components, this platform is considered as a strong contribution to the industrial competitivity of Europe within ESNII.

N-SET facility (2016)

Goals: • Qualification of ASTRID SGU, detection systems Status: Conceptual design by CEA being performed. Partial funding from National Loan.

N-TRIPOT facility (2016)

Goals: • Qualification of Fuel Handling chain, of large components Status: Conceptual design by CEA being performed. Partial funding from National Loan.

NADYNE facility (2016)

Goals: • Fuel assemblies qualification, (TH characterization scale 1; possibility to realize fast thermal transients), qualification of large EMP Status: Conceptual design by CEA being performed. Partial funding from National Loan.

8) PAPIRUS Platform (CEA Cadarache)

This platform is the currently existing platform located in Cadarache. Several facilities are or will be refurbished in the next future. Here the six most important facilities to be built, considered of highest priority, are reported here.

NBB (Na boiling bench) (2014)

Goal: • Characterization of Na boiling for core safety assessment and development of detection system. Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

TRIBONA facility (2014)

Goal: • Long term behaviour in Na of rotating or moving components, with regards friction and wear, qualification of coatings, arms (Remote Handling Systems) Status: Conceptual design by CEA being performed. Partial funding from National Loan.

NECRINA facility (2015)

Goal: • Qualification of innovative quality control systems (cold traps…) • Qualification of chemical instrumentation i.e. plugging-meters, LIBS…. Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.




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NAUSOUT (2015)

Goal: • Development & Qualification of non intrusive instrumentation (US, acoustic, out of Na) Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

TRANSCONTA facility (2016)

Goal: • Study of corrosion and mass transfer in relevant conditions in primary vessel with selected structural materials. Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

N-GRIGNOTIN (2016)

Goal: • Na-H2O interaction and wastage effects, tests on Acoustic Detection Systems, … Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

9) PLINIUS-Na platform (CEA)

FOURNAISE facilities (1 and 2): (2016)

Goals: • Capability to study interaction Na-corium Molten corium (50 – 300 kg) • Capability for study of debris formation, code validation • Potential use for core-catcher design studies Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

10) PLATEAU Platform (CEA)

Water mock-ups (2013)

Goals: • Thermal hydraulic and mechanical analysis for design or safety purpose, in upper and lower plenum (Permanent or transient thermal stratification, Above Core Structure behaviour, thermal fatigue, gas behaviour…) (water system: equipped with dedicated mock-ups; scale: 1/10 to 1/3,). • Characterization of gas ingress (vortex…) in the primary Na (air+H2O mock-up) Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

11) ATHENA Platform (CEA)

Water mock-ups (2013)

Goals: • Validation of assembly hydraulics in water : flow-structure interactions, flowrate, velocity, pressure drop, vibration effects, wear, cavitation and rod-drop kinetics, gas behaviour, Validation of inter-assembly flows and pressure drops Status: Conceptual design by CEA to be carried out. Partial funding from National Loan.

2.2.2 Safety research

In parallel, research needs are also required for safety assessment of the ASTRID prototype and future sodium-cooled fast reactors. Though these safety assessments are strictly tied to design features not yet fixed in details for ASTRID (especially regarding the innovative designs), some common research needs for safety issues were clearly identified. They aim to ensure that the main general safety objectives are fulfilled through:




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• minimizing the risks attached to the sodium use,

• practically precluding large energy release in case of an extreme design basis accident,

• minimizing the risk of severe accidents and the vulnerability to external events and aggressions,

• diversifying the safety systems,

• developing an improved instrumentation for early detection of abnormal situations and in-service inspection and repair,

• assessing the impact of minor actinide bearing fuels.

The R&D efforts should include code development, validation and qualification as well as experimental programmes to be performed either in out-of-pile facilities or in-reactor testing facilities.

The safety needs relevant for future SFR projects were widely analysed within the TAREF (Task on Advanced Reactors Experimental Facilities) group that identified the fuel behaviour under operation, under incidental/accidental transients within the design basis and under severe accident as the first priority issues and top-tier safety needs. In particular, the SFR fuel melting with a potential risk of generalized core melting is a key safety issue which implies that fuel melting occurrence conditions have to be known and predicted with a sufficient accuracy all along the in-reactor fuel life. Although the experimental database on mixed U-Pu oxide fuel of moderate burn-up is extensive, uncertainties still remain and extrapolation to other conditions (e.g. higher burn-up fuels) and other types of fuel (e.g. carbides, with minor actinides…) that are anticipated either for ASTRID or future SFR projects types of fuel, is limited.

Within the safety related experimental tests that will have to be performed in support of ASTRID and future SFR projects safety analyses, it will be possible to perform slow transient tests (though not in fully representative conditions) in the existing reactors . The severe accident related studies will have to be performed in TRIGA-ACPR (whose suitability for fast reactor transient tests however still needs to be assessed) and CABRI reactor, extensively used in the past for these purposes but whose availability at long term for fast reactor tests still have to be studied.




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3 LFR technology

3.1 MYRRHA

SCK•CEN, the Belgian Nuclear Research Centre in Mol has been working for several years on the design of a multi-purpose fast-neutron irradiation facility in order to replace the ageing BR2 reactor, a multi-functional materials testing reactor (MTR), in operation since 1962. MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications), a flexible fast spectrum research reactor (50–100 MWth) is conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying core with MOX fuel, cooled by liquid lead-bismuth (Pb-Bi). This experimental technology facility will serve as a basic research infrastructure for both the fast reactor as well as ADS applications.

MYRRHA will be operational at full power around 2023. During the 2010–2014 period the following items will be accomplished:

• the Front End Engineering Design (FEED) and the associated R&D programme. • the licensing process. • the set-up of the international consortium.

Construction of the facility and assembly of the components is foreseen in the period 2015–2019. Three years (2020–2022) are foreseen for the full commissioning of the facility. The total investment cost is currently estimated at 960 M€.

Figure 2: Scheme of MYRRHA.

MYRRHA will be a multipurpose flexible fast spectrum irradiation facility and will contribute in various fields:

• Sustainable energy: development of critical fast spectrum reactors • Sustainable fission energy: to demonstrate the physics and technology of an Accelerator Driven

System (ADS) for transmuting long-lived radioactive waste • Enabling technologies for renewable energies: production of neutron irradiated silicon




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• Health care: production of radioisotopes for nuclear medicine • Science: fundamental research for the generation of new expertise in various fields.

As a direct consequence of the desired high flux levels, and hence high power density, a compact core is needed and therefore, the central hole in the core which houses the spallation target should be of limited dimensions (∼ 10 cm). SCK•CEN opted for liquid metal as a coolant. Lead-bismuth eutectic (LBE) was selected due to its low melting temperature (124.5°C), allowing the primary systems to function at rather low temperatures. The sub-criticality level of around 0.95 has been considered as an appropriate level for a first of a kind medium-scale ADS.

MOX fast reactor fuel technology has been chosen due to the large experience in Europe and in particular in Belgium. A maximum plutonium enrichment of 35% was considered based on the available manufacturing and qualification experience by Belgonucleaire in the past.

To profit from the thermal inertia provided by a large coolant volume, we opted for a pool-type system in which the components of the primary loop (pumps, heat exchangers, fuel handling tools, experimental rigs, etc.) are inserted from the top in penetrations in the cover. The loading of fuel assemblies is foreseen to be from underneath, which is not the classical approach of the sodium fast reactors. The reasons behind the approach are firstly to keep a large flexibility for the experimental devices loading from the top and secondly, from the safety point of view, the fact that all structures including the spallation module are in place before starting the core loading.

The pool vessel, which contains the MYRRHA core internals, is located in an air-controlled containment environment. Furthermore, several factors lead to the decision to design both operation and maintenance (O&M) and In-Service Inspection & Repair (ISI&R) of MYRRHA with fully-remote handling systems.

Experimental positions for fuel or material irradiations are the following:

• In IPS closest to spallation target (6 positions) o dpa: 18 dpa/EFPY o appmHe/dpa: 0.30 – 0.40 o loops can be installed with separate heating/cooling, allowing for the appropriate

environment (lead, LBE, Na, NaK, gas,…) and the desired irradiation temperature • Close to target module for fusion materials

o dpa: about 30 dpa/EFPY (360 EFPDs) o appmHe/dpa: 6.4

• In hottest fuel assembly o dedicated irradiation fuel assembly, but no “loop-type”, limited volume o results in hottest pin clad:

§ dpa: about 30 dpa/EFPY § appmHe /dpa: about 3.8

Besides these very high fast flux irradiation positions, about 10 additional irradiation channels are available in the reflector. In addition, facilities are foreseen for medical radio-isotope production (99Mo), consisting of an in-core water loop surrounded by a stagnant gas volume and a Cd screen (to prevent thermalized neutron re-entering the fast core). For silicon doping, a large out-of-core irradiation device is being designed.

3.2 ALFRED

Recent developments carried out in last decade (EU FWP-5, FWP-6 and FWP7) in the frame of Accelerator Driven Systems (ADS) as well as Fast Reactors, have evidenced Lead as an emergent technology potentially




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complying all Generation IV goals (with specific reference to nuclear cycle sustainability). Thanks to the Lead high boiling point together with the inert nature of the coolant, favourable neutronics and safety characteristics has attracted the attention of Research Organization and Industry as a credible alternative to the other fast neutron reactors technologies under scrutiny and development. It has however to be pointed out that the state of the art of the technology is not yet fully developed. The most challenging issue to be thoroughly analyzed is the compatibility of the structural materials.

It has to be highlighted that all the actions taken so far for the development of the LFR technology have been part of the EU FWPs complementing Member States national programmes, representing one of the most relevant EU investment for fission research in the last decade.

During FP6, the conceptual viability of a large scale LFR (600 MWe) in terms of economics, safety as well as sustainability have been demonstrated by the outcome of the ELSY (European Lead System) project.

Following this effort, the FP7 project named LEADER was started in April 2010. Its main aim is to provide the conceptual design of a demonstrator (ALFRED – Advanced Lead Fast Reactor European Demonstrator) based on already available industrial technology in order to shorten as much as possible the construction phase.

The construction and the operation of LFR demonstrator (ALFRED) would allow the accumulation of important experience derived from design and operation activities and to achieve an adequate level of confidence on the path to the prototype, especially from an industrial and economical point of view. Hence, a real base for the use of LFR technology is expected to be achieved with the construction of a demonstrator, this is a key steps towards industrial deployment.

It is supposed that ALFRED will be developed during 2010–2025 and the following phases are planned: conceptual design, decision point (2013), detailed engineering, specifications drafting and tendering, construction of components and civil engineering, on site assembly and commissioning. Design phase is expected to start in 2014 and the construction after 2017. The start of operation is planned for 2025. Some feedback from LFR ETPP (MYRRHA) experience is expected to be used to optimize the final design of ALFRED.

