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Overview of experimental facilities in support of Sodium-cooled Fast Reactors (SFR)


In this LMFNS Catalogue 79 facilities in support of development of sodium cooled fast reactors (SFR) operating with mainly water, air or sodium are presented. An overview of these SFR facilities by their most relevant research field (main application) is presented in Table 3 of the IAEA publication, which is copied below.

From the total number of 79 facilities located in 10 countries:
  • 57 facilities are operational and available for research,
  •  7 facilities are currently on standby mode
  • 10 facilities are under construction or upgrade/modification
  •  5 facilities are under design
  • 10 facilities can be also used for LFR development 
To navigate yourself through the SFR facilities presented in the LMFNS Catalogue on the page "LMFNS Facilities Databaseuse the Overview table below.

For detailed information on these facilities see the SFR profiles at the "LMFNS Facilities Database"​ page. 
For this: 1) in the "Reactor type" search box select "SFR" and 2) click on the Facility Profile you are interested in.

​Overview of experimental facilities in support of Sodium cooled Fast Reactors (SFR) 

 Content Editor

Profile Country Facility Main application
          Zero power DBA and DEC Thermal Hydraulics Coolant Chemistry Materials System & components Instrumentation ISI&R Cross cutting facility
SFR 1 ChinaCEDI​​    
SFR 2 China ESPRESSO    
SFR 3 China FRIYG-l  
SFR 4 China HTMTSL  
SFR 5 China HTTCSL  
SFR 6 China MSSPD  
SFR 7 China SIPHON    
SFR 8 China TSBS    
SFR 9 France BACCARA  
SFR 10 France CARNAC  
SFR 11 France CHEOPS - NADYNE_Esa      
SFR 12 France CHEOPS - NAIMMO_Esa      
SFR 13 France CHEOPS - NSET_Esa    
SFR 14 France CORONA   
SFR 15 France DIADEMO Na    
SFR 16 France DOLMEN     
SFR 17 France FUTUNA 2       
SFR 18 France IRINA       
SFR 19 France LIQUIDUS    
SFR 20 France MASURCA  
SFR 21 France MECANA   
SFR 22 France PEMDYN    
SFR 23 France PENELOPE    
SFR 24 France PLATEAU      
SFR 25 France PLINIUS 2    
SFR 27 France VKS2   
SFR 28 Germany ALINA        
SFR 29 Germany DRESDYN    
SFR 30 Germany KASOLA          
SFR 31 Germany NATAN        
SFR 32 Germany SOLTEC          
SFR 33 India 500kW Sodium Loop      
SFR 34 India BIM     
SFR 35 India INSOT Creep loop  
          Zero power DBA and DEC Thermal Hydraulics Coolant Chemistry Materials System & components Instrumentation ISI&R Cross cutting facility
SFR 36 India INSOT Fatigue loop​  
SFR 37 India LCTMF    
SFR 38 India LCTR       
SFR 39 India LEENA  
SFR 40 India SADHANA    
SFR 42 India SGTF    
SFR 43 India SILVERINA Sodium Loop   
SFR 44 India SOWART       
SFR 45 India Sub Assembly Hydraulic Test Rig     
SFR 46 India TSTF
SFR 47 Japan AtheNa         
SFR 48 Japan CCTL  
SFR 49 Japan MELT    
SFR 50 Japan PLANDTL     
SFR 51 Japan SAPFIRE    
SFR 52 Japan SWAT        
SFR 53 Korea Republic of ITSL    
SFR 54 Korea Republic of SOFUS  
SFR 55 Korea Republic of STELLA-1      
SFR 56 Korea Republic of STELLA-2       
SFR 57 Latvia RIGADYN      
SFR 58 Latvia TESLA        
SFR 59 Latvia ST-300        
SFR 60 Russia 6B​        
SFR 61 Russia AR-1​    
SFR 62 Russia B-2      
SFR 63 Russia BFS1      
SFR 64 Russia BFS2      
SFR 65 Russia Pluton  
SFR 66 Russia Protva-1        
SFR 67 Russia SGDI    
SFR 68 Russia SGI    
SFR 69 Russia SPRUT    
SFR 70 Russia V-200       
SFR 71 USA ALEX          
SFR 72 USA Creep lab-MPM   
SFR 73 USA Fracture Mechanics Laboratory   
SFR 74 USA Hydraulic lab-MPM  
SFR 76 USA Material Processing Facility  
SFR 77 USA METL          
SFR 79 USA UW Sodium loops 1&2​      
Total by main application: 0 14 42 11 29 30 34 10


