Review of Euratom projects on design, safety assessment, R

EPJ Nuclear Sci. Technol. 6, 36 (2020)© K. Mikityuk et al., published by EDP Sciences, 2020https://doi.org/10.1051/epjn/2019007

NuclearSciences& Technologies

Available online at:https://www.epj-n.org

REVIEW ARTICLE

Review of Euratom projects on design, safety assessment,R&D and licensing for ESNII/Gen-IV fast neutron systemsKonstantin Mikityuk1,*, Luca Ammirabile2, Massimo Forni3, Jacek Jagielski4, Nathalie Girault5,Akos Horvath6, Jan-Leen Kloosterman7, Mariano Tarantino8, and Alfredo Vasile9

1 PSI, Forschungsstrasse 111, 5232 Villigen PSI, Switzerland2 JRC, Westerduinweg 3, 1755 LE Petten, The Netherlands3 ENEA, Via Martiri di Monte Sole, 4, 40129 Bologna, Italy4 NCBJ, A. Soltana 7, 05-400 Otwock/Swierk, Poland5 IRSN, 13115 St-Paul-lez-Durance, France6 MTA EK, Konkoly Thege M. �ut 29-33, 1121 Budapest, Hungary7 TU DELFT, Mekelweg 15, 2629 JB Delft, The Netherlands8 ENEA, FSN-ING, C.R. Brasimone, 40032 Camugnano, Italy9 CEA, 13115 St-Paul-lez-Durance, France

* e-mail: k

This is an O

Received: 12 March 2019 / Accepted: 4 June 2019

Abstract. Nine Euratom projects started since late 2011 in support of the infrastructure and R&D of the sevenfast reactor systems are briefly presented in the paper in terms of key objectives, results and recommendations.

1 Introduction

In November 2010 Sustainable Nuclear Energy TechnologyPlatform (SNETP) set up a Task Force comprisingresearch organisations and industrial partners to developthe European Sustainable Nuclear Industrial Initiative(ESNII) addressing the need for demonstration of Genera-tion-IV Fast Neutron Reactor technologies, together withthe supporting research infrastructures, fuel facilities andresearch and development (R&D) work.

SNETP has prioritised the different Generation-IVsystems and is proposing to develop the following projects:the sodium-cooled fast neutron reactor technologyASTRID as the reference solution; the lead-cooled fastreactor ALFRED supported by a lead-bismuth irradiationfacility project MYRRHA as a first alternative; the gas-cooled fast reactor ALLEGRO as a second alternative. TheMolten Salt Fast Reactor (MSFR) is considered as a veryattractive long-term option.

The EU framework programs have supported the R&Dactivities on these five systems as well as on two otherGeneration-IV technologies: European Sodium FastReactor (ESFR) and Swedish Advanced Lead Reactor(SEALER). All seven fast neutron systems are presentedin Figure 1.

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The paper briefly presents in terms of key objectives,results and recommendations nine Euratom projectsstarted since late 2011 in support of the infrastructureand R&D of the seven fast reactor systems presented above(see Fig. 1). Table 1 presents the list of the projectacronyms, participants and coordinators. Figure 2 presentsdomains and categories of advanced systems, while Table 2gives more details about the R&D areas. Finally, Figure 3presents the budgets and time spans of the presentedprojects.

2 SARGEN_IV: Proposal for a harmonizedEuropean methodology for the safetyassessment of innovative reactors with fastneutron spectrum planned to be builtin Europe

2.1 Key objectives

The safety of innovative reactors needs to be addressed in acomprehensive and robust manner while demonstrating alevel of safety acceptable for the general public. Having aEuropean consensus on themethodology and safety criteriathat will be used to assess innovative reactors becomes ofprime importance with an impact on any further designactivities.

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Fig. 1. Seven fast neutron systems supported by the considered EU project: ASTRID (a); ALFRED (b); MYRRHA (c);ALLEGRO (d); ESFR (e); SEALER (f); MSFR (g).

Table 1. Participants and coordinators of the considered EU projects.

Fig. 2. Domains and advanced systems of interests of theconsidered EU project.

2 K. Mikityuk et al.: EPJ Nuclear Sci. Technol. 6, 36 (2020)

With the goal of preparing the future assessment ofthese advanced reactor concepts, the European projectSARGEN_IV gathered safety experts from 22 partnersfrom 12EU Member States: recognized European Techni-cal Safety Organizations (TSOs), the Joint ResearchCentre of the EC, Designers and Vendors as well as fromResearch Institutes and Universities in order to:
– identify the critical safety features of the selected Genera-tion-IV concepts, relying on the outcomes from existingprojects from the 7th Framework Programme (FP7);
–
develop and provide a tentative commonly agreedmethodology for the safety assessment, relying on theoutcomes of the investigations carried out within
international organizations (such as IAEA, WENRA,OECD/NEA), on national practices presently in use andon practices proposed within other European FrameworkPrograms projects;

–
identify open issues in the safety area, mainly addressingand focusing on assessment relevant ones, detect andunderline new fields for R&D in the safety area(addressing methodological, theoretical and experimen-tal issues, as well) in order to provide a roadmap andpreliminary deployment plan for the fast reactor safety-related R&D.
The project partners were convinced that fostering theharmonization of the various European safety approacheswould have been very beneficial and would have stream-lined Euratom contribution to Generation-IV Internation-al Forum in the safety field. It was also meant to improverelationship between safety assessment needs and researchprogrammes efficiency in the development of new concepts.

A particular attention was addressed to take intoaccount the lessons learned from the Fukushima-Daiichinuclear accident that will impact significantly theresearch and development needed for demonstration ofGeneration-IV reactor safety.

2.2 Key results2.2.1 WP2: identification of the major safety features

In the project, a review on the safety issues was performedfor each ESNII concept: SFR, LFR, GFR and MYRRHA

Table 2. R&D areas of the considered EU project.

TH & CFD Neutronics Fuel Seismic Multiphysics

SILER xALLIANCE x xJASMIN x x xESNII Plus x x x x xVINCO xSESAME xSAMOFAR x x xESFR-SMART x x x x

Focus of SARGEN_IV project is safety assessment

Fig. 3. Budget (a) and time span (b) of the considered EU project.

K. Mikityuk et al.: EPJ Nuclear Sci. Technol. 6, 36 (2020) 3

FASTEF. A list of the initiating events was also identifiedand categorised according to their occurrence frequency.

A conclusive deliverable [1] gathered themain results foreach of the three concepts and a focus was performed toidentify phenomena able to affectmore thanone concept, i.e.
– for the coolant: sensitivity to impurities, coolant activity,retention of fissions products, toxicity, opacity;
–
for the structural materials: corrosion, erosion, irradia-tion behaviour;
–
issues in relation with fast reactors: sensitivity toblockage, power density, core compaction, reactivityvoid effects, handling hazards, failure of the coresupporting structures;
–
management of the three safety functions (reactivitycontrol, decay heat removal, containment);
–
capability to cool the core by natural circulation; – sensitivity to external events (flooding, earthquake); – considerations on the Fukushima-Daiichi TEPCO events(extreme flooding, extreme earthquake, total loss ofelectric supply, accident management);
–
categorisation of initiating event organised by challenges:challenge to clad integrity, challenge to reactor bound-ary, containment challenge.
This work gave a useful guidance for the identificationand the prioritisation of the R&D needs respective to theidentified safety issues. In particular it was pointed out that

efforts have to be performed to define the severe accidentfor each concept and to develop requirements for thecontainment in order to practically eliminate large andearly releases.

