2. Heavy Liquid Metal Research and
Development Activities: towards the LFR
The LFR has been identified by the Generation IV International Forum (GIF) as a technology with great potential to meet the needs for both remote sites and central power stations.
In the GEN IV technology evaluations, the LFR system is top-ranked in sustainability, proliferation resistance and physical protection, and rated good in safety and economics.
A Lead Fast Reactor Provisional R&D Steering Committee has been set up in the year 2005 in the frame of the GIF initiative with members from USA, Japan, South Korea and Euratom.
The European contribution to the development of the LFR is decisive and includes results of Russian ISTC projects, the past FP5 and ongoing FP6 activities on ADS, like IP-EUROTRANS, the Integrated Infrastructure Initiative VELLA and the ELSY project.
An overview of the performed and planned actions towards the realization of the LFR is presented in the following.
2-1. GENERATION IV GOALS AND SYSTEMS
Nuclear energy is expected to play an essential role in the frame of energy needs worldwide in a safe, environmentally clean and economic way.
The wide potential of nuclear energy as an attractive option from the viewpoints of security of supply and global climate change policy is increasingly recognized by energy policy makers, industry leaders and technical experts.
A rising interest in nuclear energy for electricity generation and non-electrical applications is becoming more and more evident as new nuclear plants are built or planned in several countries, including emerging and developing countries.
The aim of the Generation IV systems under development as proposed by the Generation IV International Forum (GIF) is to enhance the future contribution and benefits of nuclear energy utilization.
Nuclear power technology has evolved up to now through roughly three generations of system designs (see fig. 1):
⇒ a first generation of prototypes implemented during the period 1950 to 1970; ⇒ a second generation of industrial power plants built from 1970 to the turn of
the century, most of which are still in operation today;
⇒ a third generation, usually called Generation III/III+, of evolutionary advanced reactors which incorporate technical progress based on lessons learnt through more twelve thousand reactor-years of operation.
The trajectory from current nuclear systems to Generation IV systems is described in “A Technology Roadmap for Generation IV Nuclear Energy Systems” [1] .
The development of Generation IV systems is based on the use of advanced technologies and designs to improve reactor and fuel cycle performance as compared with current nuclear systems and further increase the attractiveness of nuclear energy.
Eight goals have been defined for Generation IV systems in four areas, hereafter reported.
Sustainability
1. Generation IV nuclear energy systems will provide sustainable energy generation that meets clean air objectives and provides long-term availability of systems and effective fuel utilization for worldwide energy production.
2. Generation IV nuclear energy systems will minimize and manage their nuclear waste improving protection for the public health and the environment.
Economics
1. Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources.
2. Generation IV nuclear energy systems will have a level of financial risk comparable to other energy projects.
1. Generation IV nuclear energy systems operations will excel in safety and reliability.
2. Generation IV nuclear systems will have a very low likelihood and degree of reactor core damage.
3. Generation IV nuclear energy systems will eliminate the need for offsite emergency response.
Proliferation resistance and Physical Protection
1. Generation IV nuclear energy systems will increase the assurance that they are very unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism.
Figure 1 . Generations of Nuclear Systems Overview
The goals adopted by GIF have used to identify and to select six nuclear energy systems which appear to be promising in the four key areas.
The six selected systems cover a variety of reactor and fuel cycle technologies including thermal and fast neutron spectra, closed and open fuel cycles and include a wide range of sizes from very small to very large reactors.
Depending on their respective technological level, the Generation IV systems are expected to become commercially available in the period between 2015 and 2030.
Table 1 Overview of Generation IV Systems
All Generation IV systems have features aiming at performance improvement and more sustainable approaches to the management of nuclear materials, e.g., minimisation of uranium requirements and waste arising.
In particular, two main means are relied upon to enhance sustainability through better use of natural resources: closed fuel cycle with reprocessing and recycling of plutonium, uranium and minor actinides using fast neutron spectra, and/or high operating temperatures of the reactor coolant ensuring high thermal efficiency.
