In this Compendium it is assumed that under the term "lead fast reactor (LFR)" all heavy liquid metal cooled fast neutron systems (FNS) are included, i.e. lead and lead-bismuth cooled reactors, both critical and subcritical installations (including LFR/ADS fast neutron systems).
In this report 72 facilities in support of LFR development are reported and classified according to their most relevant research fields (main application). In Table 4 an general overview of the LFR facilities is presented.
From the total number of 72 facilities located in 13 countries and EU:
- 56 facilities are operational and available for research,
- 7 facilities are currently on standby mode
- 6 facilities are under construction or upgrade/modification
- 3 facilities are under design
- 12 facilities can be also used for SFR development
As should be noted, for any research field (a part for the zero power facilities for V&V and licensing purposes), several facilities are currently available, attesting the great interest worldwide on the LFR development, which involves mainly European Union, China and Russia.
To navigate yourself
through the LFR facilities presented on the LMFNS Catalogue on the page "LMFNS Facilities Database" use the Overview table below.
For this: 1) in the "Reactor type" search box select "LFR" and 2) click on the Facility Profile you are interested in.
Overview of experimental facilities in support of Heavy Liquide Metal cooled Fast Neutron Systems, i.e Lead/LBE cooled fast reactors (LFR)
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).
SPECIFIC INTEREST FOR HEAVY LIQUID METAL COOLANTS (LEAD and LBE-eutectic)
The attractiveness of lead–bismuth eutectic alloy as a coolant is due to its moderate melting point (125°С), high boiling temperature (1638°С), practically eliminating the possibility of its boiling onset in the high temperature areas, and very low chemical activity with respect to air, water and steam, thereby preventing any risks of explosions and fires. Low working pressure in the circuit increases the reliability and safety of the components, simplifies the design and manufacturing technology and significantly facilitates operation of the primary system components. In the stage of designing NPPs for nuclear submarines , the benefits of lead–bismuth eutectic outweighed its drawbacks, such as corrosion and erosion activity with respect to structural materials, as well as polonium accumulation under radiation.
The interest of designers of the larger size NPPs with fast reactors using liquid lead is due to its low chemical activity with respect to water and air, its high boiling point (1745°С) and the insignificant rate of polonium accumulation under radiation. In addition to the drawbacks of lead–bismuth eutectic, the lead melting temperature is a higher melting point (327°С).
Both LBE and lead coolants dissolve the oxygen. By carefully controlling the oxygen concentration in the liquid metal, stable oxide films can be formed on the steel surface . The oxide films separate the steel from the corrosive liquid metal, and then the heavy direct dissolution of the steel is inhibited. Corrosion of stainless steels by liquid lead and LBE can be significantly mitigated by applying oxygen control technology. The corrosion inhibition efficiency of the technology depends on not only the oxygen concentration in the liquid but also the type of structural steels as well the system operating conditions. The low concentration of oxygen leads to vanishing of the oxide film on steel cladding and corrosion, while high oxygen concentration results in oxide build-up in the coolant. For maintaining the protective films formed through reactions between oxygen and steel constituents and also avoiding contamination of the liquid, the oxygen concentration must be controlled instantaneously in a certain range. The operational limits are narrower for the pure lead thus making the oxygen control system in lead coolant more sophisticated. The oxygen control technology, including oxygen measurement and control methods, has been well developed. However, more studies are needed to determine the reasonable operating range of oxygen concentration in a non-isothermal loop system.
LFR design objectives and on-going R&D
Extensive R&D efforts are ongoing worldwide, addressing issues related to lead and LBE technology concepts including those performed under GIF. Research activities are ongoing and are expected to continue in the future targeting to design and construct an LFR/ADS demonstrator.
R&D efforts are necessary for completing the design, support the pre-licensing and starting with the construction of such systems. Such activities require the identification of the technological gaps, which, in some cases, are design dependent, and represent the key topics for LFR/ADS developments.