Below are the excerpts from the IAEA Nuclear Energy Series publication "Experimental Facilities in Support of Liquid Metal Cooled Fast Neutron Systems. A Compendium​"​ (in progress). ​​​



Transport of thermal energy from the fast reactor core to the turbine requires high heat removal reliability. The total heat transfer surface and mass of the intermediate heat exchangers (IHX) and the steam generators (SG) in a fast reactor are less than those of the SGs in a pressurized water cooled reactor (PWR). It is achieved through higher temperature differences across the heat exchanger and higher heat transfer coefficients, which are due to the better heat transfer properties of liquid sodium. 
Sodium has a relatively high boiling temperature (882°C), the cooling system can operate at near atmospheric pressure and, therefore, the vessels and heat exchanger tubes can have thinner walls. Sodium is non-corrosive to structural materials used in the reactor and its density and viscosity are in the same range of values as water allowing using matured water coolant technology for the development of sodium cooled systems (e.g. primary pumps). Thus, sodium-cooled systems possess unique parameters providing for superior reliability, operability, maintainability, and long lifetime. These characteristics facilitate reducing the life-cycle costs of SFR. However, strong chemical activity of sodium with air and water possesses the risk of explosions and fires that has to be managed by design provisions and operating procedures to avoid the consequences of the potential reactions with these compounds.


Experimental activities and major R&D topics in support of SFR development

The experimental activities in support of SFR respond to much diversified needs, in terms of topics, but also by the nature of the tests. Indeed, two kinds of tests had to be covered:
  • fundamental physical tests; 
  • qualification of components, systems, or circuits at a scale representing the components of the reactor.
The past experience on SFRs is very important. The associated experimental facilities in support of SFR development have permitted to run many different experimental activities. However many experimental activities are still desired to respond to different emerging needs driven by some new objectives of SFR development driven by new requirements. Indeed, most of SFR projects aim at responding to the targets defined by the GIF. The innovations in the development of new SFRs and support of existing ones are the major motivations for the experimental activities. 
Some of these experimental needs require presence of sodium, but not all of them. In some cases, representativeness can be achieved by using simulant fluids (water, air, sodium-potassium alloy). In this document, the experimental facilities that can use water and air as simulant coolants are also presented.

Consequently, the main R&D requiring experimental activities are driven by the major topics considered in innovative SFR designs: 