Table 1 summarizes the main characteristics of the six Generation IV systems, and hereafter more details are reported.
GFR – The main characteristics of the gas-cooled fast reactor are self-generating cores with fast neutron spectrum, robust refractory fuel, high operating temperature, high efficiency electricity production, energy conversion with a gas turbine and full actinide recycling possibly associated with an integrated on-site fuel reprocessing facility. A technology demonstration reactor needed to qualify key technologies could be put into operation by 2020.
LFR – The lead-cooled fast reactor system is characterized by a fast-neutron spectrum and a closed fuel cycle with full actinide recycling, possibly in central or regional fuel cycle facilities. The coolant could be either lead (most likely option), or lead/bismuth eutectic. The LFR can be operated as: a breeder; or a burner of actinides from spent fuel, using inert matrix fuel; or a burner/breeder using thorium matrices. Two size options are considered: a small transportable system of 50 to150 MWe with a very long core life; and a large system of 300 to 600 MWe. In the long term a very large system of 1200 MWe could be envisaged. The LFR system could be deployable by 2025.
MSR – The molten-salt reactor systems present the very special feature of a liquid fuel. MSR concepts, which can be used as efficient burners of TRU from spent LWR fuel, have also a breeding capability in any kind of neutron spectrum (from thermal to fast), when using the thorium or U-Pu fuel cycle. In both options, they have a very interesting potential for the minimization of radiotoxic nuclear waste.
SFR – The sodium-cooled fast reactor systems use liquid sodium as the reactor coolant, allowing high power density with low coolant volume fraction. The reactor unit can be arranged in a pool layout or a compact loop layout. Plant size options under consideration range from small (50 to 300 MWe) modular reactors to larger plants (up to 1500 MWe). The two primary fuel recycle technology options are advanced aqueous and pyrometallurgical processing. A variety of fuel options are being considered for the SFR, with mixed oxide for advanced aqueous recycle and mixed metal alloy for pyrometallurgical processing. Owing to the significant past experience accumulated in several countries, SFR systems will be deployed by 2020. SCWR – Supercritical water-cooled reactors are a class of high temperature,
high pressure water-cooled reactors operating with a direct cycle and above the thermodynamic critical point of water (374°C, 22.1 MPa). The higher thermodynamic efficiency and plant simplification opportunities afforded by a high-temperature, single-phase coolant translate into improved economics. A wide variety of options are currently considered: both thermal-neutron and fast-neutron spectra are envisaged and both pressure vessel and pressure tube are considered. The operation of a 30 to 150 MWe “Prototype Of A Kind” is targeted for 2022.
VHTR – The very-high temperature reactor is a next step in the evolutionary development of high-temperature reactors. The VHTR technology addresses advanced concepts for helium gas-cooled, graphite moderated, thermal neutron spectrum reactor with a core outlet temperature greater than 900°C, and a goal of 1000°C, specified to support production of hydrogen by thermo-chemical processes. The reference reactor thermal power is set at a level which allows completely passive decay heat removal, currently estimated to be about 600 MWth. The VHTR is primarily dedicated to the cogeneration of electricity and hydrogen, as well as to other process heat applications. It can produce hydrogen from water by using thermo-chemical, electro-chemical or hybrid processes with reduced emission of CO2 gases.
2-2. EUROPEAN LEAD COOLED SYSTEM
The Generation IV (GEN IV) Technology Roadmap [1], identified the six most promising reactor systems and fuel cycle concepts and the R&D necessary to advance these concepts for potential deployment.
Among the promising reactor technologies the LFR has been identified as a technology with great potential to meet the GEN IV requirements.
LFR system is top-ranked in sustainability because a closed fuel cycle is considered, and in proliferation resistance and physical protection because it employs a long-life core.