The technological issues of LFR development can be summarised by the following main topics [9-11]:
- Material studies and coolant physical-chemistry
Corrosion of structural materials in lead-alloy is the main issue for the design of LFR/ADS. The topic is related to lifetime limits and circuit design. The following sub-issues are considered:
Corrosion/dissolution in Lead-alloy of
- fuel cladding;
- reactor internals, components, and heat exchangers;
- corrosion at high temperatures (related to development for long term perspectives);
- corrosion/erosion of pump impeller materials;
- corrosion inhibitors development (i.e. coating).
Efforts have been devoted to short/medium term corrosion experiments in stagnant and flowing LBE. Few experiments were carried out in pure lead, and new experimental campaigns are planned. Research activities are still needed on medium/long term corrosion behaviour in flowing lead and LBE under oxygen control.
Tests at higher temperature (at least 650°C ) and higher velocity (up to 2ms-1) are still missing being identified as needed to cover long term development of the technology as well as to perform specific tests under particular conditions (i.e. DEC).
The embrittlement and degradation of structures by liquid metal is still an open issue. The condition represents the lower bound of mechanical properties of liquid metal exposed materials. It consists in the reduction of fracture toughness and of ductility of un-irradiated materials after long term exposure to lead and lead alloys. In order to be reproduced with a typical standard procedure it is necessary to experience and to standardize tests with respect to:
- liquid metal embrittlement (LME);
- stress corrosion cracking;
- fretting in HLM.
These experiments are necessary and should be conducted for both static and flowing conditions in testing machines where the specimens are exposed to the liquid metal.
Full development of GEN IV programmes foresees the future increase of reactors operating temperatures (beyond 550°C). This evolution is aimed to enhance the thermodynamic efficiency and introduce the co-generation processes. This challenging goal requires to test:
- “new” materials such as ODS steels, refractory alloys, SiC composites, “MAX” phase materials;
- coated materials.
Moreover, the reactor vessel, the structural materials, the internals and the fuel cladding are subjected, with different extent, to several degradation mechanisms such as neutron irradiation, thermal ageing and corrosion.
In case of LFR systems, research activities are generally aiming at understanding, quantifying and predicting such effects on critical components of a nuclear power plant. Focus is given on the
- performances of the materials in terms of neutron irradiation induced embrittlement;
- behaviour of stress corrosion;
- neutron irradiation induced effects such as creep and swelling.
The main objective is to determine whether or not irradiation will promote embrittlement and corrosion attack by HLM.
The current status of knowledge is not completely addressed, and more experimental investigations are needed, providing high quality data on the material behaviour. It is expected that assessments of fuel cladding and structural core materials, subjected to both high temperature in a lead environment and fast neutron flux, are critical issues.
Summarizing, the following main issues on irradiation performance of candidate materials are of primarily importance for LFR systems development:
- corrosion in HLM under irradiation (coated and uncoated material);
- irradiation embrittlement of selected materials;
- irradiation creep;
The impact of irradiation on materials is a critical issue for LFR development. The experimental infrastructures needed to address these issues are currently available irradiation machines and research reactors.
Concerning the impeller pump materials, the relatively high velocity of liquid metal implies that structural materials are possibly subjected to severe corrosion-erosion conditions that might not be sustained in the long term. The materials of the pump impeller have to satisfy such demanding requirements which deserve specific experimental installation as:
- capability to withstand to an exposure to high temperature lead (up to 480°C, and higher for long term perspective);
- capability to withstand to corrosion/erosion effects due to high relative coolant velocity (10ms-1, and up to 20ms-1);
- demonstration of reliability and performances of the pump for a long term operation.
Chemistry control of the coolant and cover gas is another critical issue of operating LFR. It is essential to control the concentrations of impurities, because of the potential for activation and also because of the possible effect on corrosion, mass transfer and scale formation at heat transfer surfaces. Therefore, the coolant chemistry control includes
- oxygen control;
- pollution source term studies;
- mass transport;
- filtering and capturing techniques;
The following specific issues are considered:
- coolant quality control and purification during operation (i.e. oxygen control, in particular, its thermodynamic control oxygen sensor reliability, coolant filtering, HLM purification, HLM cleaning from components);
- cover gas control (i.e. radiotoxicity assessment of different elements, migration flow path into cover gas, removal and gettering).