Major R&D topics for innovative SFR designs
  • Thermal-hydraulic behaviour (for operation and safety)
This large topic covers many subjects to be studied. Of course it relies on the use of the past experience and the several specific codes. But some complementary experimental validation and qualification still remain a necessity such as: 
  • Thermal-hydraulic of the primary circuit (for new reactors designs), natural convection of the primary loop in particular for the feasibility demonstration of passive DHR system;
  • Behaviour of subassemblies and studies associated to potential situations of sodium boiling;
  • Gas entrainment and creation of vortex on free surfaces;
  • Evaluation of sodium aerosols behaviour on the gas plenum and its influence on the heat transfer coefficients, upper core structure vibrations, and thermal fatigue of the upper core structure due to core outlet temperature fluctuations;
  • Thermal-hydraulics of main components such as IHX or SG;
  • The internal thermal-hydraulics of the fuel subassembly (pressure drop, cavitation, effects of vibration, etc.);
  • Thermal-hydraulics of gas in case of use of Brayton cycle for the power conversion system;
  • Thermal stratification and mixing phenomena in the primary vessel (fluctuations of the temperature at the boundary of stratified region).
  • Some of these tests can be performed using water. That is the reason why in this report in the section devoted to facilities in support to SFRs, some water facilities are listed and detailed. They are devoted to cover basics thermal-hydraulics issues. 
  • However in some cases, mainly when a high temperature range of test or the proper heat exchange conditions are required (such as for the sodium aerosols behaviour or final qualification of sub-assemblies or operation of a new design of SG) some sodium facilities have to be used. It can be relatively large sodium facility implying an important investment cost.
  • Improvement of system reliability and operation (availability, safety, investment protection, etc.).
  • This objective relies, among others, on the performance of instrumentation for continuous monitoring but also on ISI&R. Several needs have been underlined to cover this topic and can depend on the reactor design choices. Just some of them requiring R&D and some examples of sodium experiments are recalled hereafter.

For continuous monitoring further development of techniques and/or performing experiments of eddy current flow meter (ECFM) for sodium flow measurements at the core outlet are needed. The examples include: 
  • development of high temperature ionization chambers; 
  • measurements of concentrations of oxygen and hydrogen in sodium; 
  • techniques for characterization of gas content;
  • detection of sodium-water reaction; 
  • fuel assemblies identification; 
  • sodium free surface level measurements, 
  • core geometry measurements; 
  • sodium leak detection.

For ISI&R the following is needed: 
  • ultrasonic under-sodium viewing (global position checking);
  • ultrasonic in sodium telemetry and surface metrology (accurate location checking); 
  • ultrasonic non-destructive examination (cracks, corrosion or thinning detection) applied on different components and systems (main welding in primary vessel, sodium gas compact heat exchangers, etc.). 
In complement to the development of sensors and signal treatment, this domain could require the use of carriers and robotic arms. It could lead to a dedicated R&D program requiring some analytical sources to validate elementary choices but also rather large sodium experimental pots (few meters diameter) in order to realize qualification of the developed technical solutions. Another field of activity is the development of under sodium repair techniques.

For all these kinds of development, specific experimental facilities are required implying, in most cases, the use of sodium as coolant, but small water installations can either be used to develop acoustic techniques in representative conditions. To cover this need many facilities have been identified. In general, these facilities are not very big in size but versatile by their application. 
  • Improvement of decay heat removal (safety) 
DHR is a major challenge for all types of nuclear reactors. For sodium cooled fast reactors passive DHR based on sodium natural convection could be imagined as a decisive argument in favour of this kind of fast neutron reactors. The behaviour of these systems operating in natural convection is a key point to demonstrate its reliability in case of total plant black out, for example. 
The system analysis, thermal hydraulics and CFD codes are the key tools for system calculations and simulations. A qualification study of these systems and validation of these codes have to be carried out based on some experimental validation. It can require a medium scale facility with a devoted design to confirm the performances of the components involved in such loop but also the operation of a complete loop under natural convection. Several countries are working today at the development of such facilities.

  • Improvement of reactivity control (safety)
In order to improve the reactivity control of SFR, further optimisation of its core design is required. However, improvement of the control rods and shutdown system are also needed to realise the deterministic safety approaches. For example, hydraulically sustained control rods or some other technical solutions, like high temperature threshold based on different physical principles for shutdown system, could be used. Consequently, their qualification in representative conditions (hydraulic tests to study vibrations, risks of up-loading, pressure drop, cavitation, etc.) and mechanical tests in order to demonstrate the operation ability of shut-down system in relevant normal or abnormal conditions are needed. 
These needs require a relatively large sodium facility with a large sodium flow at high temperature (up to around 700°C). 