2.2.2 Develop and provide a tentative commonly agreedmethodology for the safety assessment

In the scope of the development and the licensing of theabove mentioned ESNII prototypes in Europe, it appearedcrucial to develop a tentative commonly agreed assessmentmethodology able to be applied to each of the four abovementioned concepts and based on the safety issuesidentified.

Firstly, it performed a review of the safety methodolo-gies proposed by international organizations and thoseissued from national practices and European consortia.This included:
– INPRO methodology proposed by IAEA and ISAMproposed by the GIF;
–
experience feedback for safety assessment from nationalTSOs approaches (from Finland, France, Belgium,Spain, Germany);
–
safety approach proposed for European projects relatedto gas cooled, lead cooled and sodium cooled fast reactors;
–
safety approach proposed by international organisations(IAEA, WENRA, NEA/MDEP).


2.3 Recommendations for the future

On the basis of the reviews mentioned above that led tonumerous recommendations, the SARGEN_IV consor-tium prepared a proposal [2] for the safety assessmentpractices targeting the Generation-IV prototypes to bebuilt in Europe.

Some of the most important recommendations are asfollows:
– the safety assessment should cover the whole nuclearplant (reactor, fresh and spent fuel storage);
–
the entire life on the plant (from commissioning todecommissioning) should be addressed;
–
safety assessment should integrate the security/safe-guards aspects;
–
the consequences of chemical releases have to be takeninto account in the design;
–
the defence-in-depth (DiD) principle remains a funda-mental principle for the safety of innovative reactors andan important topic is to define accurately the level 4 ofDiD for each concept;
–
accident sequences that could lead to large or earlyreleases have to be practically eliminated.
3 SILER: Seismic-Initiated Events RiskMitigation in Lead-cooled Reactors

SILER is a collaborative project, partially funded by theEuropean Commission in the 7th Framework Programme,aimed at studying the risk associated with seismic-initiatedevents in Generation-IV Heavy Liquid Metal reactors, anddeveloping adequate protection measures. The attention ofSILER is focused on the evaluation of the effects ofearthquakes, with particular regards to beyond-designseismic events, and to the identification of mitigationstrategies, acting both on structures and componentsdesign. Special efforts are devoted to the development ofseismic isolation devices and related interface components.

Two reference designs, at the state of developmentavailable at the beginning of the project and coming fromthe 6th Framework Programme, have been considered:ELSY (European Lead Fast Reactor) for the Lead FastReactors (LFR), and MYRRHA (Multi-purpose hYbridResearch Reactor for High-tech Applications) for theAccelerator-Driven Systems (ADS).

3.1 Key objectives

One of the main goals of SILER was the development andexperimental qualification of seismic isolators for lead-cooled reactors (but applicable to any other nuclear plant).

3.2 Key results

Two device typologies have been considered: High Damp-ing Rubber Bearings (HDRBs) and Lead Rubber Bearings(LRBs). Both isolators have been designed (for ELSY andMYRRHA, respectively), manufactured and tested indifferent sizes, even to the full scale, which results to be

greater than one meter, due to the huge mass of the reactorbuildings. In particular, a prototype has been subjected tothree-directional dynamic tests (at the Department ofStructural Engineering of the San Diego University) underthe real service loads up to failure, which occurred wellbeyond the design conditions.

The adoption of base isolation provides a greatreduction of the acceleration and inertial forces in thestructure, providing very important benefits to thecomponents and the structure itself, but introducessignificant relative displacements between the isolatedand conventionally founded parts of the plant. Thus, aseismic gap of suitable width shall surround the entireisolated “island”. Of course, it shall be adequately protectedfrom bad weather (included floods) and other possibledamages, and kept free during the whole life of thestructure, in order to allow for relative movements in caseof earthquake. Moreover, all the service networks andpipelines crossing the seismic gap shall be provided withsuitable expansion joints. In SILER, both devices havebeen developed and successfully tested in full scale and inreal operational conditions, even beyond the design limit(see Figs. 4–6). It is worth noting that, due the severeseismic condition assumed in the design of nuclear plants,the relative displacement can reach 0.7–0.8m in beyond-design situations.

In SILER, several critical components of ELSY andMYRRHA (like vessel, pumps, proton beam, etc.) havebeen numerically modelled and carefully analysed undersevere seismic conditions, taking also into account theeffects of the sloshing of the liquid lead and the soil-structure interaction.

Particular attention has been devoted to the cost-benefit analysis related to the adoption of seismic isolation,which resulted to be positive. Moreover, according to theindication of EC, the main results of the project have beendisseminated through the organization of seminars,courses, workshops and the implementation of a web site(http://www.siler.eu).


In particular, guidelines for design, manufacturing,qualification, installation and maintenance of seismicisolators for nuclear plants have been delivered. Thisdocument is particularly important, due to the lack ofinternational rules regarding the seismic isolation ofnuclear plants (at the time of the project at least).

More information about the SILER Project mainresults can be founded in references [3,4].

4 ALLIANCE: Preparation of ALLEGRO �implementing advanced nuclear fuel cyclein central Europe

Gas cooled fast reactors (GFR) represent one of the threeEuropean candidate fast reactor types, the two other beingsodium cooled fast reactor (SFR) and lead cooled fastreactor (LFR). Technically, GFR is a realistic andpromising complementary option thanks to its specific

http://www.siler.eu

Fig. 6. Three-directional dynamic test performed at SRMD on afull-scale (1350mm diameter) HDRB. After partial damageoccurred close to 300% shear strain (almost three times the designvalue), the isolator was successfully subjected to a full cycle underthe design load at the design conditions.

Fig. 4. Sketch of the pipeline connecting the seismically isolated reactor building of ELSY and the turbine, provided with two flexiblejoints to adjust the relative displacements.

Fig. 5. Full scale pipeline expansion joint during seismic tests atthe ELSA laboratory of the JRC of Ispra.


advantages connected with high temperatures. The GFRconcept was mainly based on studies performed in Francein the late 1990s and was further developed within the EU5th and 6th Framework Programmes, respectively. It alsoincluded the development and safety assessment of a smallexperimental plant called at the time ETDR (Experimen-tal Technology Demonstration Reactor). This plant wasregarded as a necessary stepping-stone to a full-sized GFR

in order to test the high-temperature fuel required by thelatter. The concept was further analysed and refined by theEU FP7 GoFastR project: the ETDR has been renamedALLEGRO (see Fig. 8) and a number of design changeswere introduced, e.g. the power was raised from the original50MWth to 75MWth. ALLEGROwould function not onlyas a demonstration reactor hosting GFR technologicalexperiments, but also as a test pad of using the hightemperature coolant of the reactor in a heat exchanger forgenerating process heat for industrial applications and aresearch facility which, thanks to the fast neutronspectrum, makes it attractive for fuel and materialdevelopment and testing of some special devices or otherresearch works.