It is rated good in safety and economics. The safety is considered to be enhanced by the choice of a relatively inert coolant. Due to the R&D needs for fuel, materials, and corrosion control, the LFR system was estimated to be deployable by 2025.
In the third call of the FP6, several European organization have taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of an European Lead-cooled System (ELSY) [2].
The ELSY project aims to demonstrate that it is possible to design a competitive and safe lead fast neutron power reactor using simple engineered features [3].
The use of compact, in-vessel steam generators and a simple primary circuit with possibly all internals being removable are among the reactor features needed for competitive electric energy generation and long-term protection of investment.
The tentative parameters of ELSY are specified in Table 2, and a sketch of the plant is reported in the figure 2.
Figure 2 . Preliminary scheme of ELSY Reactor with, at the right, the sectional views of one of the four identical assemblies, each made of two steam generating
units and one primary pump.
The power is tentatively sized at 600 MWe because only plants of the order of several hundreds MWe will be economically productive on the existing well interconnected grids.
On the other hand, a larger plant will increase the lead mass and the associated mechanical loads on the reactor vessel and its supporting structure.
Pure Lead is chosen as reference coolant.
The use of lead as coolant permits to have some advantages in terms of neutronic performance, plant simplification, improved safety and proliferation resistance. The lead coolant, in fact, presents some useful characteristics, among which:
its has high boiling point, which permits to have an un-pressurized primary system, significant enhancements in passive safety and a potentially high core outlet temperature, with a resulting good plant efficiency;
it has favourable neutronic and thermal-hydraulics characteristics;
it is compatible with air/steam/water/CO consenting plant simplification with respect to sodium cooled reactors;
it is considered a more attractive choice with respect to LBE due to its lower price, higher availability and lower Plutonium production.
The choice of a large reactor power suggests the use of forced circulation to shorten the reactor vessel avoiding excessive coolant mass and reducing mechanical loads on the reactor vessel.
Due to the favourable neutronic characteristics of the coolant, the fuel rods of a lead-cooled reactor, similarly to LWRs, can be placed on a relatively large lattice and this results in low pressure drop through the core. The needed pump head, in spite of the higher density of lead, can therefore be kept low (of the order of one to two bars) with reduced requirement of pumping power.
As a result in the European 80 MW, LBE-cooled XADS [4] a simple gas lift as pumping system with 24 parallel riser pipes could be selected, instead of mechanical pumps, to enhance the primary coolant natural circulation to the specified flow rate.
A test section of this gas lift system has been installed in the CIRCE facility (at the ENEA site of Brasimone) with one full-scale riser pipe. The test result confirms the suitability of gas lift for a small-power reactor [5] [6] .
The proposed thermal cycle could be 400 °C at core inlet to have sufficient margin from the melting point of lead and only 480 °C at core outlet with many advantages in term of reduced corrosion, improvement of the mechanical characteristics (reduced creep) of the structural steels, and reduced thermal shocks in transient conditions.
In terms of efficiency of electricity energy production, the elimination of the secondary system would compensate the reduction of 62 K of the core outlet temperature in comparison to SuperPhenix Na cooled reactor (SPX1).
In fact, a supercritical cycle becomes possible, even with a lower steam temperature.
PLANT CHARACTERISTIC TENTATIVE PLANT PARAMETERS
Power 600MWe
Thermal efficiency 40 %
Primary coolant Pure lead
Primary system Pool type, compact Primary coolant circulation (at power) Forced
Primary coolant circulation for DHR Natural circulation + Pony motors Core inlet temperature ~ 400°C
Core outlet temperature ~ 480°C
Fuel MOX with consideration also of nitrides and dispersed minor actinides
Fuel handling ELSY will seek innovative solutions
Main vessel Austenitic stainless steel, hanging, short-height Safety Vessel Anchored to the reactor pit
Steam Generators Integrated in the main vessel Secondary cycle Water-supercritical steam Primary Pumps Mechanical, in the hot collector
Internals Removable to the greatest possible extent, (objective: all removable)
Inner Vessel Cylindrical
Hot collector Small-volume, above the core
Cold collector Annular, outside the Inner Vessel, free level higher than free level of hot collector DHR coolers Immersed in the cold collector
Seismic design 2D isolators supporting the main vessel
Table 2.Tentative parameters of the ELSY plant
In figure 3 the thermal cycle is compared with the technological limits extrapolated from the European experience on LBE
The reactor vessel designed to operate at the colder temperature of 400°C would be in a safe condition even assuming to temporarily lose the oxygen activity control in the melt [7].