  • ​Optimization of the handling route (availability, economics)
Handling of the fuel subassemblies in an SFR is significantly different from handling rod bundles in water reactors. First of all, the opacity of sodium requires working “blind” as long as the fuel subassemblies are inside the reactor or in the sodium storage tank. Systems to check for movements and obstacles (ultrasonic “viewing” in particular) have been developed to remedy this drawback. Then, the subassemblies need to be cleaned from the sodium which may remain attached to them before they can be stored in water. These operations require radiological protection and they are performed using remote-controlled equipment. 

The experience feedback showed a gradual extension of the durations of the core renewal campaigns, due, on the one hand, to equipment ageing (more frequent failures) and, on the other hand, to stricter assembly movement control procedures requiring a greater number of checks and hold points during the operations. Additional R&D is necessary to improve the handling and cleaning speeds in order to preserve optimum reactor availability. 

The reliability of this fuel handling route is becoming a critical issue for the reactor concept without external storage vessel. Two main constraints have to be considered: the handling of assemblies with high residual power and the requirement to treat on-line the fuel assemblies from the sodium internal storage to the used fuel assemblies’ pool. 

These constraints induce the necessity to develop innovative handling systems, in comparison with the previous ones and more efficient fuel assemblies cleaning processes (to be defined and qualified). 
To validate the fuel handling systems, the following R&D needs have been identified: 
  • in-air tests for handling mechanical devices such as, for example, fixed arm charge machine (FACM), direct lift charge machine (DLCM); 
  • tests in sodium for some of these devices to examine behaviour of some handling systems in argon and sodium aerosols, seismic behaviour of some handling systems, etc.

  • Design simplification (economics, performances, periodical inspection)
This topic covers very different actions. It can concern the primary vessel and its internals design (for example to be able to address all the periodic inspection), but also the development of electromagnetic pumps for the secondary circuit (components requiring few maintenance actions) or the arrangement of inlet and outlet piping of the main components of the secondary sodium loop. 
For example, concerning specifically the large electromagnetic pump, the development of tools allowing confirming the design represents a field of the potential R&D. In support, experiments using sodium are needed. 
This kind of R&D study requires a sodium facility with sodium flowing at high speed (10m·s-1) and devoted instrumentation to quantify magneto-hydrodynamic instabilities.

  • Improvement of core performances (economics, availability)
This topic requires technological program already mentioned in the thermal-hydraulic paragraph (here above), but also neutronics calculations and potential support of some zero power facilities. 
In this regard, development, validation and qualification of coupled codes (thermal-hydraulics and neutronics) are the key issues.
  • Improvement of the containment control including sodium risks (safety)
  • Sodium aerosols behaviour in atmosphere following a large sodium fire could require some R&D support. Indeed, such studies could bring margins in the safety demonstration, for example with a better knowledge of sodium carbonation in contact with air;
  • Depending on the insulating material choice, the issue of the corrosion in case of sodium leak can either be a concern.
  • Elimination of occurrence of a large sodium/water reaction (economics, availability and safety)
Risk of sodium-water interaction is a major concern in case of use of Rankine (water-steam) cycle. Sodium in the secondary circuit and water of the tertiary circuit are respectively circulating outside and inside steel tubes of the steam generator. In case of tube rupture, the following interaction can be accompanied by relatively complex phenomena (such as wastage and multiple tubes rupture). 
Past experimental programs provided substantial results concerning leak flow rate evolution, pressure waves propagation and mass transfer within the secondary circuit, damages caused to the adjacent exchange tubes and problems (efficiency and rapidity) arising from the sodium-water reaction detection. Representative tests can be envisaged in order to complete experimental database used in different codes representing these phenomena. 