The three respective nuclear research institutes of theCentral European region (�UJV, Řež, Czech Republic,MTA EK, Budapest, Hungary, and V�UJE, a.s., Trnava,Slovakia) agreed in 2010 to start a joint project aiming atthe preparation of the basic documents in order to form the

Fig. 8. Schematic drawing of the ALLEGRO Reactor (courtesyof Petr Darilek, VUJE).

Fig. 7. ASTEC-Na calculation scheme and modelling capabilities.


basis for a later decision on the construction and operationof the ALLEGRO gas cooled fast reactor in one of thecountries. CEA, France, supports this effort by variousmeans, especially by transferring the documents of theearlier design efforts (under appropriate Non-DisclosureAgreements) to the project participants. NCBJ, Swierk,Poland, joined the project in 2012, i.e. ALLEGRO issupported in all the four Visegrad-4 (V4) countries. Theproject ALLIANCE has been launched in 2012 by the ninemember organizations (see Tab. 1).

4.1 Key objective

The aim of the project ALLIANCE was to continue theelaboration of basic documents needed for high leveldecisions and licencing of ALLEGRO. The ALLIANCEproject [5–7] focused on the preparatory phase fordeveloping the ALLEGRO gas cooled fast reactordemonstrator. The main nuclear parameters (like powerdensity, burnup etc.) would be similar to those of theplanned 2400 MWth power GFR. The core built up fromthe initial fuel type will be replaced by a core of ceramic fuelfor the second half of ALLEGRO operation. Safety analysisperformed within the previous EU GoFastR projectcovered the consequences of most initiating events andmost of the ALLEGRO relevant issues were analysed.Safety principles of the ALLEGRO reactor will be based onthe WENRA requirements and the study of GIF [8], addedto the actual national safety rules of the hostingcountry. Moreover, in formulating siting requirements


and requirements concerning the design to reduce theimpact of external hazards, the results of the Europeanstress tests following the Fukushima events were applied.Nevertheless the current design of ALLEGRO does notfully satisfy these requirements. One of the main reasons isthat the safety margin of the stainless steel cladded mixedoxide (MOX) fuel chosen for the initial ALLEGRO core of75MWth power is rather low and cannot provide thenecessary protection against core melting after a Fukush-ima-type accident (though the margin is acceptably largeconcerning Design Basis Accidents, i.e. accidents whichmay occur with a very low but not negligible probability).

4.2 Key results

A new strategy for developing the ALLEGRO reactor wasprepared, and accepted by the Partners in 2015. The maincomponents of this strategy are: (a) to reduce ALLEGROpower from 75MWth to 10MWth and to find the optimumcore configuration (switch from MOX to UO2); (b) tooptimize nitrogen injection (launch time, duration) and thebackup pressure in guard containment; (c) to increase mainblowers inertia to avoid short term peak temperature forthe loss of coolant accident+blackout case and/or todevelop a design with a gas turbine in the secondary sidecoupled to the primary blowers (this is the solution alsoadvised for GFR).

A new systematic Roadmap was prepared to cover alldesign, safety and experimental aspects of ALLEGROdevelopment.

The ALLEGRO consortium is represented by V4G4Centre for Excellence, a legal entity registered in Slovakia.The main goal of V4G4 is to establish R&D facilities toinvestigate fuel development issues, helium technologyrelated problems, issues related to structural materials andto construct a non-nuclear 1:1 mock-up of ALLEGRO.

TheRealisationPhaseof the “ALLEGROProject”will bestartedwhenever theobjectivesof thePreparatoryPhasearereached, approximately in 2025. The realisation phase willinclude the preparation of the basic design, licensing (sitepermit, construction license, etc.), construction, operationand decommissioning of the ALLEGRO reactor.

As ALLEGRO will be a result of a joint effort on theregional (or even European) level, an internationalconsortium should be formed to finance the entire project.The desired and potentially possible governance structureapplicable in the Realisation Phase was discussed withinthe ALLEGRO project almost from the very beginning. Itwas found that the existing EU structures (e.g. “ERIC �European Research Infrastructure Consortium”) are notapplicable as they are appropriate only for infrastructuresused for basic research and they practically exclude thejoint financing by governments and the industry. In case ofnuclear development infrastructures the contribution fromboth sides is absolutely needed. The different governancemodels were discussed in detail in the project deliverables.


The Design and Safety Roadmap was elaborated whichconsists of about 80 tasks in order to elaborate a new

conceptual design with satisfactory safety features by 2025.A pre-conceptual design will be prepared and discussed onthe European level by 2020. The Roadmap clearly fixes theachievements needed for the pre-conceptual design and theconceptual design by tasks. Leading and participatingmember organisations are declared for each task. Themanpower needed and eventual investment costs are alsoestimated per task. The first version of the Research-Development-Qualification Roadmap is also prepared. Itconsists of those experimental tasks which are necessary tocomplete in order to ensure a sound basis for the design.

One of the main challenges of the ALLEGRO design isassociated with final resolution of the emergency decayheat removal from the core. This problem is a key issue forfeasibility and safety acceptance of the GFR. To continuewith development of the ALLEGRO GFR demonstratordesign, complex set of tools is necessary, allowing reliablesimulation of both operation and safety relevant events, upto severe accidents.

5 JASMIN: Joint Advanced Severe accidentsModelling and Integration for Na-cooled fastneutron reactors

The project was launched in 2011 in support of both theESNII roadmap and the Deployment Strategy of SNETPfor the enhancement of Sodium-cooled Fast neutronReactors (SFR) safety through the development of newcapabilities to simulate innovative reactor designs [9]. Theproject was focussed on the primary phase of SFR coredisruptive accidents, considered as the overture to largerscale core destruction. However, the code integratedfeatures, which represents a good opportunity for simulat-ing in a single code what is generally simulated in separatecodes, were also considered through the in-containmentsource term modelling.

5.1 Key objectives

The project aimed at enhancing the current capability ofanalysis of severe accidents in SFRs by developing a newEuropean simulation code, ASTEC-Na from the existingASTEC platform developed by IRSN and GRS for LWRs.It was motivated by the need for new simulation tools withupdated models, extended modelling scope and flexibleplatforms in replacement of the current available codes forSFR safety studies developed in the 1980s.

Then, it was intended to provide ASTEC-Na with:
– improved physical models (accounting for recent LWRand SFR research);
–
a modern architecture and a high flexibility to ease itscoupling with other tools and the accounting forinnovative reactor designs;
–
extended capabilities to evaluate the consequences ofunprotected accidents on materials relocation and fissionproducts and aerosols behaviour, once released.
Most important activities consisted in the developmentof new models and in their verification upon experimentaldata and through code benchmarks.


5.2 Key results5.2.1 ASTEC-Na model development

The three ASTEC-Na modules that focussed on themodelling efforts were CESAR, ICARE-SFR and CPA*.The final in-vessel and ex-vessel modelling capabilitieslisting the models that were developed are displayed inFigure 7. The ICARE module development particularlybenefited from accurate fuel thermomechanical and fissiongas models issued from SCANAIR1 for describing the pinbehaviour during accidental transients which makes it verypromising for assessing future SFR designs [9]. A highlyflexible point-kinetics model was also implemented withthe possibility to use time-dependent reactivity coefficientprovided by neutron physics codes [10]. The in-contain-ment source term modelling in CPA* was focused on theNa-particle generation from pool fires and their chemicalageing. Other source term issue (like fission productscrubbing in Na pools, etc.) was left out. Late incorporationof the FEUMIX module, simulating sodium pool & spraycombustion, greatly enhance the code capabilities but stillsource term modelling in ASTEC-Na needs to be extendedto the missing models.