Since ALFRED shall be based on available industrial technology it uses wrapped hexagonal Fuel Assembly weighed down (at list with primary pumps off) and supported by a lower support grid anchored to the inner vessel. FA is maintained in position laterally by a cylindrical inner vessel and from the top by springs connected to the reactor roof. No refueling machine is foreseen inside the reactor vessel. The steam generators are classical bayonet type featuring double walls with continuous monitoring to avoid water-lead interaction in case of SGTR accident. ALFRED is equipped with two redundant and diverse DHR systems operated in a passive way. Primary system features low pressure drop to enhance natural circulation which permits to have a grace time higher than 30 min in case of unprotected loss of flow.

The main parameters of ALFRED selected in the FP7 LEADER project are presented below:

Power 300 MWth

Thermal efficiency 40%

Primary Coolant Pure lead

Core inlet temperature ∼400°C

Core outlet temperature ∼480°C

Fuel MOX, considerations for nitrides (with and without MA)




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Fuel cladding temperature Maximum 550°C

Cladding material T91 as a reference, other materials to be investigated

Primary system Pool type, compact, forced circulation (at power), natural circulation for DHR

Primary system pressure loss < 0.15 MPa (Unprotected Loss of Flow grace time >30 min)

Steam generators 4 or 8, integrated in the main vessel

Primary pumps 4 or 8, integrated in steam generator, suction from hot collector

Secondary cycle Water-superheated steam at 180 bar, 450°C

Reactor Vessel Hanged, Austenitic stainless steel

Safety Vessel Anchored to reactor pit

Reactor Internals Removable

Seismic design provisions 2D isolators supporting the reactor building

Table 1: The main parameters of the ALFRED prototype.

ALFRED has the mission to demonstrate the correct operability of all heat transport systems including the power production system by connection to the grid. ALFRED is a scaled down version of the ELSY concept (industrial reactor concept) with similar (but not necessarily identical) characteristics.

The main objectives of the LFR Demonstrator are:

• to achieve an high safety standard and to enhance non-proliferation resistance; • to assess economic competitiveness of LFR technology, including high load factors; • to demonstrate better use of resources by closing the fuel cycle; • to validate materials selection.

Moreover the demonstrator shall confirm that the newly developed and adopted materials, both structural material and innovative fuel material, are able to sustain high fast neutron fluxes and high temperatures.


The work package WP3, by the deliverable D3.2, has identified and catalogued 33 facilities which either exist or are under advanced design, having been financed by national governments. The categorization of these facilities in terms of their use and status has been performed as well.

The afore mentioned work allowed to identify the list of facilities which are necessary in the next years to support the LFR development comprehensive of pilot plant (MYRRHA) and demonstrator (ALFRED) requirements. The synthetic list contains: a) existing infrastructures which need some upgrade, b) new facilities which are in advanced construction phase and c) the completely new facilities which have been only conceived so far. The synthetic list (reported below) indicates the facilities and projects having priority within the EU in view of the LFR development.




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Facilities Technical goals Present status and funding Availability

COMPLOT Fuel assembly thermal-hydraulic and Hydrodynamics. Operability of the control rods Spallation target thermal-hydraulics

Engineering design. Partial Funding by Belgium

End 2012

HELENA Fuel assembly thermal-hydraulic in forced convection Corrosion-erosion of pump impeller materials Valve Qualification in Lead Corrosion – erosion on structural material

Engineering design. Partial Funding by Italy

End 2012

NACIE Fuel Assembly thermal-hydraulic in mixed convection. Transition from forced to natural circulation.

Facility available. Upgrade to be implemented

End 2012

CIRCE Large scale testing of TH LBE-components. TH integral testing and transient analysis

Facility available. Upgrade to be implemented

End 2013

E-SCAPE Pool thermal-hydraulics and CFD validation


End 2012

DEMOCRITOS Pool thermal-hydraulics in water Engineering design. Partial Funding by Belgium

End 2012

ATHENA I Steam generator tube rupture HLM/Water interaction code (SIMMER) assessment

Engineering design. Partial Funding by Italy

End 2013

ATHENA II • Large scale testing of thermal hydraulic components

• TH integral testing of primary and secondary loop.

• FA Manipulation • Safety Assessment (flow blockage,

flow induced vibration, control rod insertion, frozen flow path, gas trapping through the core)

Conceptualization phase in Italy. No funding presently foreseen

End 2014

HLM Pump Test Loop

Hydraulic performances endurance and reliability of LBE pumps

Conceptualization phase in Italy. Possible location in Romania. No funding presently foreseen

End 2014

POIROT (test unit is INTRIGE)

Integral test for in-vessel remote handling

Conceptualization phase in Belgium. No funding presently foreseen

End 2014

RHAPTER Remote handling test components Under construction in Belgium. Partially funded.

End 2011




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Seismic core damage facility

Earthquake testing of core assembly and structures for damage assessment (possible modification to existing facility)


Beyond 2014

Facility for fuel coolability during refueling

To assure experimentally the capability to cool fuel elements during refueling for a specific flask design and cooling system (forced and/or passive) ) (to be merged into the ATHENA II goals)


Beyond 2014

Lilliputter-2 To test the HLM technology components: valves, fittings, filters, seals, thread connections…


TELEMAT Corrosion in in lead to high temperature (up 700°C)

Foreseen refurbishment of facility in Germany

End 2011

Electra Small reactor for Education and training (first phase – electric heating)

Engineering design. Partial Funding by Sweden

CRAFT Corrosion in non-developed flow large test section conditions (multiple pin experiment)


End 2012

LIMITS 3 HLM Materials Fatigue Engineering design. Partial Funding by Belgium

Middle 2012

LIMITS 4&5 HLM Materials fracture toughness Engineering design. Partial Funding by Belgium

End 2012

Facility for creep fatigue

HLM Materials Creep fatigue Engineering design by Italy. No funding foreseen

End 2012

Access to Russian BOR60 reactor

Irradiation damage of materials under fast spectrum

Two test campaigns are ongoing, one funded by Italy and one by Belgium

End 2012

Facility for chemistry control of cover gas (HLM Lab)

Fission gas and spallation/activation products release and capture

Engineering design by Belgium. Partial Funding by Belgium.

End 2011

HELIOS III Coolant chemistry control based on gas oxygen control


End 2011

Mass transport loop SCK (MEXICO)

Mass trans & impurities deposition rates

Conceptual design by Belgium. Partially funded

End 2013

MYCENE Chemistry control proof principle loop in LBE

Conceptual design by Belgium. Partially funded.

End 2013

Mass transport loop KIT

Mass trans & impurities deposition rates

Conceptual design by Germany. No funding presently foreseen

End 2013

Hot cell + furnace with O2 control (ITU, NRG, Chalmers)

Fuel – coolant interaction Conceptual design by Germany, Holland and Sweden. No funding presently foreseen

End 2013




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Facility for core melt propagation

Core melt propagation and fuel dispersion in severe accident (sub-assembly scale on a reactor).

Conceptual design by Italy and Belgium. No funding presently foreseen

Beyond 2014

Table 2: Lead and LBE facilities and projects – technical goals, status, funding, and availability.

The synthetic list has been traced on basis of the R&D plans of MYRRHA and ALFRED and is a direct consequence of the identification of needed infrastructures which was reported in deliverable D3.2.

Due to the lack in Europe of fast neutron irradiation facilities the list includes also the proposal to access to BOR60 reactor for material testing purposes.

3.4 Needed infrastructures

The important areas of investigations for MYRRHA and ALFRED can be collected in three groups: 1) Thermal-hydraulics and components tests, 2) Chemistry control, and 3) Materials. Each one includes issues which are related to a wide range of operating conditions. The general objectives are above all gathering experimental data for geometries and boundary conditions consistent with the prototype design and generating databases for computer code development and validation.

Within WP3 of the ADRIANA project, the following identification of the needed infrastructures for ALFRED and MYRRHA has been done on the basis of the LFR key topics (see [3] and [4]). Presently the needs are associated with the specific facilities of the synthetic list and express the rationale which is behind the synthetic list of the previous section 3.3.

1. Experimental facility for corrosion testing of materials in lead environment at high temperature: the long term materials need to be tested at temperatures as high as 700/800°C. TELEMAT.

2. Experimental facility for Steam Generator tube rupture having an interaction volume meaningful for testing a significant portion of SG and representative for “direct” extrapolation to reactor vessel dimensions. This facility shall provide a realistic reproduction of the tube rupture. ATHENA I.

3. Experimental facility for heat exchange components (SG, DHR,…). The facility will permit to test in a safe way components which will have a heat exchange as high as 1 to 2 MW in case of MYRRHA and 5 to 10 MW for LFR. CIRCE, ATHENA II.

4. New facility or, alternatively refurbishment of the existing ones, for investigating pool thermal hydraulics of HLM. The alternative option may be selected following an upgrade of the instrumentation. E-SCAPE; DEMOCRITOS; ATHENAII.

5. A facility to perform experimental qualification, in full scale, of the fuel elements, manipulation and core manipulation in normal and accidental conditions. This should have multipurpose functions. INTRIGE.

6. Facility to support the designer with full scale basic tests (e.g. (i) operability of the handling machine, (ii) operability and insertion speed of the control rods, (iii) capability to cool fuel elements during refuelling) INTRIGE, Facility for fuel coolability during refueling, COMPLOT.

7. Facility to test relevant mechanical components before installation in the Demo. ATHENA II (for large components). Lilliputter 2 (for valves and tube fittings).

8. A facility for studying the sources of core damage events and investigating severe accident assessment. NACIE, Facility for core melt propagation.




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9. Experimental facility to perform integral qualification tests of the main coolant pump. Equipment shall include qualification instruments for thermal hydraulics, vibration dynamics, mechanical forces, overall performances. HLM Pump Test Loop, HELENA.

10. An experimental facility for corrosion tests, having larger test section (larger diameter) than those currently available. This facility should be able to test the performances of a specific number of pin simultaneously with the coolant flowing up to 2 m/s. CRAFT.

11. A coolant chemistry facility for LFR design investigating the coolant and a facility for studying the fission gas released in the cover gas. Facility for chemistry control of cover gas, HELIOS III, Mass transport loops, CIRCE.

12. Test facilities aimed at investigating the fuel coolant interaction (basic chemistry) and the fuel dispersion in primary system. Hotcell + furnace with O2 control.

13. Test facilities aimed at seismic testing. Seismic core damage facility.

In principle, it is observed that the diversity of the facilities for addressing the different topics and for testing the designs of both MYRRHA and ALFRED is satisfactory. Some concerns are raised on the availability of some of them in covering the requirements of MYRRHA and ALFRED plans on time. Indeed, some facilities might be overloaded for future experimental programmes.