Sodium-water-air reaction occurs when two leaks (water and sodium pipes or envelops) intervene in the same premise due to an external accidental event, for example, that causes rupture of both sodium and water circuits of the steam generator. The risk of explosion has to be considered for the explosion of hydrogen in presence of air. Development of the process involves complex phenomena and their interactions: 
  • pressure peaks; 
  • gaseous bubble growth; 
  • combustion, explosion; 
  • sodium fragmentation, etc. 
There is a need in validated model for these phenomena.
Another way to drastically reduce the risk of sodium-water reaction could be the use of another energy conversion system, such as the use of Brayton cycle with pressurized nitrogen or supercritical carbon dioxide as coolant. 
With this perspective an important field of R&D concerns the different components of this innovative tertiary loop: heat exchanger and turbine mainly. For the heat exchanger, a compact design using printed circuit heat exchanger (PCHE) type modules in order to get higher heat power density compared to a standard shell & tubes heat exchanger type could be preferred. 
However, a PCHE type heat exchanger has never been industrially used with these fluids and it is, therefore, necessary to:
  • test mock-ups in order to qualify the concept; 
  • quantify the heat exchange correlations and investigate the spatial distribution of fluid flows; 
  • investigate challenges of the operation with sodium in narrow channels (draining, cleaning, potential self-plugging, stop / restart, inspection, etc.). 
In respond to various requirements, for example, concerning supercritical CO2, the studies are dedicated to the: 
  • stability of operational conditions near the critical point; 
  • industrial development of the turbine;
  • Na–CO2 interaction, which has less potential consequences than Na–H2O interaction, particularly with regards to the wastage effect. 
The whole operation of gas circuit cycle should be evaluated for all conditions (normal and accidental ones) with support of modelling a system loop eventually. 

  • Cross-cutting topics (e.g. material studies, improvement of system reliability)
The SFR system raises a number of material issues due its environment i.e. corrosion phenomena. The following corrosion effects are needed to be studied:
  • stainless steel in contact with high quality sodium (low impurities level – lower oxygen concentration); 
  • related mass transfer, mechanical behaviour of structures for vessels, pipes and internal components; 
  • special focus on cladding material used for the fuel assemblies. 

The main goal is to confirm the performance of new structural materials of e.g. 
cladding (AIM2, ODS), 
  • new hard coatings materials (tribology studies are then required) with regards to the expected operating conditions (high burn-up, temperature, dose, stress, etc.); 
  • new innovative energy conversion system (ECS), etc.
  • Improvement of behaviour in severe accident conditions
The development and qualification of severe accident codes and mitigation devices for SFR require a comprehensive experimental program. It can encompass:
  • in-pile experiments; 
  • prototypic corium experiments;
  • simulant material tests. 
In particular, depending on the pin design, in-pile experiments could be necessary to study their behaviour during severe accident transients and of in-core mitigation devices. 
In order to better understand the phenomena during severe accidents and to qualify technical solutions to mitigate its consequences, corium experiments are required first at small medium scale then at larger scale, mainly for:
  • fluid-corium interaction; 
  • corium relocation; 
  • core catcher issues.