5.2.2 ASTEC-Na model verification and validation

The CESAR thermos-hydraulic module, where the modelsdeveloped most, pointed out good performances (i.e.boiling onset) for the single phase where the quality ofASTEC-Na results were found similar to what exhibited bymore mature codes. For two-phase thermal-hydraulics, thepressure drop calculated by the 5-equations model wasgenerally too large and some deviations were found in thecalculation of the two-phase front radial propagationinherent to the 2D axial-symmetric model. In ICARE,though the RIA model showed its capability to reproducethe dynamics of the physical phenomena (i.e. internalpressure built-up, gap closure, clad straining, etc.), somedeviations from data trends during PCMI2 (i.e. axial fuelexpansion, clad deformation) were observed that couldprevent from an adequate molten fuel pressurization andclad failure calculation. The mechanistically based ap-proach for fission gas simulation (requiring data notnecessary available within SFR conditions) prevent from aconclusive RIA model validation.

The point kinetics model was found reliable to calculatethe power evolution in a pool-type SFR during transientstill the flux shape is not excessively perturbed. The validityof the model up to hexcan failure that depend on thematerial relocation and thus on the transient might beovercome, thanks to the ability of ASTEC-Na to use time-dependent reactivity coefficients but will require perform-ing a lot of iterations (to have adequate coefficients for atime period, the state of the core during this time period hasto be known).

1 SCANAIR is a simulation tool developed in IRSN for reactivity-initiated accident (RIA) in LWRs.2 PCMI : Pellet-Clad-Mechanical-Interaction.

The verification of in-containment source term model-ling in CPA* was not fully conclusive as key phenomenaremained described by heavily parameterized models.However, the deviations from data trends, in airborneconcentration of aerosols and their chemical compositions,highlighted a need for further review and extension of theimplementedmodels. Code benchmarking could not help asASTEC-Na was at the forefront of in-containment sourceterm modelling.


The ASTEC-Na tool, though offering great opportunitieswas still far from being mature at the end of the project TheSWOT analysis performed in analysing the code weak-nesses and threats allowed to point out the priorities infuture development of the missing models and, beyondASTEC-Na, to make some key recommendations for anyforthcoming development and validation of a safetyanalytical tool:
– extend the validation of prototypic MOX fuel thermos-mechanical and fission gas models to the high tempera-ture domain covered in SFR transients;
–
perform further analytical work on in-containment andin-vessel fission product behaviour to alleviate thescarcity of experimental data in the open literature.
As for ASTEC-Na, the development of an interfacewith a fuel irradiation and a neutron physics codes tominimize as far as possible the user work was stronglyrecommended and the continuation of the sensitivitystudies on the RIA model key parameter warmly advised.

6 ESNII Plus: Preparing ESNII forHORIZON 2020

6.1 Key objective

The aim of this four-year cross-cutting project was todevelop a broad strategic approach to advanced fissionsystems in Europe in support of the European SustainableIndustrial Initiative (ESNII) within the SET-Plan. Theproject involved private and public stakeholders, includingindustry, research and academic communities (see Tab. 1).

6.2 Key results6.2.1 Organisation of ESNII to capitalise on opportunitiesin Horizon 2020 and beyond

Ways to coordinate the work of ESNII between EC and thenational R&D programmes were analysed. Central to thiscoordination is establishing the funding mechanisms thatcan be used to gainmaximum leverage for funding obtainedfrom the EC’s Framework Programmes and for theMember State programmes.

6.2.2 Future financial and legal models for ESNII

Three challenges were identified:
– funding ESNII and SNETP. This is of the order of k€per partner, obtained from member organisations


combined with Euratom grants (FP7 and Horizon2020);

–
funding the R&D carried out on ESNII systems. This is ofthe order of M€, and is obtained from Member Statenational programmesandEuratomgrants (Horizon2020);
–
funding design and construction of the ESNII demon-strators. This is of the order of G€ and must be obtainedfrom Member State national programmes.
6.2.3 Strategic Roadmapping

The aim of the taskwas to facilitate anddefine areas for jointprogramming between national actors, Member States andthe EU. This task hence served as a first benchmarkingexercise of joint proposals with variable common objectivesand partnerships for Horizon 2020EU programmes. Aworkshop was organised and the following topics wereidentified: MOX Fuel, Austenitic and Ferritic-MartensiticMaterials, In-core Instrumentation and RCC-MRX code.

6.2.4 Support to facilities development

Functional specifications of the R&D facilities related toSFR, LFR and GFR were identified with particularattention to the specificities and the unresolved issues.The availability and capabilities of irradiation infrastruc-ture in Europe were reviewed in order to support thematerial and fuel development.

6.2.5 Siting and licensing requirements for the newgeneration of fast reactors

The specific requirements for licensing Generation-IVreactors are currently not explicitly included in the existinglegal framework at the national level, even if there are plansor intentions to modify the legislations to improve thenuclear safety and to address the new reactor generationdevelopment. In order to survey the requirements andrecommendations that may be used in the process oflicensing Generation-IV systems, by capturing and inte-grating the international experience, an overview on theexisting standards and recommendations (WENRA, GIF,EUR, CORDEL, MDEP and IAEA documents), with theconsideration of Fukushima lessons learnt was performed.The conclusions drawn could be found in [11].

6.2.6 Prospective analysis of supply chain

Fast reactors, selected at European level as next generationNuclear Energy Systems, pose undeniable challenges froma technological point of view. In order to support theforeseen deployment strategy, a survey of the existingsupply chain has been thoroughly carried out in terms of itscapabilities and potentialities with respect to FastReactors needs. The identified challenges of the EU nuclearindustry with respect to Fast Reactors can be found in [12].

6.2.7 Potential of small modular fast reactors

Although the “economy of scale” was privileged as soon asnuclear was considered for civil applications, exceptionsare represented by installations in remote regions or by

specific technologies fitting in the small to medium powerrange. Opportunities offered by SMR based on fastreactors technologies were analysed, with a particularfocus on LFR and the EU context. Two main potentialapplications were identified: installations of SMRs havingpower in the order of 100 MWe for the compensation ofrenewables, or multi-units sites with a total power in therange 350–700 MWe for the replacement of fossil fuelpower plants and the supply of process heat to industrialclusters.

6.2.8 Potential of cogeneration fast reactors

The additional opportunity of fast reactors designed forcogeneration applications (i.e., production of electricityand process heat) is made possible by the elevatedtemperatures characterizing the primary circuit of suchreactors, compared to traditional LWRs. A state-of-the-art overview on the EU cogeneration market withemphasis on opportunities for fast reactors was comple-mented by technical recommendations and by a top downcost estimate for an LFR system in a cogenerationapplication.