- Core integrity, moving mechanisms, maintenance, ISI&R
The fuel manipulator (or handling) system is used for the purpose of controlling the reactor power distribution (by off-line shuffling). The system is also used for storing and handling the fuel assemblies during its overall lifetime (i.e. from the arrival up to the spent fuel storage).
In current LFR/ADS designs, refuelling and shuffling are performed remotely. The design and the operation of such machine, has to be tested before the installation in the reactor, for demonstrating its capability to fulfil its functions in reliable and safe way. This requires the assessment in an experimental facility of the prototype machine as well of its component for qualification purposes. On this regard, the following sample experimental activities are considered:
- cold testing of components (in air);
- testing of submerged components;
- reliability of fuel handler components;
- integral test, including fuel recovery strategy and in-core rescue strategy.
In this frame large pool facilities (i.e. CIRCE, COMPLOT, CLEAR-S - the profiles are presented on the CD-ROM in this Compendium) might be in principle suitable for addressing the issues, thanks to their high flexibility. The fuel assemblies maintain and position the fuel within the reactor core, provide the cooling of the fuel and ensure shielding from radiation streaming from the core. Therefore, different issues need to be experimentally tested to verify the suitability of the design features, including ensuring the structural integrity of the fuel assembly. In particular, the following issues need to be considered:
- mechanical and structural integrity of fuel assembly, in connection with fuel loading procedure;
- wide range of operating conditions;
- flow induced vibrations;
- spacer grids - fuel pin interactions.
Moreover, impact of neutronics on control and shutdown systems should be taken into account in core design. The shutdown systems should be designed to meet requirements for the normal and abnormal up to accident conditions.
The following areas of investigations related to the issues related to the core cool ability; integrity and the behaviour of the safety systems (i.e. control rods) need to be addressed:
- loss of core cool ability and integrity; identification of the initiating events, including those connected with control rods mechanisms design (e.g. control rod withdrawal and ejection);
- fuel coolant interaction;
- fuel degradation mechanisms and behaviour (up to its release in the primary system);
- dispersion and relocation of fuel in the primary system;
- impact of seismic loads and sloshing;
- control rod(s) design and mechanisms;
- testing of control rods mechanisms operation and performance;
- reliability of control rods mechanisms and components;
- impact of seismic loads and sloshing: demonstration that the system is qualified to design basis earthquake to permit shutdown rods to drop into the core.
- Steam generator functionality and safety experimental studies
The LFR designs are pool type reactors, which have the SGs (or the heat exchangers) inside the reactor vessel. This implies that the interaction between the secondary circuit coolant and the HLM may occur. Thus, the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a safety issue in the design, but also in the preliminary safety analysis, of these reactor types.
There are two major topics of investigation in case of Steam Generator Tube Rupture (SGTR) postulated event:
1) to understand the phenomena involved in the accident scenario,
2) to study how to prevent or mitigate the consequences of the event, reducing the primary system pressurization.
The areas of investigation include:
- pressure wave propagation across the primary system,
- steam transport in the primary system,
- steam entrainment into the core,
- lead/LBE-water interface phenomena;
- rupture/leakage detection systems;
- tube rupture mitigation countermeasures.
Concerning the SG, the main qualification studies regard the:
- design validation;
- unit isolation on demand;
- pressure drop characteristics;
- behaviour of the components in normal operation (e.g. forced, mixed and natural convection), operational transients and accident conditions.
Apart from the SG, in a reactor there are several auxiliary systems, which need to be qualified for nuclear applications, such as:
- decay heat removal system (DHR);
- dip-coolers and isolation condenser;
- reactor vessel auxiliary cooling system (RVACS);
- fuel assembly transport system;
- spent fuel assembly transport and cooling system.
- Thermal hydraulics
Since several years, the research on heavy liquid metals thermal-hydraulics has been conducted in many countries. Nevertheless, open issues are still pending for LFR development.
The objectives of the activities in relation to pool thermal-hydraulics are twofold:
1) gathering experimental data in geometry and with boundary conditions which may improve the knowledge of phenomena/processes at component and system levels;
2) generating databases for supporting the development and demonstrating the capability of computer codes to predict phenomena/processes relevant for the design and safety.