Experimental studies at facilities in support of SFR

The scientific basis for implementation of the sodium coolant technology has been created as a result of more than fifty years' experimental work. 
Comprehensive studies on thermal hydraulic, mechanisms of turbulent heat exchange, boiling and condensation, physical chemistry, technology and corrosion of structural materials in sodium have been carried out at experimental facilities with circulating sodium. Special attention was given to techniques and measurement technology including development of the unique detectors for measuring temperature and velocity of the coolant, etc.
Detailed data on temperature profiles and heat exchange have been obtained and further used in the reactor core calculations. During the hydrodynamic studies of fuel assemblies and reactor flow sections maximum attention has been given to the measurements of velocity profiles, tangential stresses and flow turbulent characteristics. Detailed thermal hydraulic studies of the full-scale reactor core models including possible overlapping of various core elements;; emerging asymmetrical displacement and deformation of fuel pins and presence of counter flows have been carried out [14, 15].
As a result of comprehensive studies of physical chemistry and sodium coolant technology [16, 17], data on various impurities in the sodium coolant, their equilibrium concentrations, solubility of oxygen, hydrogen, carbon, etc. as well as the kinetics of the reactions occurring in such systems have been obtained. On the basis of these data and as a result of study on corrosion of structural materials the admissible content of impurity in sodium and in protective gas has been determined. The concentrations of oxygen, carbon, hydrogen and nitrogen impurities are limited by corrosion processes, and for gaseous fission products (Cs, Sr) – by radiation environment.
These results confirm the low rate of corrosion of structural materials in sodium at equilibrium concentrations of oxygen and hydrogen during normal operation. The characteristics of the leak-before-break phenomenon (i.e. self-growing micro and small leaks due to a failure of the tube, through which water flows into sodium and their transition into a large leak) have been studied. Requirements to the leakage detection systems have been established. Various methods of cleaning contaminants from the coolant have been studied [18, 19]. Cold and hot (getter) traps have been recommended for the practical applications, and carbon traps have been recommended for purification from caesium. Various designs of coolants’ sample analysis techniques for hydrogen, oxygen, and various forms of carbon, metal impurities have been developed and tested at the facilities. 
Patterns of sodium burning, release of aerosol products and their transport in the building and to the environment have been also studied. The tools and localization systems for combustion suppression and capture of aerosol products, systems of concrete structures protection, methods and means for cleaning and decontamination of the equipment and building after sodium burning have been developed. 
The first priority objective of R&D studies on SFR is the reactor itself.
The main directions of experimental studies on sodium coolant in fast reactors that have been performed and those which still need to be carried out at existing facilities, can be summarized as follows [14-20]:

Thermal hydraulics
  • Obtaining of the main parameters of heat transfer (empirical relations and constants) and temperature distribution of fuel pins for all conditions and operating modes (geometry change, power excursion, peaking factors, etc.);
  • Retrieving data for local turbulent characteristics for the liquid metal single-phase and two-phase channel flow and pool boiling, taking into account the influence of large-scale eddy flows on the thermal stratification;
  • These results will be used for the verification of computer codes and for obtaining constants required for models, which take into account the materials, operation parameters and designs proposed for the future NPPs; 
  • Physical chemistry and sodium coolant technology
  • Obtaining data on solubility, diffusion and dispersion characteristics of complex heterogeneous liquid metal systems and their behaviour, taking into account the spontaneous nucleation of the crystalline phase from a supersaturated solution;
  • Studying of the effect of coagulation processes in the circulation loop. (Coagulation rate is determined by the hydrodynamics of circulating coolant);
  • Studying of mechanisms and kinetics of formation and disintegration of heavy oxides and carbons in a non-isothermal circuit;
  • Discovering of the minimum allowable concentration of oxygen and other impurities in sodium and sodium-potassium alloy to secure the performance of structural materials;
  • Improvement of methods and systems of sodium purification: suspended solids, radioactive impurities;
  • Justifying regimes that, in case of placing the cold traps in the reactor tank, exclude the accumulation of hydrogen in the cold traps of the primary circuit;
  • Development of methods of trapping and reliable localization of tritium emitted during various manufacturing operations.
  • SFR safety:
  • The main objective of SFR safety studies is to ensure the safety of the primary circuit and the reactor core. The following is deemed necessary for the future SFR safety studies: 
  • Justification of the design conditions, excluding the formation of eddy flows in outlet channel of the reactor core and on the surface of sodium pool (gas trapping), passive circulation zones (stratification phenomenon, temperature fluctuations); 
  • Analysis of the consequences of possible non-standard mode of operation (e.g. locks, triggering emergency core cooling systems (ECCS), boiling) and the development of measures that exclude the transition into an accident; 
  • Rationale and development of new technical proposals to avoid the large external sodium systems in Emergency Heat Removal Systems (EHRS) and their other deficiencies;
  • Optimizing of the characteristics of air heat exchanger of the EHRS of SFR: it is necessary to justify the distribution of air flow at the inlet of the fuel assembly, check the sodium circulation resistance in parallel circuit system to study the processes of heat transfer and thermal strength of the wire-wrapped rod bundles in non-stationary Design Extension Conditions (DEC);
  • Examining new design solutions capable to ensure reactor core cooling at sodium boiling. Namely, the sodium plenum above the reactor core. It is required to determine the boundaries of unstable operation, study the dynamics of the propagation of the boiling region in a fuel assembly;
  • Testing passive safety systems (PSS) proposed for protection against overheating of passive safety devices, triggered by reducing the flow rate through PSS and by raising the temperature at its outlet.