6.2.9 Core physics

Benchmarking activities of neutronic codes used in Europeand recommendations for their application to the differentadvanced concepts were performed. Main safety parame-ters of the three EU demonstrators were calculated withthe main codes used in Europe. R&D needs to improve thecore safety were identified.

6.2.10 Fuel

MOX fuel properties catalogue was updated throughadditional measurements performed during the project onsamples previously irradiated in European reactors. Theeffect of burn-up on thermal conductivity was, for the firsttime, measured on MOX fuel for fast reactors with high Pucontent.

6.2.11 Seismic behaviour

The work focused on the modelling and analysis of thebehaviour of the demonstrators by implementing seismicisolators including experimental verifications proving theirefficiency in accidental conditions.

6.2.12 Instrumentation

Instrumentation and measurement techniques relevant tosafety and in service inspection and repair were developedrelated to fuel cladding failure detection, coolant chemis-try, thermal hydraulics characterisation and in-serviceinspection and repair.


–
ESNII shall continue organizing the EU R&D onsustainable nuclear energy systems. Coordination withnational member states programs needs to be encouraged.


–
The facilities for developing experimental programs shallbe preserved and stronger cooperation facilitated toavoid duplications and improve budget utilization.
–
A regulatory framework for Generation-IV reactors hasto be built by countries interested in Generation-IVsystems deployment to develop and maintain thecompetences for licensing process. The documents ofIAEA, WENRA, NEA and EUR may be used in theprocess of developing national Generation-IV systemslicensing requirements.
–
Concerning the industrial supply chain, further specifi-cations on Generation-IV specific components will beneeded to verify if there are suppliers for them.
–
Possible contribution of fast neutron systems toimplementation of SMRs in Europe should be furtherinvestigated.
–
In the core physics area, R&D must be pursued toimprove the safety.
–
Measurements of MOX fuel properties using existing andfuture irradiation experiments, in particular those havingan important impact on safety must be continued.
–
Seismic devices and the corresponding modelling have tobe encouraged for future projects of demonstrators.
–
Competences in instrumentation must be preserved insome key European laboratories to support the safeoperation of the nuclear installations.
7 VINCO: Visegrad Initiative for NuclearCOoperation

7.1 Key objective

Project VINCO (Visegrad Initiative for Nuclear COoper-ation) was Coordination and Support Action (CSA)carried out jointly by Visegrad countries (Czech Republic,Hungary, Slovakia and Poland) and France. Mainobjective of the VINCO project was to establish acooperation network in Visegrad Group and Francefocused on studies of gas-cooled reactor technology,mainly Gas-cooled Fast Reactors (GFR). This Actioncomplements already established V4G4 Centre of Excel-lence Association and represents the next stage of capacitybuilding in nuclear technologies in Central Europeancountries, focused mainly on ALLEGRO Project (seeFig. 8). Taking into account that development of a newnuclear technology becomes too complex and too costlyfor small and medium-sized countries the need ofinternational cooperation becomes obvious. Visegradcountries decided thus to join their efforts and developcomplementary specializations in participating countries,namely, reactor design and safety analyses in Slovakia,helium technology in Czech Republic, fuel studies inHungary and material research in Poland. This group iscompleted by France, which started already studies ongas-cooled reactors, however, mainly due to current focuson sodium technology, had to slow down studies on GFRs.However, significant knowledge has been gathered earlierin French CEA, therefore its participation in furtherstudies carried out in V4 countries is fully justified andbeneficial for the project.

Main objectives of the VINCO project were thus:
– development of the principles of cooperation and rules ofaccess to existing and planned infrastructure;
–
identification of the specific objectives of the R&Dactivities in the cooperating countries;
–
description and analysis of the existing research, trainingand educational equipment and capabilities;
–
determination of the investment priorities in cooperatingcountries; and
–
setting up of joint research, educational and trainingprojects.
A close cooperation with CEA, France ensured betterdescription of the investments needed in Visegrad Region,tightening of pan-European cooperation and strengtheningof the role of V4 countries, helping them to evolve fromusers to the suppliers of R&D capabilities in nucleartechnologies. Amajor expected impact of the project wouldbe setting up of a distributed regional research centrespecialized in nuclear technologies needed to developGeneration-IV reactors and to improve safe operation ofexisting and planned Nuclear Power Plants in the region.

7.2 Key results

Activities carried out in the frames of the VINCO projectallowed to strengthen the links between the partners,establish running cooperation, especially in the field ofsimulation capabilities in participating institutions, initiatecommon educational and training actions and exchange thepractices of experimental works in hot cell laboratories.Financial and legal framework analysis in V4 countriescarried out within the project helped to identify the possibleinternational cooperation schemes in V4 countries. Mutuallearning and exchange of scientific staff between thelaboratories took place, mainly in form of benchmarklearning exercises on both, the neutronic and the thermo-hydraulic analyses and were devoted to the development ofinputmodels aswell as the efficient use of various calculationtools utilized by different users. Several joint events wereorganized, such as School, workshops and exchange visits.An important part of the project was related to educationalissues. Database of (nuclear) Educational Resources hasbeen prepared and a brochure on Generation-IV technologyprepared and printed. Finally, communication campaignswere organized to provide the information about nucleartechnology for a broader public and establish contact withdecision makers in the V4 Region.


Recommendation for future actions constituted an impor-tant part of VINCO project activities. Main conclusion wasthat cooperation through the international agreementwould bring advantages in the form of reaching of thecritical mass required for such a project, clearly definedstructure, competitions and responsibility. An obstacle canbe politically and procedurally demanding scenario, as thewording of such agreement should be supported by a broad-political agreement of all countries. The ALLEGROEducation and Research Centre (ALLEGRO ERC) was


evaluated as the most promising scheme of cooperation forthe development of the GFR technology and generally forthe development of any Generation-IV nuclear systemtechnology after 2020. The Centre (possibly a part ofESFRI Roadmap) can integrate existing scientific andresearch resources of V4 countries, both human andtechnical, aiming the EU to keep up with other leadingteams around the world in developing advanced nuclearpower sources, with focus to GFR. The integrating aim ofthe ALLEGRO ERC is to prove feasibility and to providesound basis for design of industrial scale nuclear GFRdemonstrator ALLEGRO.

A long-term expected impact of the project is thestrengthening of inter-regional cooperation of V4 countriesin nuclear technologies and related educational activitiesby sharing available infrastructures, expertise, trainingand educational capabilities. Specialization and exchangeof information should allow for the acquisition of the state-of-the-art equipment answering the common needs ofEuropean research institutions related to the developmentof Generation-IV of nuclear reactors.

After completion of the project we may state that themain lines of the expected project impact remain valid.Moreover, VINCO project helped us to identify newobjectives for collaboration within V4 countries, namely,development of High Temperature Gas-cooled Reactor(HTR) technology, a topic especially important in Poland.HTR reactors may produce steam at 550 °C, which isnecessary for chemical industry and may constitute anecessary step in the implementation of more demandingGFR technology. These activities will be carried out in theframes of NOMATEN Centre of Excellence established inNational Centre for Nuclear Research in close collaborationwith strategic partners: CEA France and VTT Finland,which recently has been approved by the EuropeanTeaming for Excellence program constituting a newresearch quality in V4 countries.