The following not exhaustive list of topics is identified as relevant at component and system levels:
- flow patterns in forced convection, including:
- stratification (inducing thermal stresses);
- stagnant zones;
- surface level oscillations;
- transition to buoyancy driven flow;
- natural convection flow:
- pressure drop;
- surface level oscillations;
- fluid-structure interaction;
- thermal fatigue issue;
- sloshing due to seismic event tests.
Moreover, it is important to test the fuel assemblies on the basis of thermal-hydraulics parameters (i.e. pressure losses, flow distribution, velocity field, clad wall temperature distribution, etc.) and the geometrical features, such as: rod bundle lattice, sub-channel geometry, spacer grids.
The following not exhaustive list of topics should be experimentally investigated for supporting LFR systems development:
- heat transfer in forced and natural convection (including transition);
- sub-channel flow distribution;
- sub-assemblies flow distribution (i.e. open wrap, wrapless);
- cladding temperature distribution and hot spots;
- pressure drop;
- fluid-structure interaction;
- flow induced vibrations;
- grid-to-rod fretting;
- fuel assembly bow
The topics of investigations involve also the study of the potential sources and the consequences of the core damage.
Integral tests are mandatory in this frame since full scale facilities are not available. Data can be extrapolated to the full scale, if the test facilities and the initial and boundary conditions of experiments are properly scaled, i.e. the scaling will not affect the evolution of physical processes important for the postulated accident scenario. This evaluation determines whether the data may be used in nuclear plant safety analyses.
On the other side, integral tests are fundamental for supporting the development and demonstrating the reliability of computer codes in simulating the behaviour of a LFR, during a postulated accident scenario: in general, this is a regulatory requirement. Applications of computer codes to accident analyses require the implicit assumptions that these codes have the capabilities to scale up phenomena and processes from test facilities to full scale plant conditions. However, the different scale, in terms of geometry, characterizing any facility and a nuclear plant does not ensure a priori that a code, which is able to reproduce a generic transient in a scaled facility, is also able to simulate with the same accuracy the same transient in a full-scale LFR.
These considerations imply that integral tests are unavoidable, and complex activities, which involve the following objectives and areas of investigations, are needed:
- phenomena and processes at system level and connected with design, safety and operation issues;
- flow blockage studies and related experimental/modelling investigations;
- simulations and analyses of a broad spectrum of accident scenarios;
- accident management procedures;
- component testing;
- scaling issues;
- generating databases for supporting licensing process;
- codes assessment and validation.
- HLM pump and corrosion/erosion studies
The main pump is a critical component for the LFR development. The reasons are related to the physical properties of the coolant, which flows inside the system, the relevance of the component in relation with the reactor safety. Indeed, the main coolant pumps in nuclear technology are nuclear grade components.
It refers to a process of rigorous manufacturing quality assurance for those components that are especially critical to reactor safety. Notwithstanding this, postulated initiating events in safety analysis and licensing refers to malfunction of main coolant pumps (e.g. single or multiple main coolant pump failure, locked rotor and shaft seizure).
It implies that experimental investigations are needed on materials, properties of the mechanical parts (i.e. impeller, bearings and housing), performances tests and reliability of the component. The R&D activities related to the pump impeller material have been already outlined. Other activities are connected with:
- bearing qualification tests;
- integral pump tests addressing pump performances and long term reliability.
LFRs impose higher requirements to instrumentation due to the higher thermal loads, higher temperatures, high fast neutron flux, corrosion effects and the opacity of liquid metals. These harsh environmental conditions of the reactors, which can be even higher than the operational temperatures that must be detected and recorded reliably in case of abnormal operation up to accident conditions, significantly limit the choice of instrumentation.
A common requirement for a LFR reactor design is therefore the understanding and development of materials and structures capable to function reliably for a long time in a harsh environment described above. Moreover the position of instruments in a system has influence on their performance.
For most of the parameters to be measured, technical solutions for out-of-pile conditions are available or are currently under development and qualification. For in-pile conditions, the improvement and qualification of instrumentation has to be initiated and promoted, especially with respect to aging and size.
Concerning neutronic issues related to LFR/ADS development, the following issues need to be addressed:
- Validation measurements for nuclear data improvement:
- threshold processes : (n, n’) and (n, xn) reactions;
- MA cross-sections.