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4 GFR technology A key milestone has been reached in 2007 with the end of the GFR preliminary viability phase. The preliminary viability of the GFR reactor system was confirmed. Further studies are now concentrated on the GFR fuel and its SiC cladding. The ALLEGRO demonstrator is planned to start with a first MOX core with metallic cladding which does not raise large R&D issues for the short term. Nevertheless, the safety demonstration of such a reactor is not completed and parallel experimental demonstrations are needed to qualify calculations codes, and later operational and accidental procedures following the licensing process.

The GFR development roadmap includes parallel viability studies of: (1) a moderate power demonstrator named ALLEGRO (< 100 MWth) without electricity generation as a necessary step towards an electricity generating prototype and further the head of series commercial reactor, then (2) the target commercial electricity generating reactor (∼2400 MWth) and its fuel; a lifetime of 60 years is aimed for the commercial reactor [5].

4.1 ALLEGRO

ALLEGRO is a project of an experimental GFR prototype which is being studied in a European framework. This experimental reactor with a thermal power around 75 MWth will not produce any electricity. Helium, a transparent and neutral gas, will be used as a pressurised primary coolant. At a reduced scale, ALLEGRO will have all the architecture and the main materials and components foreseen for the GFR, except the power conversion system. Its safety principles are those proposed for the GFRs: core cooling through a gas circulation in all situations, ensuring a minimal pressure level in case of a leak thanks to a specific gas-tight envelope surrounding the primary system. It will also contribute to development and qualification of innovative refractory fuel elements that can withstand high temperature levels. The objectives assigned to ALLEGRO demonstrator concern the following three domains:

• The major issue is to demonstrate and qualify at a pilot scale the key GFR technologies and to confirm the expected qualities of GFRs.

• The experimental reactor will also be used to perform irradiation testing under fast neutron spectrum.

• At least an additional loop implemented in the primary system will provide a pilot test capacity of high temperature components or heat processes.

In a logic where this reactor will be built before a GFR refractory fuel element is completely validated, such a validation being precisely one of the objectives assigned to the reactor.

A 10 MWth derivation from the main heat exchanger/cooling system could be developed to connect an experiment using high temperature heat coming out from the core, to simulate a process at a significant scaling. The table below gives an overview of the main core performance and the experimental capability:

Core Management Fast neutron flux Dose In core volume

25–30% Pu

75 MW

F1 – 660 EFPD or

F5 – 2000 EFPD

8.4 × 1014 n/cm2/s

(E > 0.1 MeV)

15 dpa/year 6 × 5 litres

Table 3: Main core performance and experimental capability of ALLEGRO.




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ALLEGRO will start-up with a core based on an improved technology: a pin type fuel with a metallic clad. Such a core will be designed to include later on several sub-assemblies of innovative technology (carbide fuel with ceramic clad) and to be finally completely set up with this type of fuel. This progressive approach will allow the qualification of new technologies under representative conditions.

ALLEGRO is designed for a core power density of 100 MW/m³. The first cores will consist of UPuO2 pins with 25% Pu, cladded with optimized 316-Ti stainless steel. Fuel pins will be the same as the one in the driver core of the PHÉNIX reactor. The fertile axial blanket will be only suppressed, which will allow taking benefit from all the knowledge acquired for many decades on this type of fuel. Six locations can hold an experimental ceramic fuel sub-assembly.

When the ceramic fuel will be qualified, after some 2000 EFPD to reach a maximum fuel burn-up of 8 at%, the substitution of the MOX core by a ceramic core will permit to raise the operating temperature at the outlet of the core. Performance of the ceramic cores are estimated to be well representative of a GFR core as it is foreseen today in terms of temperatures, damages to structures, and fuel burn-up.

At the present time ALLEGRO is under pre-conceptual design studies and its start-up is scheduled by 2025.


With respect to different technical areas, the following conclusions on needed experimental facilities for GFR development in Europe in support of the ALLEGRO programme can be drawn [5]:

High temperature materials behaviour

Facilities exist to fabricate and characterise at a laboratory scale SiC or composite fibre reinforced SiC (SiC/SiCf) , vanadium alloys, ODS steels, in particular at CEA Saclay and their continuous support is of course recommended. Furthermore, the development of ALLEGRO will need pilot scale fabrication capacities with larger investments and industrial partnership. The lack of fabrication capacities for the vessel material was also highlighted.

Thermal hydraulics

Integral System Test facilities like SALSA, HE-FUS3 and HELITE in air and helium are highly recommended and need therefore investments. Separate Effect Test facilities like ESTHAIR, HECO, HEBLO, and ITHEX are highly recommended. Only part of them is available today, however these facilities have to be adapted. HECO needs to be constructed.

Qualification of components

A large number of existing or planned facilities must be used for the test and qualification of important components (thermal barriers, seals, blowers, heat exchanger); big investments are needed, either to adapt the facilities to GFR needs, or to build the facilities themselves. In addition, a non nuclear specific facility to qualify the optical vision, which is essential for the GFR, either for the core instrumentation or for fuel handling is needed at high priority. This facility could be used to qualify the fuel handling process also.

Operation – Accidents

Several existing or planned helium loops that can fulfil the needs have been identified. In particular, a helium loop with a 10–20 MW order of magnitude heat capacity will be necessary on the ALLEGRO development roadmap, to qualify operational and accidental procedures at relevant scale.

Severe accidents




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MERARG can be used to measure gas released from fuel sample and for plate out of fission products on plate samples. A test facility investigating plate-out in a prototypical geometry of a given GFR component (for example turbine, compressor, heat exchanger,…) is not available.

PLINIUS is a European experimental platform dedicated to the study of severe accidents using large masses of prototypic corium (i.e. high temperature molten mixtures containing depleted uranium oxides prototypic of the melt that could arise during hypothetical severe accidents). It is currently used for LWR corium types but can be adapted to GFR. Nevertheless, such an adaptation cannot be defined before the completion of more analytical tests on the high temperature fuel and clad behaviour in severe accident conditions.

4.3 Summary of needs for GFR development

The materials development appears as a high project priority, especially because the SiC cladding concept design and the fabrication of first pins for potential irradiation in ALLEGRO in 2026 is very constraining. It is a continuing process anyway, because a full ceramic core could be fabricated around 2035 (the first ceramic core of ALLEGRO). In line with the materials characterizations, the severe accidents behaviour analytical tests on advanced GFR pins are also of a high priority. The safety demonstration of such a GFR reactor is not completed and parallel experimental demonstrations are needed to qualify calculations codes, and later operational and accidental procedures following the licensing process.

Regarding components and systems, big investments are needed, either to adapt the existing R&D facilities to GFR needs, or to build the new specific facilities. In addition, a non nuclear specific facility to qualify the optical vision, which is essential for the GFR, either for the core instrumentation or for fuel handling, is needed at high priority.

A helium loop with a 10–20 MW order of magnitude heat capacity will be necessary on the ALLEGRO development roadmap, to qualify operational and accidental procedures at relevant scale.

The completion of more analytical tests on the high temperature fuel and clad behaviour in severe accident conditions is necessary before possible adaptation of PLINIUS, a facility for severe accidents studies, for GFR needs.

The development of ALLEGRO will need pilot scale fabrication capacities with larger investments and industrial partnership.




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5 Crosscutting R&D activities

5.1 Instrumentation and ISI&R

In a nuclear reactor, adequate instrumentation for plant operation and safety, maintenance, service, inspection and repair is required. Fast systems, liquid metal cooled or gas cooled have higher requirements (see Table 4), due to the higher thermal loads, higher temperatures, high fast neutron flux, the opacity of liquid metals and last but not least the automatic reactor shut down or protection system. A balanced instrumentation is necessary to operate the reactor and to early detect deviations from normal operation as well as the origins of the deviations to allow adequate reaction. Therefore, the sensor systems composed of sensor, transmission lines, data acquisition system and user interface has to be reliable, accurate, even under harsh environmental conditions, corrosion resistant, and resistant to high neutron and γ flux. Furthermore, sensors should be miniaturized and qualified and able to operate from shut down to accidental conditions.

System SFR GFR LFR ADS (LBE)

Core inlet / exit temperature (°C) 350 / 550 300 / 850 450 / 550 270 / 410

System pressure (MPa) 0.2 – 0.4 7 – 10 0.2 – 0.6 0.2 – 0.6

Neutron fluence (n/cm²s) ∼1016 ∼1016 ∼1016 ∼3.7 × 1015

Typical in-core velocity (m/s) 5 – 8 80 – 120 1.5 – 2 ∼1.5

Risk of reaction with air / water high no low low

Table 4: Typical operation and environmental conditions of Gen IV systems.

One common need for all the nuclear reactor designs (fission or fusion reactors) is the understanding and development of materials and structures capable of functioning reliably for a long time in an environment described above. The main objective of qualification is to assure that the sensor system as described above is reliable over the envisaged operation time. Since this can only be done in prototypic or even hasher (benefit of shorter exposure times) environments, the need for a fast demonstrator becomes obvious.

Instrumentation for normal operation, plant transients and accidental situations are a key issue for innovative systems, since the operator requires in all situations sufficient and reliable information of the core, the safety systems and the power conversion system. Moreover, for the opaque liquids such as sodium, lead, LBE (SFR, LFR, ADS) maintenance operations have to be performed under visual control to avoid unintended damages of core internals. Under accidental conditions, a clear figure of the facility state is required for effective decision making. This should be assisted by a facility model based on an expert system.

For gas cooled systems (GFR) with high core outlet temperature (see Table 4), the control of the temperature field in the core or at least at the core outlet is an essential issue. Since the database for GFR is




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rather scarce, experiences from HTR instrumentation are used if applicable. For liquid metal systems, the large database and the operating experiences of SFRs in France, Japan, and India are taken into account.

Generally, for all purposes adequate instrumentation and control systems are available for out-of-core application. A large variety of measurement techniques exist for liquid metal systems, widely tested and qualified for sodium, limited for lead or LBE. For in-core applications the situation is much worse since very limited instrumentation is available to withstand the harsh environmental conditions.

One of the outcomes of deliverable D5.1 [7] was a quantitative overview of available instrumentation for the physical issues which is illustrated by Table 5. A green check mark ü means “fit for purpose” and a red check mark ü indicates that the technology is available, but R&D activities are necessary for adaptation and/or improvement to meet the specific needs. A question mark ? is an indication, that a lack of proven technology exists or that other limitation hinders application to reactor application, or which has not yet been addressed. The dash ― indicate that no information could be found.