8 SESAME: Thermal hydraulics simulationsand experiments for the safety assessmentof metal cooled reactors

8.1 Key objectives

The thermal-hydraulics is recognized as one of the keyscientific subjects in the design and safety analysis of liquidmetal cooled reactors [13]. SESAME project focuses on pre-normative, fundamental, safety-related, challenges for thefour liquid-metal fast reactor systems (ASTRID,ALFRED,MYRRHA, and SEALER)presented in Section 1(see Fig. 1) with the following objectives:
– development and validation of advanced numericalapproaches for the design and safety evaluation ofadvanced reactors;
–
achievement of a new or extended validation base bycreation of new reference data;
–
establishment of best practice guidelines, Verification &Validation methodologies, and uncertainty quantifica-tion methods for liquid metal fast reactor thermalhydraulics.
The goal is to improve the safety of liquid metal fastreactors by making available new safety related experi-mental results and improved numerical approaches. Thesewill allow system designers to improve the safety relevantequipment leading to enhanced safety standards andculture. Due to the fundamental and generic nature ofSESAME, developments will be of relevance also for thesafety assessment of contemporary LWRs.

8.2 Key results8.2.1 Liquid metal heat transfer

A fundamental issue is the modelling of the turbulent heattransfer over the complete range from natural, mixed andconvection to forced convection regimes. Current engi-neering tools apply statistical turbulence closures andadopt the concept of the turbulent Prandtl number basedon the Reynolds analogy. This analogy is no moreapplicable for liquid metals, and robust engineeringturbulence models are needed. Within the SESAMEproject, the main focus was the extension of the validationbase for mixed and natural convection regimes and forgeometrically complex cases, together with further devel-opment and implementation of selected promising modelsin widely used engineering codes.

8.2.2 Core thermal hydraulics

Although experiments in liquid metal are being carried outin the European context on wire-wrapped fuel assembliesand to a lesser extent on fuel assemblies with grid spacers,the data to be retrieved from those experiments will belimited to pressure drops and the thermal field and will notinclude detailed information on the flow field. To derivereference data for the flow field in wire wrapped fuelassemblies, a combination of experimental data andreference high fidelity numerical simulations was set-up.Such need was not only recognized in Europe, but also inthe US. A collaboration was established between theEuropean and US partners allowing to maximize thebenefits of both validation campaigns and to close the gapin the validation process of wire wrapped fuel assemblies.

Missing data for spacer-grid fuel assembly design werealso produced by performing experiments in a liquid metalrod bundle. Such experiments were performed for gridspacers without blockages and with blockages and wereaccompanied by CFD simulations.

8.2.3 Pool thermal hydraulics

Although it is recognized that pool thermal-hydraulics assuch is highly design dependent, the development andvalidation of modelling approaches for pool thermal-hydraulics is not. In order to improve the validation base,liquid metal experiments were performed at differentscales. Firstly, an experiment in the TALL-3D facilitywhich includes a small pool in which thermal stratificationand mixing phenomena can be studied. Four large scaleexperiments have been performed in the CIRCE facilityusing the so-called ICE test section which has beeninstrumented such that relevant data for thermal

Fig. 9. CFD Model of ALFRED primary loop. (Courtesy of CRS4, SESAME Task 3.1.2).


stratification and flow patterns can be extracted. TheESCAPE facility, a scaled down mock-up of MYRRHA, isinstrumented such that relevant data for thermalstratification and flow patterns can be extracted. Inparallel, CFD approaches were developed for all facilitiesmentioned and validated using the experimental data.Finally, the validated CFD approaches were applied tothe MYRRHA and ALFRED reactor design (see Fig. 9).

8.2.4 System thermal hydraulics

Traditionally, the analysis of nuclear system behaviour isperformed using system thermal-hydraulics codes. Suchanalyses are validated using integral design specificexperiments or reactor data from prototype, test, ordemonstration reactors. In recent years, the traditionalapproach of using system thermal-hydraulic codes issupplemented with new multi-scale approaches in whichsystem thermal hydraulics codes are coupled to detailedthree-dimensional CFD approaches. SESAME projectaimed at extending the validation base by providingreference data at different levels. The validation data wereprovided in loop scale by experiments in the NACIE-UPfacility, focused on the multi-scale coupling of thebehaviour in the fuel assemblies and the loop system.Scaling up, the CIRCE facility in the so-called HEROconfiguration is used to provide experimental validationdata. Real reactor data were provided from the Phénixreactor end of life tests. This data allowed validation of thethree-dimensional effects to a much larger extent than thenatural circulation test data which were previously used.Finally, the approaches under development will be appliedto the MYRRHA and ALFRED lead-cooled reactordesigns.

8.2.5 Guidelines and education

One of the main goals of the SESAME project is theestablishment of Best Practice Guidelines, Verification &Validation methodologies, and Uncertainty Quantificationmethods. To this purpose work meetings have beenorganized in Stockholm (2016) in which the currentpractices and experiences of all partners and some invitedexperts from outside the project have been compared and

discussed.With respect to education and training, a lectureseries was organized in 2017 hosted by VKI in Belgium.During the lecture series, experts from the projectdisseminated their knowledge on experimental techniquesand modelling approaches. The textbook [14] was pub-lished as one of the main deliverables to the nuclear liquidmetal community at large. Finally, an internationalworkshop on nuclear liquid metal thermal hydraulicswas hosted by NRG in Petten, with more than 70 lectures,and participants from the entire world.


SESAME project improved the safety of liquid metal fastreactors not only in Europe but also globally by makingavailable new safety-related experimental results andimproved numerical approaches. These outcomes willallow designers to improve the safety of their reactors,which will finally lead to an enhanced safety culture. Forthe future, it is recommended to keep the successfulapproach of SESAME in which experiments, modelling andsimulations moved hand-in-hand. New projects, based onthe outcomes of SESAME, would be implemented enlarg-ing the community, strengthening the partnerships,improving the synergies, leading innovation, enhancingsafety culture at the European and international level.

9 SAMOFAR: a paradigm shift in reactorsafety with the Molten Salt Fast Reactor

The ultimate aim of nuclear energy research is to develop anuclear reactor that is truly inherently safe and thatproduces no nuclear waste other than fission products. TheMolten Salt Fast Reactor (MSFR) has the potential toreach these goals. The most characteristic property of amolten salt reactor is the liquid fuel, which providesexcellent options for reactivity feedback and decay heatremoval. Furthermore, the continuous recycling of the fuelsalt enables one to design a reactor either as a breederreactor with in situ recycling of all actinides, or as a burnercapable of incinerating the actinide waste from otherreactor types.

Fig. 10. Schematic design of the primary circuit of the MSFR showing the reactor vessel and emergency draining system for the fuelsalt.