- Validation measurements for licensing & operation:
- uncertainty reduction on cross-sections (Pb-MA-Pu-241, Pu-242, in high energy range <1 MeV);
- determination of flux gradients in fast spectrum;
- reactivity effects (voiding of HLM coolant in MOX core, secondary scram system).
The above areas and topics were identified for zero power reactor experiments (such as, e.g., MASURCA and VENUS).
Concerning data libraries and neutronic code validation, the task includes experiments producing benchmarking data for:
- validation and improving minor actinides cross sections
- non elastic neutron scattering thresholds
- libraries for innovative materials (SiC, ZrC, Zr3Si2 and others)
Concerning the operational and control issues the following needs to be addressed carefully
- determination of flux peaks and gradients
- absorber and reflector worth
- degraded geometry studies
Experimental studies at facilities in support of LFR
Concerning LFR, recent developments carried out in last decade regarding ADS as well as FR, evidenced lead coolant as an emergent technology potentially complying with all Generation IV goals (with specific reference to nuclear cycle sustainability). Thanks to the high boiling point of lead together with its inert chemical nature, its favourable neutronics and safety characteristics, the LFR technology have attracted the attention of research organizations and industry as a credible alternative to the other fast neutron reactors technologies under research and development.
The most challenging issue to be thoroughly analysed is the compatibility of the structural materials and coolant chemistry handling. It has to be highlighted that important actions have been taken so far for the development of the LFR technology .
Since the complete development of the fuel cladding material cannot be done in a short time, most of the current designs foresee retaining the relevant design choices adopted for ELSY and ELFR (see Ref. ). Namely to cautiously limit the bulk core outlet temperature to 480°C, to adopt the already qualified material for cladding and improve it against HLM corrosion by corrosion resistant coats. The absence of chemical reactivity of lead allows placing the SG directly in the primary pool without any intermediate system. The adoption of this advantageous configuration requires, however, an accurate analysis of the consequences of the SG tube ruptures (SGTR) accident in terms of damage to the surrounding structures. R&D studies (both, calculations and experiments) have been performed on this issue to implement provisions having the purpose to prevent and limit possible effects of the shock wave and to avoid the entrance of water vapour into the core.
Comprehensive thermal hydraulic simulations of large components and plant configuration have been performed in full scale at large HLM experimental facilities. They allow even to point out the thermal mechanical effects of thermal striping and stratification in order to optimize the design and the materials.
There is no yet operating experience and feedback on reactors cooled by pure lead. About 80 reactor years of experience and feedback were accumulated during operation of lead-bismuth eutectic cooled reactors used for Alfa/Lira-class submarines and land-based facilities in the former Soviet Union. The related feedback as well as experience from licensing of these reactors is, however, not easily available.
On the basis of analyses performed in the frame of GIF Risk and safety working group (RSWG) (see Ref. ), the main safety features and issues for LFR can be summarized as follows:
Materials and coolants
- Molten lead is corrosive and is oxidized if oxygen concentration is not controlled;
- Molten lead might erode structural materials, metallic impurities produced by corrosion can be transported in the primary system;
- Lead vapours are chemically toxic;
- Large specific weight of lead and total primary inventory might, in case of external excitations, challenge structural integrity of systems or components;
- Large quantities of coolant in the main vessel of pool LMFNS (both SFRs and LFRs) may lead to complex flow patterns and interactions between the coolant and structures;
- Lead is optically opaque.
The specific aspects related to the main safety functions, such as reactivity control, heat removal and confinement of radioactivity, can be outlined as follows:
- Ruptures of steam generator tubes might lead to over-pressurization of the primary circuit, sloshing and steam/water entrainment resulting in a positive reactivity insertion;
- Lead density reactivity coefficient might be positive in some core regions;
- Loss of core geometry (core compaction) might lead to a positive reactivity insertion and power increase.
- Heat removal safety aspects
- Lead has high freezing point (327°C) with a potential for coolant solidification. Mechanical stresses might be exerted on structures during unfreezing if a proper melting sequence is not applied;
- Accumulation of corrosion products might lead to coolant blockages;
- Confinement of radioactivity:
- Corrosive properties of molten lead could challenge confinement barriers, in particular the cladding of the fuel pins.