SFR LFR ADS (LBE) GFR out-of-

pile in-pile

out-of-pile

in-pile out-of-

pile in-pile

out-of-pile

in-pile

Neutronics and Gamma ü ü ü ü ü ? ü ü

Pin failure detection

ü ü ü ? ü ? ü ?

Boiling detection ü ü ― ― ― ― N/A N/A

Thermal hydraulics data (S/A) ü ü ü ? ü ? ü ü

Fluid chemistry control ü ü ü ü ü ? ü ü

Displacements / Vibrations (loose parts)

ü ? ü ? ü ? ? ?

Leak detection of primary/secondary system enclosure

ü ü ü ü ü ü ü ü

In-vessel structure topology ― ? ― ? ― ? ― ?

ISI&R ü ü ü ? ü ? ü ü

Remote handling ü ü ü ? ü ü ü ü

Table 5: Overview of the applicability of available instrumentation.

As can be seen here for most of the issues to be measured, technical solutions exists for out-of-pile conditions or are presently in the status of improvement and qualification. For in-pile situations, the improvement and qualification has to be initiated and promoted, especially with respect to the ageing and size issues. Another issue is the very wide operational range required for in-core instrumentation for neutron sensors as well as for safety related instrumentation for the reactor protection system. Safety related instrumentation is focused on core monitoring and detecting early deviation from normal operation




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and/or local abnormal states, e.g. pin failure, local coolant voiding, sub assembly blockages, unintended control rod movements. These needs are on top of normal operational.

The status of the instrumentation is discussed in deliverable of WP5 [9]and in [2]and [8] and will not be discussed here. The main issue is to elaborate a roadmap for instrumentation improvement and qualification.

Most important issue of instrumentation and especially core instrumentation is to qualify innovative sensors for long term application in a fast reactor. The two levels of qualification, non-nuclear prototypic conditions and nuclear prototypic environment, can be performed in parallel. However, instrumentation has to withstand long term irradiation in a fast neutron field. In Figure 3, a roadmap for instrumentation qualification is proposed starting with available innovative sensors in 2011 and ending with a set of qualified instrumentation in 2021 or at least in 2025 depending on irradiation capabilities.

Figure 3: Roadmap proposal for qualification of core instrumentation.

Furthermore, Figure 3 gives some information about the costs of instrumentation development, improvement and qualification. The costs are directly connected to the irradiation costs; however, they can be reduced in combining irradiation with irradiation campaigns necessary for other components, materials or devices.

The sensor to be qualified should be installed together with a well known and proven system to allow on-line checks to determine long term degradation and to support destructive examinations tests after irradiation. The roadmap contains two tracks: 1) the high priority ones which are necessary to allow successful operation of a Gen IV prototype and 2) a long term aspect to increase diversity and to reduce sensor sizes and costs.

High priority instrumentation

It is quite clear that the first step is to demonstrate the high quality requirements of a Gen IV system, whose safety features are comparable or better to that of a Gen II or Gen II+ reactor. If this is proved in the




33

prototypes, the efficiency as well as the transmutation capability can be optimized. Therefore instrumentation must be qualified to guarantee a safe operation. Typical high priority instrumentation is:

• neutron flux detectors with a wide range,

• fast response and high robustness,

• flow sensors for an early and rapid detection of flow variations in sub-assemblies,

• divers sensors for temperature at core outlet with high reliability,

• fine spatial resolution and low interference with the fluid, and

• fast an reliable boiling detection for SFR.

For ISI&R the high priority instrumentation is under liquid metal viewing to assist the operator during maintenance and repair and to detect loose parts in the reactor vessel. The latter is difficult to achieve because of the required space for the detector.

Long term evolution

For the long term evolution a fast prototype reactor should be available to further test and qualify the high priority instrumentation in a prototypic environment as mentioned above. During the ADRIANA project, several issues were found which can only be solved in the long term. One prominent aspect is the ageing of in-core instrumentation, especially the electrical insulation of cables which are used in the vicinity of the core (inlet, outlet, blanket, etc.). Also vibration detection on S/A level and core level to early identify possible damage have to be qualified in a prototypic fast reactor.

The separation in short term on “high priority” and “long term” requires a periodic review of the status and the ongoing development, to correctly address needs and capabilities for qualification. To achieve this task, the instrumentation section should be updated in a two-year interval, starting from 2012. This would allow updating the overview and ranking of instrumentation, to eliminate outdated instrumentation, and to track the ongoing qualification process.

5.2 Material testing reactors, irradiation and fuel testing facilities

From the ADRIANA work packages on SFR, LFR/MYRRHA and GFR, the following needs for irradiation facilities (including experimental devices) and transient test devices were identified.

SFR:

Materials: The mechanical behaviour of materials for cladding and wrapper applications has to be investigated; for the cladding, austenitic or ODS steels are foreseen. The key issues are swelling under irradiation, the irradiation and thermal creep, the toughness, the embrittlement under irradiation, the evolution of the DBTT… but also their behaviour in nitric acid during reprocessing.

Fuel behaviour under operation and under incidental/accidental transients within design basis: The need of fuel pin irradiation capabilities under representative conditions of fast neutron flux in order to address fuel safety issue and to give basic information and validation data for fuel operation and transient behaviour is pointed out.

Fuel behaviour related to severe accidents: To support the required additional R&D work on both fuel safety and severe accident issues, experimental testing facilities are needed. In particular, with regard to




34

fuel accident behaviour, as the fuel transient evolution is mainly governed by thermal and mechanical effects intimately coupled with fission gas and solid fission products, only neutron heating mode can provide a reliable simulation of the events. The need for transient testing reactor with possible use of irradiated fuel pins under representative conditions of SFR accident transients is highlighted for addressing safety issues of high priority (no available European facility in short to mid term period). CABRI has been identified to be the dedicated facility in Europe to perform SFR accident transients and fuel safety studies within representative conditions. However, its availability at long term for SFR tests is not yet guaranteed.

LFR/MYRRHA:

Irradiation effects on materials: It is necessary to complement the database on irradiation performance of candidate materials:

• Corrosion in lead under irradiation • Irradiation embrittlement of selected materials (e.g. T91) • Irradiation creep (mechanical load, fission gas) • Swelling

Results of irradiation effects associated with HLM exposure are presently expected by GETMAT project. A facility able to perform these tests is the BOR60 reactor in Dimitrovgrad, Russia.

Advanced fuel R&D qualification: In a first stage, the focus will be on MOX fuel with 30–35% reactor grade Pu on which significant feedback exists from R&D for sodium reactors and FBR (PHÉNIX, SNR, MONJU, SUPERPHÉNIX):

• a database on normal fuel pin behaviour was developed: fission product release, fuel restructuring, densification and swelling, clad swelling, clad embrittlement, clad corrosion, clad fatigue, clad elongation and fuel/clad evolution;

• an assessment of the failure margin was made;

• envelope conditions for this fuel was established (normal operation conditions, anticipated operational occurrences and postulated accidents in a regulatory context).

Still, additional R&D will be required on the issues of clad-coolant interaction, clad-coolant-fuel interaction, corium-coolant interaction, He embrittlement (mainly for ADS), failed fuel pin behaviour and the maximum acceptable damage in the subassembly.

Pre-qualification should be done on laboratory scale, followed by qualification by irradiation tests of a prototype in representative conditions; the final demonstration is to be done in the ultimate demonstrator facilities (MYRRHA/ALFRED).

In a second phase, new fuels (MA mixed MOX fuel, nitride fuel, carbide fuel) will be needed: for these fuel types, all aspects related to fuel R&D will need to be redone.

GFR:

The fuel and core materials development process includes fabrication, then irradiation on samples in representative conditions, then irradiation on a single pin, then pin bundles and finish with irradiation of a sub-assembly (sub-assembly level) and further, irradiations of a whole core with advanced GFR technology (qualification at the core level).

Fuel irradiation on samples (fissile compound): The fuel irradiation on samples requires representative fast neutron fluxes, i.e. of the order of 1.2×1015 n/cm²/s (E > 0.1 MeV) which can be reached mainly in other fast




35

reactors like PHÉNIX (unfortunately shutdown), JOYO, MONJU (Japan), BOR60, BN600 (Russia) and in the future facilities MYRRHA or ALLEGRO (GFR project).

Clad irradiation on samples: The irradiation of clad materials is less demanding in terms of flux representative conditions, provided the level of damage dose reached is sufficient. A lot of Material Testing Reactors (MTR) can bring their contribution such as ATR (USA), HFR (Netherlands), BR2 (Belgium), OSIRIS (France), JHR (France, under construction).

Single pin irradiations: Again, representative fast flux conditions are necessary with, if possible, circulating helium coolant. Reactors that can bring contribution to this matter are: HFIR (USA), JOYO, MONJU (Japan), BOR60, BN600 (Russia), MYRRHA and ALLEGRO.

Irradiations on pin bundles, sub-assembly and whole core irradiations: These levels of qualification tests refer to the global fuel and S/A technology and must be preformed in relevant conditions either in terms of flux, doses/burn-up, and coolant environment. Only the ALLEGRO reactor can reach these conditions.

Analysis of the existing and future capabilities in view of the needs

The irradiation facilities and the associated experimental devices in operation, under construction and in planning phase were surveyed and a ranking exercise was made in order to identify the optimal installations for testing on short term and also to set the priorities for funding on a medium time scale (see Ref. [10]). The following general conclusions can be drawn:

• Although some partial irradiation tests can be performed in existing thermal spectrum facilities (with specially designed experimental devices to obtain more of less appropriate environmental conditions and/or neutron spectra), ultimately qualification tests need fast spectrum irradiation facilities. These do not exist today in the EU. Facilities like BOR60 in Russia can offer a temporary solution for the most urgent problems, at a cost of the order of 0.5 M€ per irradiation campaign (of one year) and with a finite capacity, especially for "foreign" experiments.

• At this moment and on short term, a number of high flux thermal spectrum MTRs are available (BR2, OSIRIS, HFR, LVR-15, …) with flexible irradiation facilities and possibilities to harden the spectrum locally (neutron screens, fast flux boosters), but always with limited fast neutron fluxes.

• Within about 5 years, the JHR will become operational and available for irradiation experiments. This is also a thermal spectrum reactor, with intrinsically the same limitations as above, but with somewhat higher (fast) fluxes. The PALLAS reactor is another reactor expected to become available for future testing, but again with a thermal neutron spectrum.

• On a longer term, beyond 2020, MYRRHA will offer very interesting irradiation capabilities (from 2023 on): it will be a flexible, fast spectrum irradiation facility, allowing for relevant irradiation tests in LBE, lead, sodium, gas,… at relevant temperatures. These testing capabilities will include not only material testing, but also fuel testing (including slow power transients).