9.1 Key objectives

The grand technical objective of the SAMOFAR project isto prove the innovative safety concepts of the MSFR byadvanced experimental and numerical techniques, and todeliver a breakthrough in nuclear safety and optimalwaste management. This objective has been split in foursub-objectives:
– delivering the experimental proof of concept of theunique safety features of the MSFR;
–
providing a safety assessment of the MSFR for both thereactor and the chemical plant;
–
updating the conceptual design of the MSFR; – creating momentum among key stakeholders.
Besides the Work Package (WP) on project manage-ment, the SAMOFAR project contains six specializedparts. WP1 deals with the integral safety assessment andthe overall reactor design (see Fig. 10) including thechemical operation plant. WP2 determines experimen-tally all safety-related data of the fuel salt. WP3investigates experimentally and numerically the naturalcirculation dynamics of the fuel salt in the primaryvessel and emergency drain tanks, and the behaviour ofthe salt in the freeze plugs during a drain transient. WP4assesses numerically the accident scenarios identified inWP1, which include the normal operation transients andthe off-normal accident scenarios. WP5 assesses experi-mentally and numerically the safety aspects of thechemical extraction processes, and the interactionbetween the chemical plant and the reactor. WP6covers the dissemination and exploitation of knowledgeand results, e.g. by education and training of youngscientists.

SAMOFAR is the latest MSR-related project in asuccessful series (MOST, ALISIA, EVOL) and started inAugust 2015. The SAMOFAR consortium consists of 11partners from the EU, Switzerland and Mexico, eachproviding a specific own contribution. Besides the partners’contribution, also observers participate to the project.

9.2 Key results

In WP1, the design of the MSFR including the emergencydraining systemhas beenupdated and assessed by a panel ofexperts. A plant simulator has been developed and is nowbeing used to define reactor control strategies andprocedures for the various operation modes of the MSFR,such as full power operation, load-following, start-up andshut-down. A risk assessment methodology has beendeveloped based on the Integrated Safety AssessmentMethodology,whichwill lead to “built-in” safety at the earlydesignstages.Other riskanalysismethodshavebeenappliedto theMSFRandhave led to the identification of postulatedinitiating events and a list of relevant design key-points.

Test calculations with the MELCOR and ASTECsevere accident codes showed that these codes can mostprobably be used, but that some data need to be added. Aspecial setup has been constructed for experimental studiesof actinides in molten fluorides and for the synthesis ofactinide fluorides. Experimental studies on the vaporiza-tion behaviour of the fuel salt revealed the retentionproperties at high temperature. It turns out that CsFremains fully dissolved in the salt, but that CsI needsfurther investigation. A test facility has been made tomeasure viscosity of salts based on ultra-sound methods.Finally the interaction of salt with water under theinfluence of gamma radiation has been investigated.

In WP3, the major experimental contributions in twolarge setups (DYNASTY and SWATH) have beenprepared. For DYNASTY, numerical research has revealedflow instabilities in a natural circulation loop with adistributed heated salt. The DYNASTY facility is inoperation to generate experimental data, which will beused for stability analysis and for the validation ofnumerical codes in WP4. The design and construction ofthe SWATH facility and the test sections in which theexperiments will be carried out have been completed.SWATH uses a molten salt between 500 °C and 700 °C toperform thermal hydraulics measurements, includingphase change phenomena and experiments on freeze plugs.


In WP4, transient calculations based on the scenariosidentified in WP1 will be performed based on leading-edgemulti-physics codes including uncertainty propagation.Verification and validation of these code systems has beendone via code-to-code comparison and by using theexperimental data generated in WP3.

In WP5, the fuel salt processing scheme has beenupdated, and thermochemical calculations have revealedthe transfer coefficients. This data has been used tocalculate the radionuclide inventory at each stage usingnew software, as well as the radioactivity, the decay heatproduction and the shielding requirements. The behaviourof uranium and iodine in the salt has been investigatedexperimentally.

In WP6, a summer school has been organized withfocus on the scientific fundamentals of fluid fuel reactors.Almost 90 MSc/PhD students and young professionalsparticipated. The SAMOFAR website (http://www.SAMOFAR.eu) acts as the portal to reach the publicand for information exchange and for archiving. TheYouTube channel (http://samofar.eu/samofar-youtube-channel/) has been updated with lectures from thesummer school and movies.


The MSFR is a reactor design at low TR level with severalpoints for improvement. To come to a realistic safetyassessment of the reactor, a more detailed design is neededwith better materials data (structural materials, fuel saltproperties, etc), validated simulation models of thespecific phenomena occurring in the MSFR, and reliabledata on the performance of components and processes.These topics are subject of the new SAMOSAFERproposal that focuses, among others, on safety require-ments and risk identification of molten salt reactorsincluding the chemical processing plant; measurementand calculation of the fuel salt retention properties;evaluation of nuclide mobility and the resulting inventoryin each compartment of the reactor including the chemicalprocessing plant; modelling and simulation of phenomenaneeded for the safe confinement of fuel salt; modelling ofheat removal processes, including radiation heat and otherphenomena; reactor operation and control to assessnormal operation and emergency operation; educationand training of students, and dissemination and exploita-tion of our results.

10 ESFR-SMART: European Sodium FastReactor Safety Measures Assessment andResearch Tools

10.1 Key objectives

To improve the public acceptance of the future nuclearpower in Europe we have to demonstrate that the newreactors have significantly higher safety level compared totraditional reactors. The ESFR-SMART project [15] aimsat enhancing further the safety of Generation-IV SFRs andin particular of the commercial-size European Sodium Fast

Reactor (ESFR) in accordance with the ESNII roadmapand in close cooperation with the ASTRID program. Theproject aims at 5 specific objectives:
– produce new experimental data in order to supportcalibration and validation of the computational tools foreach defence-in-depth level;
–
test and qualify new instrumentations in order to supporttheir utilization in the reactor protection system;
–
perform further calibration and validation of thecomputational tools for each defence-in-depth level inorder to support safety assessments of Generation-IVSFRs, using the data produced in the project as well asselected legacy data;
–
select, implement and assess new safety measures for thecommercial-size ESFR, using the GIFmethodologies, theFP7 CP-ESFR project legacy, the calibrated andvalidated codes and being in accordance with the updateof the European and international safety frameworkstaking into account the Fukushima accident;
–
strengthen and link together new networks, in particular,the network of the European sodium facilities and thenetwork of the European students working on the SFRtechnology.
Close interactions with the main European andinternational SFR stakeholders (GIF, ARDECo, ESNIIand IAEA) via the Advisory Review Panel will enablereviews and recommendations on the project’s progress aswell as dissemination of the new knowledge created by theproject. By addressing the industry, policy makers andgeneral public, the project is expected to make ameaningful impact on economics, EU policy and society.

10.2 Key results

The project is currently close to the end of the second yearand the key results obtained at the first phase of the projectare summarised below [15].

10.2.1 Proposal of new safety measures

The key idea is to make a next step in developing the large-power (1500 MWe/3600 MWt) SFR concept, following upthe “line” of the Superphenix 2 (SPX2), European FastReactor (EFR) and ESFR designs and using the set of theGIF objectives as a target. In particular:
– the ESFR core design modifications were aimed atimproving the core map symmetry; optimizing the voideffect; and facilitating the corium relocation toward thecorium catcher;
–
the ESFR system modifications were aimed at simplify-ing the overall design (see Fig. 11) and improving thesafety functions: control of reactivity, heat removal fromfuel, and confinement of the radioactive materials.
10.2.2 Evaluation of core performance

After the new core design was proposed the studies werelaunched to check how this core design will influence theneutronics and fuel performance. In particular:

http://www.SAMOFAR.eu

http://www.SAMOFAR.eu

http://samofar.eu/samofar-youtube-channel/

http://samofar.eu/samofar-youtube-channel/

Fig. 11. General view of ESFR-SMART reactor.