- Anyway, recent project (i.e. ELSY of the EU, (see Ref. ) has shown that lead properties, previously considered serious drawbacks of the lead technology can be turned into advantages by proper design:
Promote the experience. Outlook for the LFR development
- Identify innovative nuclear research and development (R&D) activities that require research reactors (RR) support;
- Identify existing (or soon to be operational) RR facilities capable of supporting innovative nuclear development;
- Quantify the capabilities of the identified facilities within the context of the required research support;
- A short vessel (less than 10m for a 600MWe reactor) is possible. That resolves the issue of robustness against the seismic loads;
- About 1.5m level difference between the cold and hot collectors at normal power operation is sufficient to feed the core and simplifies the primary system configuration;
- Refuelling machine operating in gas would resolve the issue related to in-vessel handling in an opaque medium at high temperature;
- No component is connected to the reactor vessel making possible replacement of all components, including the cylindrical inner vessel, through reactor roof penetrations and eliminating the need of repair under hot lead; Thanks to the hard neutron spectrum, most threshold fission reactions of fissionable isotopes are effectively triggered, determining low minor actinides equilibrium concentrations in the fuel.
FUTURE DEVELOPMENT OF FACILITIES IN SUPPORT OF LFR
According to the previous overview, starting from the existing facilities already mentioned in the Compendium, toward LFR demonstration (up to Technological Readiness Level 7 - TRL7), the construction of the following facilities is recommended:
- Experimental facility for corrosion testing of materials in lead/LBE environment at high temperature (650°C) and high velocity (2ms-1);
- Experimental facilities for studies on LME, fatigue, creep, stress-corrosion cracking up 650°C and in HLM (with Oxygen control systems - OCS);
- Experimental facility for SG tube rupture having an interaction volume meaningful for testing a significant portion of SG and representative for “direct” extrapolation to reactor vessel dimensions. This facility shall provide a realistic reproduction of the tube rupture (i.e. refurbishment of CLEAR-S);
- Experimental facility for the heat exchange components: SG, DHR system. The facility will permit testing in a safe way components having a total heat exchange power as high as 2 to 10 MW for LFR;
- New facility or, alternatively, renovation of the existing ones, for investigating of pool thermal hydraulics of HLM. The alternative option may be selected following an upgrade of the instrumentation of the existing facilities (i.e. ESCAPE, CIRCE, CLEAR-S);
- Facilities (i.e. COMPLOT) to perform full-scale experimental qualification of the fuel elements in a hydraulic flow through the core in normal and accidental conditions. It should include multipurpose functions to support the designer with the full scale basic tests (e.g. operability of the handling machine, operability and insertion speed of the control rods, capability to cool fuel assemblies during refuelling) to test relevant mechanical components before implementing in the demonstration reactors;
- A facility for studying the originators of core damage events and investigating severe accidents;
- More experimental facilities to perform integral qualification tests of the main coolant pump (i.e. CLEAR-S). Equipment should include qualification instruments for thermal hydraulics, vibration dynamics, mechanical forces, and overall performance;
- An experimental facility for the corrosion tests, having larger test section (larger diameter) than those currently available. This facility should be able to test the performances of a specific number of pin simultaneously with the coolant velocity up to 2ms-1;
- A coolant chemistry facility for LFR design investigating the coolant and a facility for studying the fission gas release in the cover gas;
- Test facilities aimed at investigating the fuel coolant interaction (basic chemistry) and the fuel dispersion in the primary system. The implementation of suitable hot cell is needed;
- Test facilities aimed at seismic testing.
Demonstration reactors BREST-300, ALFRED, MYRRHA, CLEAR-I are currently under development and will be created aiming:
- to demonstrate (TRL7) the HLM-cooled nuclear system technology;
- to achieve high safety standards and to enhance non-proliferation resistance;
- to assess economic competitiveness of the LFR technology, including high load factors;
- to demonstrate better use of resources by closing the fuel cycle;
- to validate selection of materials.
The above mentioned demonstration reactors are also designed to confirm that the newly developed and adopted materials, both structural material and innovative fuel material, are able to sustain high fast neutron fluxes and high temperatures.