• On a similar or even longer term, the fast reactor prototypes ASTRID, ALLEGRO, ALFRED are planned to be constructed. They will provide relevant conditions in terms of neutron spectrum and environment (coolant and temperature), but their main aim is to serve as a demonstration facility, which might limit their flexibility as an irradiation facility.

5.3 Laboratories with hot cells

The main R&D directions with implications on hot cells (for fuel fabrication/characterization and/or for post-irradiation experiments) can be summarized as follows:




36

SFR:

• development/characterization of mixed U-Pu fuels with novel designs as driver fuel for SFR prototypes

• safety of Minor Actinides (MA) oxide bearing fuels and associated recycling processes (treatment, refabrication) for MA recycling modes (heterogeneous and homogeneous),

• preparation/characterization of dense fuels (carbide, and possibly also nitride or metal) and associated recycling processes (treatment, partitioning, refabrication) to be qualified in a second phase of operation of the SFR prototype as alternative fuels for this type of reactor featuring enhanced safety

LFR/MYRRHA:

Very similar needs exist for LFR development.

• In the short-term, an essential goal is to confirm that ready-to-use technical solutions exist (with uranium-plutonium MOX without minor actinides as the main candidate), so that fuel can be provided in timing with the MYRRHA (and the ALFRED) operation. Existing experience in MOX fabrication indicates that fuel with 15–35 wt% enrichment in reactor grade plutonium in pellets of 90–97% of theoretical density can be produced. Such fuel was indeed fabricated, qualified and used in a number liquid metal fast reactors and experimental fast reactors cooled by sodium.

• In the mid-term, the possibility of using advanced MA (Minor Actinide) bearing fuels as well as of achieving high fuel burn-ups has to be assessed. New fuels with 2.5–5 at% and later even 10–20 at% of MA in heavy metal will have to be developed; nitride fuels are considered as an option for this purpose.

• In the long term, the potential for industrial deployment of advanced MA-bearing fuels and the possibility of using fuels that can withstand high temperatures to exploit the advantage of the high boiling point of lead will have to be investigated.

GFR:

Dense, high thermal conductivity fuels are required. Carbide fuel is selected as reference. The main challenges for developing GFR fuel mentioned in the SNETP SRA include:

• pre-selection in 2009 of a limited number of viable solutions of fuel elements (design, materials),

• selection of front end (fabrication and re-fabrication) fuel cycle processes,

• selection of a reference and a back-up fuel around 2013 based on the knowledge of materials properties derived from irradiation tests,

• optimization of the fuel through irradiations at higher burn-up, transient tests, and simulation of accidental conditions,

• preliminary design studies and simulation of normal and abnormal operating transients of plate and pin fuel sub-assemblies,

• confirmation of reference GFR fuel concept by 2019.




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Analysis of the existing and future capabilities in view of the needs

From an analysis of the existing hot laboratories capabilities and the needs expressed above, the following conclusions can be drawn:

• Quite extensive hot laboratory capabilities already exist in many European countries (see the overview of existing lab-scale facilities described in D6.1 [11] and D8.1 [12])

• With respect to specific needs for fast reactor R&D, emerging limitations can be identified in terms of flexibility (type of isotopes and compounds that can be handled, in particular with reference to Minor Actinides bearing fuels and to non-oxides fuels such as nitrides and carbides) and/or throughput.

• Flexible tools will be required in "pin-scale" facilities to produce a very diverse array of experimental pins for irradiation tests at experimental facilities during the first phases of the design of future fuels for SFR, LFR and GFR.

• Later, larger scale demonstrative irradiation tests will rely on the development of "pilot-scale" fabrication facilities and will require also industry involvement.

• Large-scale transmutation experiments could involve the recovery and handling of important amounts of Minor Actinides.

• Finally, PIE characterization facilities will have to address specific issues, like pure atmospheric conditions for nitride/carbide fuels, establishing materials properties under Gen IV relevant conditions (e.g. high temperatures) and compatibility with coolants (sodium in particular).

5.4 Zero power reactors

Generally, zero power reactors offer neutron physics measurements in clean state reactor, i.e. without burn-up. These measurements are typically divided to the following areas:

• Core safety assessment

• Nuclear data measurements

• Core calculations validation

On zero power reactors, a set of benchmarks can be carried out using the new materials and geometries intended for Gen IV nuclear power systems. Then the resulting libraries can be consequently used for various code validations.

5.4.1 Expressed needs for SFR, LFR and GFR

The ADRIANA work packages WP2 (SFR), WP3 (LFR), and WP4 (GFR) have identified their needs regarding zero power reactors regarding their respective R&D programmes and also capabilities of zero power reactors with respect to the Gen IV fast neutron reactor systems (see Ref. [13]).

For the SFR development WP2 has specified the following topics to be studied at the existing zero power reactor facilities:

Doppler & fuel expansion reactivity feedback

• The critical facility MASURCA is dedicated to the neutron studies of fast reactors lattices.

• The adaptability of the MASURCA core allows the validation of innovative SFR core designs: Physics

and neutron parameters of new lattices (low Pu enrichment, dense fuels, compact reflectors and




38

shielding, physics and neutron parameters of large fast reactor cores (zone decoupling, power map

control, absorber reactivity worth amplification, instrumentation issues).

Reactivity feedback due to sodium voiding

• The critical facility MASURCA is dedicated to the neutron studies of fast reactors lattices.

• Void sodium effect reducing (Na upper plenum, moderator introducing)

• Degraded or incidental configurations in which several material zones would be redistributed.

Regarding the LFR systems, the following areas and topics were identified by WP3 for zero power reactor experiments:

Validation measurements for nuclear data improvement

• Threshold processes: (n,n’) and (n, xn) reactions

• MA cross-sections

Validation measurements for licensing & operation

• Uncertainty reduction on cross-sections (Pb-MA-241Pu, 242Pu, in high energy range < 1 MeV)

• Determination of flux gradients in fast spectrum.

• Reactivity effects (voiding of HLM coolant in MOX core, secondary scram system).

Operation/control

• Coupling of accelerator, neutron source, Pb core.

• Perform a (low-power) coupling experiments: continuous beam, beam interruptions, pulsed experiments on fast sub-critical lead multiplying system with different sub-criticality levels.

And finally, for the GFR development in Europe WP3 summarised the research topics to be studied at zero power reactors as follows.

Existing calculation tools and nuclear data libraries have to be validated for GFR designs. The wide range of validation studies on SFRs must be complemented by specific experiments pointing the specificities of gas-cooled designs which are mainly: slightly different spectral conditions, innovative materials such as He and various ceramic materials (UC, PuC, SiC, ZrC, Zr3Si2,…), and peculiar abnormal conditions (depressurization, steam ingress,…).

The first ALLEGRO core will involve standard fuel and cladding, no sodium at all and a steel reflector. The demonstration of ALLEGRO core and subsequent gas-cooled reactor cores presumably will involve ceramic (carbide) fuel and structure elements, innovative materials and reflectors, resulting in a somewhat softer spectrum. Spectra even softer have been achieved in the MASURCA 1A’ and 1B experiments (metal fuel, no oxygen nor sodium, large amounts of graphite, fertile blankets) and in the ZPR-3 53 and 54 experiments (large amounts of graphite, steel reflectors).

Nevertheless, an analytic and systematic validation would require specific additional experiments resulting in a specific experimental programme ENIGMA proposed for MASURCA. It involves a spectrally representative reference core made of (U,Pu)O2, UO2, graphite and void rods, and a steel reflector. Several substitution patterns in the central region of the core have been devised to study the effect of spectral shifts (harder / softer spectrum), innovative material introduction (Si, SiC, Zr,…), low-grade Pu, heterogeneity effect of absorber worth, but also a few substitution patterns in the axial reflector based on




39

innovative reflector materials (ZrC, Zr3Si2,…). Possible simulation of He reactivity worth and steam ingress reactivity worth could also be addressed. A full-scale ENIGMA programme could last over 3 years. The current refurbishment process at MASURCA delays the possible performance of this programme after 2016. However, the reference core was charged in MASURCA in 2006 and was characterized just before the shutdown for refurbishment (so as to provide a clean characterization reproducibility check after refurbishment).

5.4.2 Conclusions for zero power reactors

There are several zero power reactors in Europe that are directly suitable for Gen IV research (MASURCA, VENUS-F) with their fast spectrum and ability to study the needed topics on given geometries. Also, there are more reactors (PROTEUS, LR-0) not having pure fast spectrum, but able to model it and having enough space to study selected topics.

Because of huge investment costs compared to the added value, there is currently no effort to build new zero power reactors in Europe. Instead, present reactors are being or have been recently refurbished to offer better R&D possibilities for Gen IV.

On the present level of assessment, provided that MASURCA will be recommissioned on time after its refurbishment, no major gap was found for the expressed needs from the SFR, LFR or GFR work packages.

5.5 Nuclear data measurements for fast reactors [14]

Model simulations, at a high level, play a crucial role for the successful and cost-effective development and safety assessment of advanced nuclear reactor systems. Construction of prototypes must evolve from initial concepts to competitive industrial applications via several intermediate validation steps. However, building of prototypes and performing test experiments is extremely costly and time consuming. In addition, the number of experimental reactors available worldwide is quite small and even declining, so that the opportunities for experimental verification of those new nuclear systems or their parts are very limited.

Therefore, model calculations must supplement experimental tests and validations. They are used for simulating system behaviour in a variety of possible configurations and running conditions, planning and interpretation of experimental tests and allow selecting progressively between different technological options using only a limited number of prototypes. Therefore, the Strategic Research Agenda of the SNETP has defined research needs for improved modelling tools and for the development of advanced simulation methods.

All important reactor aspects, such as reactor kinetics, heat production, induced radioactivity, radiation damage, dose rates, and fuel radio-toxicity, result from interactions between particles and nuclei. To model the nuclear system characteristics with sufficient accuracy, a precise simulation of all relevant nuclear reactions is necessary. For nuclear power production, neutron-induced reactions are definitely the most important interactions, besides reactions induced by light-charged particles and photonuclear reactions. Measurements of so called microscopic or differential nuclear data must establish the experimental basis for the development of model codes and for the adjustment of model parameters. These measurements provide also information for the basic physics understanding of nuclear reactions mechanisms and nuclear structure.

In view of the fact that all the three reactor systems dealt with in the ADRIANA project (SFR, GFR, LFR, and also an ADS system) use a fast neutron spectrum, the availability of accurate nuclear data in the fast energy region is the basis for precise reactor calculations. Since the focus of the activities concerning nuclear data




40

measurements was in the past mainly for a thermal spectrum, additional experimental measurements and their detailed analysis and interpretation are required for the fast spectrum systems and for innovative materials. This is particularly true for fuels containing minor actinides for their transmutation in fast spectra. To this end the European Commission supported in the 5th, 6th, and 7th EURATOM Framework Programmes several projects related with measurement issues of microscopic nuclear data, such as HINDAS, EUROTRANS-NUDATRA, EFNUDAT, CANDIDE, NUDAME, EUFRAT, ERINDA, and ANDES.