The second important element is the steam generator (SG). The substantiation of thermal hydraulic regimes of SG and its system of automatic protection is very important task, which can be specified as follows:
  • To improve safety of a large size SG it is necessary to develop and test the materials and the overall design, which secure the sustained process of self-growing leakages and operative repair in case of water ingress  in sodium;
  • For advanced large-blocks SG, experimental validation of thermal-hydraulic regimes and testing of the system of automatic protection is required; 
  • It is necessary to continue research to improve the leak detection systems by applying various methods: measuring concentrations, vibro-acoustic and other methods.
  • Elimination of sodium fires and increase of hydrogen safety: 
  • Development and experimental validation of measures and technical solutions that minimize the amount of sodium leaks and its contact with the air are needed.
  • Particular attention has to be given to the “leak before break” concept.
  • To improve the efficiency of aerosol containment systems new aerosol filters for emergency ventilation system triggered by sodium leakage and general ventilation system have to be developed.
  • Other:



Sodium coolant is used in primary and secondary circuits of most fast reactors. The third circuit is usually designed as a steam turbine cycle. 
Intensive heat removal in heat-exchange apparatus of NPP is achieved due to good thermal physical properties of sodium. Its disadvantage is the violent reaction with water, accompanied by the formation of alkali, hydrogen and heat. Thus the sodium-water steam generators are most critical elements of the scheme, which largely determine the performance of nuclear power plants.

Extensive experience of operation of sodium-water steam generators confirmed their efficiency. However, establishing the reliable SG design still remains one of the priority tasks.

Water leaks in sodium took place practically in all steam generators of all NPPs with fast-neutron reactors: "Enriko-Fermi" (USA), "Phoenix" (France), Prototype Fast Reactor (PFR), (Great Britain), BN-350 and BN-600 (USSR). Interaction of sodium with water was detected even in the steam generator with two-wall tubes at experimental reactor KNK-II (Germany). At reactor "Phénix", after nine years of operation, all steam evaporation units had been replaced. At PFR, from the very beginning of its operation start in 1974, numerous leaks of water to sodium were observed with the tube plate of U-shaped steam generators in welded joints of pipes. Mainly for this reason, the average load factor of PFR was only 11%.

For today there are no structural materials that can guarantee sufficient stability in a zone of reaction of sodium with water. Therefore, operation of steam generators even with small leaks is inacceptable. Sodium-water steam generators should be equipped with: leak detection, dehydration and drainage of sodium, filling of steam generator cavities by inert gas. This suite of systems is named «Emergency protection system of the steam generator» (EPS). All systems should act by following the established algorithm automatically.

Construction of a large-scale thermal hydraulic facility [26, 27] is deemed to be necessary for experimental assessment of the emergency protection system ensuring safety of SFR steam generators in case of leakage of water (steam) into the shell side (sodium circuit). Thus seal failure of "heat-transfer" surface in SG modules can be simulated by injections of water (steam) into inter pipes space of modules and be fixed by the leak monitoring system. In addition, studies to define parameters of the effective work of the detection systems, minimization of pollution and design damage at leaks of water (steam), of hydrodynamic and temperature effects etc. can be performed at such a facility.