–
six-batch burnup calculations were performed using aMonte Carlo code and the core state specification at theEnd of Equilibrium Cycle were defined, including the 3Disotopic composition needed to calculate the reactivitycoefficients and kinetics parameters as well as the 3Dpower distribution for the following-up thermal-hydrau-lic analysis;
–
fuel performance for a typical cycle was analysed with anumber of fuel performance codes and the gap heatconductance correlation was derived for the subse-quent steady-state and transient thermal-hydraulicanalyses.
10.2.3 Benchmarking of codes

One of the specific objectives of the project is to performfurther calibration and validation of the computationaltools for each defence-in-depth level. In particular:
– a new calculational benchmark has been proposed for thestart-up core of the Superphénix (SPX) Sodium FastReactor based on open publications;
–
a computational exercise on sodium boiling modellingwas organized based on a KNS-37 sodium loop
experiment featuring sodium boiling in pin-bundlegeometries.

10.2.4 Experimental programs

Two specific objectives of the project address new experi-ments: (1) to produce new data to support calibration andvalidation of the computational tools for each defence-in-depth level; (2) to test and qualify new instrumentations inorder to support their utilization in the reactor protectionsystem. In particular:
– new test on chugging boiling regime (CHUG) waslaunched to support the computational activities onanalysis of the ESFR behaviour under sodium boilingconditions;
–
new test on corium jet impingement (HAnSOLO) wasstarted using a water-ice system as amodel of the corium-catcher system;
–
safety rules were formulated for designing a new high-temperature sodium facility;
–
Eddy-current flow meters (ECFM) was qualified for apositioning above the fuel assemblies in order to detectpossible blockages of the sodium flow.



Since the project is only at the second year norecommendations for the future are provided.

11 Conclusion

The paper briefly presents nine Euratom projects startedsince late 2011 in support of the infrastructure and R&D ofthe seven fast reactor systems.

The SARGEN_IV project was the first opportunityto gather together various experts of fast reactors safetyfrom TSOs, research institutes, utilities and universities.The project allowed very fruitful exchanges providing asynthesis on identification and ranking of the safetyissues and the proposal for a harmonization of the safetyassessment practices that could be used further for eachof the concepts proposed by the ESNII. Beside showinghow difficult it is to have a detailed safety assessmentwhen the design of the reactor is not well defined, theSARGEN_IV project contributed to the harmonisationof the different methodologies, crucial for developing aconsistent assessment platform which could be usedeffectively in the decision-making process and to makethe different innovative reactor types publicly acceptablein Europe.

The SILER project demonstrated that the technologyfor the seismic isolation of nuclear facilities already existsand that the main components like isolators (in particularHigh Damping Rubber Bearings and Lead RubberBearings) and flexible joints for pipelines (even the morecritical ones) are reliable enough to guarantee the safety ofthe plant, even in the case of beyond design events. SILERalso confirmed the significant advantages given by seismicisolation, not only in terms of reduction of the seismicactions on the structure and most critical components, butalso from the economical point of view, thanks to thepossibility of standardizing the design of the reactorbuilding, making it substantially independent of theseismicity of the construction site. Some activities ofSILER continued in the ENSII Plus Project (see Sect. 6),regarding the design of the seismic isolation systems of theASTRID and ALFRED reactors.

The ALLIANCE project is helping to realise thevision of a next-generation GFR in one of four centralEuropean countries during the next decade. Outcomes willhelp meet EU energy and climate targets.

The JASMIN project has fostered a collaborativework on the integral Beyond Design Basis Accident(BDBA) ASTEC-Na code development and validation.The project, relied on the PIRTs produced within theprevious CP-ESFR FP7 project, capitalized the largeamount of knowledge produced since 40 years in this field inthe ASTEC-Na code development by collecting andsharing some past experimental program results, anddisseminated it to end users. JASMIN end-products werethe final version of the ASTEC-Na code and the associatedvalidation experimental matrices. Both might be used inthe future not only for R&D activities but also forindustrial applications. Cross-cutting issues were also

investigated and led to the conclusion that the soundbases of ASTEC-Na and the existing similarities withPb-cooled and Pb-Bi reactors, turn it to be a good option todevelop an ASTEC-LM (Liquid Metal) version.

The ESNII Plus project prepared the ESNII structu-ration and deployment strategy, to ensure efficientEuropean coordinated research on Reactor Safety for thenext generation of nuclear installations, linked withSNETP SRA priorities. To achieve the objectives ofESNII, the project coordinated and supported thepreparatory phase of legal, administrative, financial andgovernance structuration, and ensured the review of thedifferent advanced reactor solutions. At the same time, theproject addressed the following technical cross-cuttingareas:
– core physics benchmarking activities of neutronic codesused in Europe and recommendations for their applica-tion to the different advanced concepts. Identification ofR&D needs to improve the core safety;
–
fuel update of theMOX fuel properties catalogue throughadditional measurements performed during the projecton samples previously irradiated in European reactors;
–
seismic behaviour, modelling and analysis of thebehaviour of the demonstrators by implementing seismicisolators including experimental verifications;
–
instrumentation development and measurement techni-ques relevant to safety and in service inspectionand repair.
The VINCO project addresses one of the mostimportant problems of the society: to find energy forfuture generations. Obviously, such a problem cannot beresolved by a small, C&S action; however, VINCOcontributes to its solving by building a research platformable to cope with one of the future concepts, gas-coolednuclear reactors, in Visegrad countries.

Within the SESAME project, new analytical andsimulation methods are being validated with referencedata. Most of these reference data are based onexperimental results and, when not feasible, are comple-mented or replaced by high fidelity simulation data(typically DNS or LES). As such, within these projects,experiments, high fidelity reference simulations andpragmatic engineering simulation will go hand-in-handproviding not only the international liquid metal fastreactor designers, but also the light water community withvaluable new reference data and modelling approaches.

The progress in the SAMOFAR project till now,which is only very briefly summarized in this paper,contains significant results beyond current knowledge,both in the fields of safety assessment, Molten Salt FastReactor design, fuel salt data, experimental evaluation,numerical algorithms and modelling, and the synthesis ofsalts and coatings. Many results were published atscientific conferences, journals and other disseminationchannels to increase the impact of the project. Theinclusion of SAMOFAR related topics in the curricula ofthe university programs has contributed to the dissemina-tion and to the education of students. The SAMOFARproject is scheduled to finish at July 31, 2019.

On one hand, the ESFR-SMART project continuesthe development of the European Sodium Fast Reactor


concept following up the EFR and CP ESFR projectsespecially in terms of safety enhancement and designsimplification. On the other hand, R&D activities insupport of the Sodium Fast Reactors in general areperformed in terms of codes validation and calibration, newexperiments and new instrumentation, support of sodiumfacilities and measurements of MOX fuel properties. Theproject is ongoing and scheduled to finish in August 2021.

Projects have received funding from the Euratom research andtraining programme under grant agreements No. 295446, 295485,323295, 295803, 605172, 662136, 654935, 661891, 754501.

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Review of Euratom projects on design, safety assessment, R