However, today the EU is faced with the situation that not many European laboratories are able to perform experiments for nuclear data measurements with the required accuracy. A recent survey on “Research and Test Facilities Required in Nuclear Science and Technology” [15] performed by an expert group for the Nuclear Science Committee of the NEA revealed a similar situation worldwide.

To cover all domains of microscopic nuclear data measurements with the required accuracies, a combination of the following three different types of facilities is needed:

1. Time-of-flight facilities for fast neutrons, 2. Charged-particle accelerators, and 3. Experimental nuclear reactors and critical assemblies.

A recurrent problem with nuclear data measurements is the difficulty in obtaining appropriate isotopic samples of (pure) raw materials, to be used as targets for the incident neutrons. This problem is exacerbated in the case of radioactive samples, minor actinides, some fission products, etc. This is an important point which demonstrates that having a good facility is not entirely sufficient in order to generate measured data; target materials are also needed. Thus specialised chemistry laboratories, equipment and staff are also required. For transmission measurements, the problem is even more acute as the masses of samples required are fairly high.

In the EU only the facilities operated by the JRC-IRMM and funded by the EURATOM programme are dedicated entirely to the area of nuclear data for fission energy applications. The EUFRAT Transnational Access project of the 7th EURATOM Framework Programme provides support for scientific groups wishing to perform measurements at the JRC-IRMM facilities (2008–2012). The n_TOF facility at CERN dedicates also a large percentage of its beam time to measurements of neutron data for fission applications, while another substantial part is devoted to astrophysics. The European Commission supported the n_TOF construction with the n_TOF-ND-ADS project.




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6 Costs and funding of European research infrastructures

6.1 SFR infrastructure cost estimation and funding

According to the Deloitte study [16], the total cost for the ASTRID demonstrator is estimated at 5000 M€; about 1 G€ is intended for the SFR technology innovations and component development, while about 4 G€ are counted for design and construction during the 2010–2022 period. The ASTRID financing plan envisages the following means of funding [16]:

1000 M€ (20%) national funding including research institutes,

1000 M€ (20%) private funding (industry),

1385 M€ (28%) EU incentives and grants,

1000 M€ (20%) EIB or EURATOM loan,

615 M€ (12%) tax exemption.

Thus, the financing of ASTRID will be realised through Public-Private Partnership (PPP). The long term deployment (2040) and overall risks imply the need for strong public involvement, however industry is willing to participate (up to 20%) and has already committed financial resources.

The income of the ASTRID programme is expected through the electricity production assuming the power of 600 MWe for a life time of 40–60 years and availability factor of 80% after a progressive evaluation phase in the first years of its operation. It is estimated that the electricity production will cover the reimbursement of a loan (corresponding to 20% of the investment) and to cover operation charges and experimental costs.

Regarding the planned research infrastructure in support of ASTRID, the estimation of the cost of investment (cost 1) for the 11 platforms mentioned in Section 2.2 and in some cases also the cost of mock-ups (cost 2) has been done. The cost of operation is not reported here, due to uncertainties in the estimation. It will be performed latter when more data on design and implementation will be obtained. Moreover, this cost mainly depends on the respective research centre and the work organization in these respective platforms. The investment cost is a total cost for a given platform and it does not take into account some already existing or anticipated sources of funding such as national funding programmes, European structural funds, etc.

These cost of the European sodium platforms are given in Table 6. For some platforms, those which are not existing today (i.e. SUSEN, AMPERE, ISA), the cost of some ancillary facilities needed to operate the platform, such as Na storage capabilities, Na purification devices, mock-up Na cleaning prior to re-use or dismantling, Na fire detection and fighting system, has been estimated and added to the cost of the facility itself.

Platform Institution Facility Existing / Project

Availability Cost (1) Cost (2)

SUSEN CV Řež Na loop and SC-CO2 loop

Project 2015 8 M€ ?




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PAPIRUS CEA Cadarache

Several Na facilities Existing &


PLATEAU CEA Cadarache

Several (H2O) facilities for FA

Existing & Project

2013 1 M€ 5 M€

ATHENA CEA Cadarache

Several (H2O) facilities for primary vessel thermo-hydraulics

Existing & Project

2013 2 M€ 5 M€

KASOLA

ALINA KIT Karlsruhe

FA bundle and primary vessel thermo-hydraulics

Existing & Project

2014 10 M€ 1 M€

PLINIUS SFR CEA Cadarache

FOURNAISE 1&2 Project 2016 25 M€ ?

Liquid Metal Platform

ENEA Brasimone

ISA Project 2014 5 M€ ?

AMPERE IPUL Salaspils

2 facilities Existing &


DRESDYN HZDR Dresden

2 facilities Existing &

Project 2015 21 M€ 6.5 M€

CHEOPS CEA Cadarache

N-TRIPOT, NADYNE, N-SET

Project 2016 70 M€ 30 M€

Na school CEA Cadarache

SUPERFENNEC + new mock-ups

Existing & Project

2014 5 M€ 0.5 M€

TOTAL 180 M€ 48 M€ +

?

Table 6: Cost estimation for European experimental platform for SFR development.

6.2 LFR infrastructure cost estimation and funding

The total cost for MYRRHA is estimated at 960 M€ and its financing plan supposes the following means of funding [16]:

40% from Belgium national funding,

15% from European countries national funding,

5% from non-European countries national funding,

30–35% from EU incentives and grants,

5–20% from EIB loan.




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The proposed financing models of MYRRHA are either the ERIC-like† or an international agreement.

For assumed 30 year operation life time, the incomes of MYRRHA are expected from commercial revenues (30–40 M€/year) linked to production of medical radioisotopes and neutron doped silicon. However, there is a risk that the facility will not meet the expected production performance.

As regards ALFRED prototype, according to the Deloitte study [16] its total cost is estimated today at about 1000 M€. Realization and financing via European consortium open to international participation is supposed today and will be specified in near future. The preliminary financing plan of ALFRED counts with the following contributions [16]:

50% from European countries national funding,

5% from non-European countries funding,

30–35% from EU incentives and grants,

10–20% from EIB loan.

The expected income of ALFRED, assuming 30 year operation life time and availability factor > 50%, will be from electricity production (30–40 M€/year). Nevertheless, like for all the prototypical nuclear power facilities, there is a risk of low availability factor in the first years of its operation.

The Belgium Federal Government has financed the preliminary phase of the MYRRHA Project with 60 M€ (up to 2014). Then, the Belgium Federal Government will support 40% of the total investment required for constructing MYRRHA, if SCK•CEN will constitute a Consortium which will finance the remaining 60% of the total investment.

The Italian Government will contribute in financing the construction of the large pool facility ATHENA-I, aimed at investigating the SGTR event (about 4 M€), as well as its upgrade for performing integral tests and qualification of components at full scale (e.g. ATHENA-II). The maintenance and refurbishment of Italian existing facilities which are intended to qualify components for MYRRHA and ALFRED is yearly funded by the government by means of 1.5 M€.

The Swedish government intend to partially support (2 M€) the design and construction of the ELECTRA reactor for education and training to be located at KTH.

The Rumanian government expressed, by official letter to ESNII, the willingness to host the ALFRED reactor.

No information is currently available regarding the planned funding of the other ADRIANA Project partners.

Since the deadline to issue the Preliminary Safety Assessment Report (PSAR) of MYRRHA, the LFR pilot system, is fixed at 2014, it is compulsory that all the testing activities which are necessary to produce the proofs to be reported in the PSAR are concluded before that date.

These testing activities represent the first experimental phase of the LFR development which includes all the infrastructures which are due by 2013. The foreseen investment consists of 30 M€, as reported in the following table. It is mainly covered by the direct funding by Belgian government and at a lesser extent by Italian government.

† ERIC stands for European Research Infrastructure Consortium.




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Investments

1st phase

Total operation

cost 1st phase

Test section cost

1st phase

TOTAL

1st phase

30.15 M€ 9.6 M€ 15.6 M€ 55.35 M€

The second phase will comprehend testing activities concerning:

• The beyond design basis accidents

• Several qualifications of industrial components

• The experimental proofs to license the ALFRED demonstrator

• Education & training

The costs of 2nd phase have the following amount:

Investments

2nd phase

Total operation

cost 2nd phase

Test section cost

2nd phase

TOTAL

2nd phase

46.1 M€ 10.75 M€ 18.99 M€ 75.84 M€

Costs Facilities

Availability date Investment Operation Test section

COMPLOT End 2012 0.7 M€ 0.3 M€/yr 1.4 M€

HELENA End 2012 1 M€ 0.5 M€/yr 0.5 M€

NACIE End 2012 0.5 M€ 0.2 M€/yr 0.3 M€

CIRCE End 2013 3 M€ 1 M€/yr 2 M€

E-SCAPE End 2012 1.5 M€ 0.3 M€/yr 0.25 M€

DEMOCRITOS End 2012 0.75 M€ 0.2 M€/yr 0.25 M€

ATHENA I End 2013 8 M€ 2 M€/yr 6 M€

ATHENA II End 2014 10 M€ 2 M€/yr 6 M€

HLM Pump Test Loop End 2014 4 M€ 1 M€/yr 2 M€

POIROT (test unit is INTRIGE) End 2014 0.55 M€ 0.25 M€/yr 1.2 M€

RHAPTER End 2011 0.5 M€ 0.25 M€/yr 0.04 M€

Seismic core damage facility Beyond

2014 2 M€ 1 M€/yr 2 M€

Facility for fuel coolability during refueling

Beyond 2014

1 M€ 1 M€/yr 0.5 M€

Lilliputter-2 0.25 M€ 0.25 M€/yr 0.25 M€

TELEMAT End 2011 0.5 M€ 0.5 M€/yr 0.5 M€




45

Electra (?) 6 M€ 1 M€/yr 1 M€

CRAFT End 2012 0.5 M€ 0.25 M€/yr 0.5 M€

LIMITS 3 End 2012 0.25 M€ 0.3 M€/yr 0.05 M€

LIMITS 4&5 End 2012 0.5 M€ 0.3 M€/yr 0.1 M€

Facility for creep fatigue End 2012 0.5 M€ 0.3 M€/yr

Access to Russian BOR60 reactor End 2012 0.5 M€/yr

(× 10 test campaigns)

2 M€

Facility for chemistry control of cover gas (HLM Lab)

Beyond 2014

1.8 M€ 1 M€/yr

HELIOS III End 2011 0.5 M€ 0.4 M€/yr

Mass transport loop SCK (MEXICO)

End 2013 0.35 M€ 0.25 M€/yr 0.5 M€

MYCENE End 2013 0.6 M€ 0.3 M€/yr 0.5 M€

Mass transport loop KIT End 2013 1 M€ 1 M€/yr 0.5 M€

Hotcell + furnace with O2 control (ITU, NRG, Chalmers)

End 2013 1 M€/yr 0.25 M€

Facility for core melt propagation Beyond

2014 20 M€ 3 M€/yr 6 M€

TOTAL 76.25 M€ 20.35 M€/yr 34.59 M€

Table 7: Cost estimation for European experimental facilities for LFR R&D.

6.3 GFR infrastructure cost estimation and funding

The total cost of the ALLEGRO prototype is estimated at 900 M€, assuming 300 M€ needed for technological development and design, 500 M€ for licensing and construction, and 100 M€ for sitting and owners costs [16]. Joint Undertaking or ERIC-like are the two considered models of realization and financing of ALLEGRO. According to the study Deloitte of funding opportunities [16], the preliminary financing plan envisages the following ways of funding:

14% from tax exemption,

35% from EU incentives and grants,

8% from private investors,

6% from EIB loan,

32% from national public research investors,

5% from hosting country public investments.

The incomes of ALLEGRO are expected from commercial revenues related to the irradiation capabilities of the facility (25+ year operation life time and availability factor > 80% are assumed).




46

In order to prepare and justify further investments, a ranking of needs and facilities considering project priorities, but also various criteria such as cost, readiness, etc. are presented in deliverable D8.3. As a result of this analysis, two phases were identified with the following proposals:

Phase 1 (2012–2017): limited investments, mainly relying on existing facilities operation for materials, thermal-hydraulics, components (thermal barriers, seals, compressor technology), severe accidents (analytical tests). For this phase, the costs were estimated as follows (in M€):

Construction or refurbishment cost

Total operation cost

Experimental device cost

TOTAL

15 M€ 83 M€ 24 M€ 122 M€

This phase would include:

• Operation of CEA, CNRS and University of Dresden benches for fabrication and characterization of ceramic clad materials,

• The operation of existing helium loops or benches: HEDYT, HETIQ, HETIMO, HELAN (CEA), HEBLO, HELOKA (KIT), HE-FUS3 (ENEA), HTHL (CV Řež)

• Investment on new loops to cover specific needs such as system code qualification allowing natural circulation, thermal barriers with circulating helium: HECO(CEA), SALSA(location to be determined), HELITE(location to be determined)

Phase 2 (2015–2020): complementary investments including new facilities to cover components development according to the licensing process of ALLEGRO (Fuel handling, Heat exchangers, He quality management, valves/check-valves, Operational/accidental procedures qualification) and potential severe accidents integral tests.

For this phase, the costs were estimated as follows (in M€):

Construction or refurbishment cost

Total Operation cost

Experimental device cost

TOTAL

32 M€ 39 M€ 12 M€ 83 M€

This phase includes in particular the construction and operation of:

• Instrumented Fuel Handling Pilot (IFHP) (location to be determined), • HELLO 20 MW large helium Loop (location to be determined), • CIGNE/HPC (CEA)adapted to GFR/ALLEGRO helium quality management needs,

• PLINIUS (CEA) adapted to GFR/ALLEGRO needs.

Costs

Technical area When

? Construction or refurbishment

Operation (total)

Experimental device

Sub-totals




47

Ceramic composite materials 2012

–2017

2 M€ 24 M€ 8 M€ 34 M€

Thermal-hydraulics 2012

–2017

8 M€ 29 M€ 10 M€ 47 M€

Components (thermal barriers, seals, compressor technology)

2012–

2017 3 M€ 18 M€ 4 M€ 25 M€

Components (Fuel handling, Heat exchangers, He quality management, valves/check-valves, Operational/accidental procedures qualification)

2015–

2020 22 M€ 33 M€ 10 M€ 65 M€

Severe accidents(analytical tests)

2012–

2017 2 M€ 12 M€ 1.6 M€ 15.6 M€

Severe accidents (integral tests)

2015–

2020 10 M€ 6 M€ 2 M€ 18 M€

TOTALS 47 M€ 122 M€ 35 M€ 204 M€

Table 8: Cost estimation for the GFR research infrastructure.




48

7 Roadmaps for European research infrastructure for fast reactor systems development

7.1 SFR technology roadmap – Milestones for the design and implementation of ASTRID

Several phases are foreseen for the design and implementation of ASTRID [1] as it is illustrated in Figure 4.

Figure 4: Main phases of ASTRID projects development.

A preliminary analysis of the possible options for ASTRID was submitted in mid-2010.

A preparatory stage was organised in view of the design phase, which will finish in early 2011, in order to finalise the work programme and organisation.

The first major project milestone has been set for late 2012 at which time the CEA will have to provide the French Government with a preliminary cost estimate and schedule for the subsequent phases, after decision to continue. Innovation will be encouraged during the first part of the preliminary design phase in order to achieve a project design that is consistent with the objectives of Generation IV reactors. Dialogue will be instigated with the French Nuclear Safety Authority (ASN) during this period, which will result in a “safety orientation report”.

Most options will have been finalised by the end of 2012. The second part of the preliminary design phase – lasting until late 2014 – will consolidate the first part, providing more information and greater consistency. The documents required during the preliminary design will be drafted. The safety options report will be written and submitted to ASN during this second phase.

The detailed design phase is planned from 2015 to 2017. The decision to build ASTRID will be taken by 2017. The start of the operation is then expected in 2022–23.

The SFR technology roadmap with the main domains to be covered with respect to the ASTRID main phases is represented in Figure 5.




49

Figure 5: The SFR technology roadmap with respect to the ASTRID programme.

7.2 LFR technology roadmap

7.2.1 Roadmaps of MYRRHA and ALFRED LFR

The plan schedule for completing the design, for supporting the pre-licensing and for starting with the construction of MYRRHA research reactor is defined. The details are available at SCK•CEN, which is the designer.

The first phase will end in 2014, when the engineering design and the support test facilities will be available (see Figure 6). A list of 10 concerned experimental facilities, which are planned to be constructed and/or operated for MYRRHA project, is reported in Annex 1 of D3.2 [4]. Among the others, the following have to be mentioned: CRAFT operative at SCK•CEN by 2012, a coolant chemistry facility for LBE scheduled for 2012, COMPLOT (SCK•CEN, 2012), E-SCAPE, RHAPTER (loop facility aiming to test fuel manipulator and moving components), INTRIGE (pool type facility), etc. In this connection, it is observed that E-SCAPE is a new test facility for component testing (i.e. similar to CIRCE).

The draft roadmap for ALFRED LFR reactor is defined in cooperation with ANSALDO Nucleare, which is the main designer. The plan is preliminary and more details are still needed. The main efforts planned in short term are:

1. to identify and to qualify the suitable coating for structural materials at higher temperature in HLM; 2. to provide a large pool infrastructure for SGTR and integral testing.

Based on the available information about the time planning of MYRRHA and LFR systems developments, a roadmap related to the experimental infrastructures is proposed in Figure 6. The figure includes existing and under design test facilities in connection with the topics of investigation.




50

Figure 6: Roadmap for experimental infrastructure needed by MYRRHA and LFR projects.

7.3 GFR technology roadmap

The short and midterm activities are concentrated on:

• the preparation of the ALLEGRO demonstrator construction with its first MOX core, • the development of advanced GFR fuel for the ALLEGRO second core and further larger reactors

(prototype and commercial reactors). First advanced GFR bundles should be ready to be irradiated in the ALLEGRO MOX core one year after reactor start-up.

Figure 7 illustrates how the main R&D domains should be scheduled in view of the planning constraints for ALLEGRO, with the assumption of:

• a decision for detailed studies and construction in 2012, • beginning of construction in 2018, • reactor start-up in 2025.

M

YR

RH

A r

equi

rem

ents

LF

R r

equi

rem

ents

Scaled down testing of SG and DHR (THEADESA, …C)

-

Testing corrosion of PS materials (TELEMATA, …C)

1

Testing protective coatings (LECORA, …C)

8

SGTR issue (small scale) (LIFUS5A, KALLSTARA)

-

SGTR issue (ATHENA-IB)

2

Sub-assembly TH (THEADESA, NACIEA, HELENAB, …)

-

SGTR issue (ATHENA-IB)

2

Pool TH of HLM (E-SCAPEB)

4 Testing corrosion of PS materials (CRAFTB, …C)

8

Integral tests (CIRCEA, …C)

3

MCP testing (CIRCEA)

7

Full scale component qualification (ATHENA-IIC, …C)

5

HLM pool TH (CIRCEA, E-SCAPEB, ...C)

4

Full scale MCP qualification (RomaniaC, …C)

7

Full scale component qualification (INTRIGEB, RHAPTERB, CIRCEA, ... C)

5

Severe accident investigations (…C)

6

MCP testing (CIRCEA)

7

Integral tests (ATHENA-IIC, …C)

3

Coolant chemistry control (HELIOS IIIB, SLEEVEA, …C)

9

Coolant chemistry control (…C)

9

Sub-assembly TH (THEADESA, NACIEA, HELENAB, …C)

-

Fuel coolant interaction (…C)

10

Fuel coolant interaction (…C) 10

Seismic testing (…C)

11

Coolant impurities transport (…C)

9

A Existing B Planned and financed C Needed infrastructure to be funded




51

Figure 7: Schedule of the main technology domains to be covered in view of ALLEGRO constraints.

The materials development appears as a high project priority, especially because the SiC cladding concept design and the fabrication of first pins for potential irradiation in ALLEGRO in 2026 is very constraining. It is a continuing process anyway, because a full ceramic core could be fabricated around 2035 (the first ceramic core of ALLEGRO). In line with the materials characterizations, the severe accidents behaviour analytical tests on advanced GFR pins are also of a high priority.

In terms of thermal-hydraulics, the provisional safety dossier due before construction will have to demonstrate the status of the MOX core and system calculation tools qualification with quantified uncertainties by 2016.

The components technology options must be cleared during the detailed design process (2012–2018), as regards in particular:

• capability of DHR blowers to cover a wide pressure range, • reliability of valves/check valves, • thermal barriers adapted to GFR accidental conditions, • demonstrated fuel handling capability with articulated arm, • core thermal instrumentation using optical fibbers and pyrometers, • helium quality management solutions adapted to both MOX and ceramic cores environments.

Operation/accident tests cover in particular the capability to switch the core cooling from the main loops to the DHR loops and their capability to operate in natural circulation in specified conditions.

For such tests, the strategy would consist in relying at first on existing facilities and to prepare in parallel the new needed investments (in particular to cover DHR natural circulation) in order to fulfil the licensing procedure requirements. Therefore, tests on existing facilities could cover the period 2012 to 2017, and later tests on new or upgraded facilities (such as qualification of operational/accidental procedures) could cover the period 2015 to 2020 with some two years overlap between the two types of tests.




52

For the case of severe accidents, the strategy consists also in relying on existing facilities for the short and midterm (analytical tests). Decisions about investments will need to be made in the middle of the detailed design phase of ALLEGRO (2016), when some severe accidents scenarios assessments will be completed.

7.4 Roadmap for instrumentation and diagnostics development

The roadmap on instrumentation for ESNII reactors is sketched in Figure 3. It is oriented close to the roadmaps developed for SFR (WP2), LFR (WP3) and GFR (WP4). In a first step “classic” and “innovative” instrumentation are handled separately. “Classic” systems already qualified in fast power plant are considered to be robust against the environmental conditions of fast reactors. As discussed in D5.1 [8], experiences gained from situations beyond normal operation are very important to optimize the instrumentation.

Innovative systems, which have been used so far only at lower temperatures, have to pass an out-of-pile qualification programme, which is a prerequisite for in-pile qualification. This also includes the combination of different sensors using adaptive monitoring techniques. When the concept is fixed qualification of the instrumentation including diverse systems has to demonstrate the robustness of the instrumentation system. In the last step, in-pile test have to provide the necessary data for the required parameters as it is discussed in Ref. [9] (D5.2).

Based on the needs of the individual systems the roadmap can be specified as follows:

Urgent (before 2017):

1. Core supervision & reactor protection system (temperatures, mass flow rate, neutronics) has to be qualified with respect to safety to early detect deviations from normal operation using modern techniques as discussed above. To extend the available set of instrumentations, innovative sensors have to qualified for nuclear applications

2. Fast localisation of pin failures (req. pool experiments) is an issue, which can avoid sub-assembly damage

3. Chemistry control of the fluid (liquid metal or gas) to minimize corrosion

4. Qualification of passive devices such as control or shut down systems in order to avoid power excursion (first out-of-pile, finally in-pile with fast reactor neutronics.

Long term: Extended research a qualification is needed for the flowing issues:

1. Ageing of advanced in-core instrumentation, especially the isolation of electric cables, can only be performed in fast systems.

2. Vibration detection on S/A and on core level to early identify possible damage, preferentially in a prototypic fast rector using a dummy S/A with a vapour generator as tested in the KNK-2 reactor (no gas allowed!).

3. Due to the PSA analyses, the diversity level has to be increased so that at least each sensor has an alternative (diverse) device accessible.

4. Leakage of molten metal to the fluid of the power conversion system should be practically eliminated by design. If this is not possible, design, detection, and localization has to be optimized to minimize the interaction.




53

It is quite clear that the first step is to demonstrate the high quality requirements of a Gen IV System, whose safety features are comparable or better to that of a Gen II or Gen II+ reactor. If this is proved in the prototypes, the efficiency as well as the transmutation capability can be optimized.




54

8 Conclusions Fast neutron reactors have a large potential as sustainable energy source. In particular, nuclear reactor systems with a closed fuel cycle and potential for minor actinide burning may allow minimization of volume and heat load of high level waste and provide improved use of natural resources compared to only 1% energy recovery in the current once-through fuel cycle with thermal reactors.

Among the fast reactor systems, the Sodium-cooled Fast Reactor has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype and commercial size reactors since the 1940s. This concept is currently considered as the reference within the European strategy. Innovations are needed to further enhance safety, reduce capital cost and improve efficiency reliability and operability, making the Generation IV SFR an attractive option for electricity production.

The commercial deployment of future SFR ―and, in general, of Gen IV fast reactors― is not expected before 2040. ASTRID, an SFR reactor demonstrator of 600 MWe power, is currently under development in France. Most conceptual options of this reactor will be finalised by the end of 2012 and will be followed by the second phase (2012–2014) consolidating the preliminary conceptual design. The detailed design phase is planed from 2015 to 2017 and the decision to build the ASTRID prototype will be taken by 2017. At the present time, the start of the operation is expected by 2022–23.

In the following 4–6 years, the eleven research platforms planned in France, Germany, the Czech Republic, Italy, and Latvia (see section 2.2) will contribute to the development of the SFR technology according to the identified needs in support of the ASTRID programme.

Regarding the two alternative technologies, LFR and GFR, at the present time the projects of fast reactor prototypes MYRRHA, ALFRED and ALLEGRO are in the pre-conceptual phase. At the same time, the roadmaps for development of these both technologies define needs and necessary experimental facilities which should be constructed in the following years (see sections 3.3 and 4.2).

The first phase of the LFR technology development will end in 2014 when the engineering design of MYRRHA and the support test facilities will be available. Construction of the facility and assembly of the components is foreseen from 2015 to 2019. Three years (2020–2022) are foreseen for the full commissioning of MYRRHA.

The development of ALFRED prototype LFR reactor is supposed during 2010–2025 period considering the design phase to start in 2014 and the construction after 2017. The start of operation is then estimated for 2025. The LFR roadmap set up the needs and priorities for the technology development in support of the two projects MYRRHA and ALFRED and drawn the plan of necessary facilities to be built in the following years for their R&D support years (see sections 3.3).

Concerning the ALLEGRO development, the preliminary schedule of the project supposes the decision for detailed studies and construction in 2012, beginning of construction in 2018, and the reactor start-up by 2025. The GFR technology roadmap identified necessary experimental facilities in support of the GFR technology development in consideration of the ALLEGRO programme needs (see section 4.2).




55

Instrumentation is a key issue in all nuclear systems and especially for Gen IV systems. Adequacy, reliability, and robustness are essential for successful operation. Instrumentation for liquid metal fast reactor originates from developments 30 years ago and from advanced sensors used in non-nuclear liquid metal applications. Both roots have to be combined to develop innovative high qualitative sensors and to qualify them for the harsh environments. High priority is identified for sensor systems necessary for safe operation of the prototypes. A long term aspect is added to further develop and mature these sensors for commercial NPP. The ISI&R is oriented to situations in commercial NPP and extended by the higher needs of fast reactors such as narrower design, higher temperatures, opaque fluid, and so on.

Finally, a set of qualified instrumentation and techniques including at least one diverse system for each issue will be available for Gen IV systems. The qualification process is part of the long term evolution based on prototypes such as ASTRID or MYRRHA.




56

Conceptual design-1Conceptual Design-2Basic designDetailed design and constructionFacilitiesFeasibility Report on MA partitioningPosition Report on MA partitioning and transmutationCore manufacturing workshop (AFC)MA bearing fuels fabrication facilityMaterialsCooalntsTH & SafetyComponents qualifiactionInstrumentationEducation & Training

Coolant chemistry control

Testing corrosion of PS materialFuel coolant interactionSub-assembly THPool TH of HLMIntegral testsSGTR issueFull scale component qualificationCoolant impurities transportMCP TestingSevere accident investigationsSeismic testingDetail designConstructionOperation

Testing corrosion of PS materialTesting protective coatings

Scaled down testing of SG and DHRSGTR issue (small scale)SGTR issueMCP TestingSub-assembly THIntegral testsFull scale component qualificationFull scale MCP qualificationPool TH of HLMCoolant chemistry controlFuel coolant interactionConceptual design, decision pointDesign phase and constructionStart of operation

Decision for detailed studies and constructionThermal-hydraulicsComponentsOperation/accidents tests on existing facilitiesDesign Process (components technology options)Operation/accidents tests on upgrade facilitiesBeginning of constructionMaterials: esp. SiC claddingOperation/accidentsReactor Start-UP

2023 2024 2025

ALF

RED

2019 2020 2021 20222015 2016 2017 20182011 2012 2013 2014

2023 2024 20252019 2020 2021 20222016 2017 20182014 20152013M

YRR

HA

ALL

EGR

OA

STR

ID2011

SFR

LFR

GFR

2012

2010

Figure 8: Roadmaps for European SFR, LFR and GFR projects.




57

9 References

[1] C. Latgé et al., “Final WP2 report on future needs and infrastructure road map supporting SFR system development”, FP7 ADRIANA project deliverable D2.2 (2011).

[2] L. Vála et al., “Matrix exercise: Map of all research facilities (existing or projects)”, FP7 ADRIANA project deliverable D8.3 (2011).

[3] P. Agostini et al., “Preliminary report describing present situation and future needs addressed by WP3”, FP7 ADRIANA project deliverable D3.1 (2011).

[4] P. Agostini et al., “Final WP3 report on future needs and infrastructure road map supporting LFR system development”, FP7 ADRIANA project deliverable D3.2 (2011).

[5] C. Poette et al., “Final WP4 report on future needs and infrastructure road map supporting GFR system development”, FP7 ADRIANA project deliverable D4.2 (2011).

[6] P. Le Coz, J.-F. Sauvage, and J.-P. Serpantie: “Sodium-cooled Fast Reactors: The ASTRID Plant Project”, ICAPP’11 Conference, Nice, France, 2011.

[7] W. Hering et al., “Report on the present situation of instrumentation, diagnostics, experimental devices for experiments and SFR, LFR, and GFR”, FP7 ADRIANA project deliverable D5.1 (2011).

[8] L. Vála et al., “Evaluation of existing research infrastructure for long-term vision of sustainable energy”, FP7 ADRIANA project deliverable D8.2 (2011).

[9] W. Hering et al., “Report on infrastructure road map and future needs for innovative systems development focused on instrumentation, diagnostics and experimental devices”, FP7 ADRIANA project deliverable D5.2 (2011).

[10] L. Vermeeren et al., “Final WP6 report on material testing reactors, transient testing reactors and hot laboratories: infrastructure needs in support of SFR, LFR and GFR development”, FP7 ADRIANA project deliverable D6.2 (2011).

[11] L. Vermeeren et al., “Preliminary report describing present situation and future needs addressed by WP6”, FP7 ADRIANA project deliverable D6.1 (2011).

[12] L. Vála et al., “Mapping of existing research infrastructures and list of research infrastructure projects”, FP7 ADRIANA project deliverable D8.1 (2011).

[13] V. Juříček et al., “Final WP7 report on future needs and infrastructure road map supporting Zero Power Reactors development”, FP7 ADRIANA project deliverable D7.2 (2011).

[14] F.-J. Hambsch (Scientific Coordinator of ERINDA) – private communication.

[15] “Research and Test Facilities Required in Nuclear Science and Technology”, NEA, OECD 2009, NEA 6293.

[16] “Funding opportunities and legal status options for the future European Sustainable Nuclear Fission Industrial Initiative of the Strategic Energy Technology Plan”, Deloitte, February 2010.

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