All reactor internals will operate under controlled oxygen conditions, whereas fuel cladding could be coated for a greater safety margin.
An improved thermal cycle at higher temperature could be adopted in the longer term, as new materials will be made available.
327 350 450 400 550 600 650 500 O2 control + alluminization
Vessel Internals Cladding 400 480 ~ ~ ~ ~550 Technological Limits O2 control
Low O2 activity
Core 400 400 Material embrittlement Lead Freezing Temperature °C Outlet Inlet
Figure 3. ELSY thermal cycle
ELSY concept meets alls the goals defined for the GENERATION IV systems, as reported in the following.
Sustainability
Resource utilization. Because lead is a coolant with very low neutron absorption and moderation, it makes possible an efficient utilization of excess neutrons and reduction of specific uranium consumption. Reactor designs can achieve a breeding ratio of about 1, and long core life and a high fuel burnup can be achieved.
Waste minimization and management. A fast neutron flux significantly reduces waste generation, Pu recycling in a closed cycle being the condition recognized by GEN IV for waste minimization. The capability of the LFR systems to safely burn recycled minor actinides within the fuel will add to the attractiveness of the LFR.
Economics.
Life cycle cost.. The economic utilization of MOX fuel in a fast spectrum has been already demonstrated in the case of the SFR, and no significantly different conclusion can be expected for the LFR except from improvement due to the harder spectrum. Because of the favourable characteristics of molten lead, it will be possible to significantly simplify the LFR systems in
comparison with the well known designs of the SFR, and hence to reduce its capital cost, which is a major cost factor for the competitive generation of nuclear electricity. A simple plant will be the basis for reduced capital and operating cost. A pool-type, low-pressure primary system configuration offers great potential for plant simplification. The use of in-vessel Steam Generator Units (SGU’s) and the consequent elimination of the intermediate circuit, typical of sodium technology, is expected to provide competitive generation of electricity in the LFR. This approach is possible because of the absence of fast chemical reactions between lead and water, although the SG tube rupture accident (i.e., pressure waves inside the SGU) must be considered in the design. The configuration of the reactor internals will be as simple as possible. The very low vapour pressure of molten lead should allow relaxation of the otherwise stringent requirements of gas-tightness of the reactor head and possibly allow the adoption of simple fuel handling systems. Corrosion by molten lead of candidate structural steels for the primary system will be minimized by limiting the core outlet temperature. Considering that there will be no intermediate circuit to degrade the thermal cycle and that the expected core inlet temperature of about 400°C is relatively high, the adoption of a high-efficiency water-steam supercritical cycle is possible. Risk to capital. A reduction in the risk to capital results from the potential for
removable/replaceable in-vessel components.
Safety and Reliability
Operation will excel in safety and reliability. Molten lead has the advantage of allowing operation of the primary system at low (atmospheric) pressure. A low dose to the operators can also be predicted, owing to its low vapour pressure and high capability of trapping fission products and high shielding of gamma radiation. In the case of accidental air ingress, in particular during refuelling, any produced lead oxide can be reduced to lead by injection of hydrogen and the reactor operation safely resumed. The moderate DT between the core inlet and outlet temperatures reduces the thermal stress during transients, and the relatively low core outlet temperature minimizes the creep effects in steels.
Low likelihood and degree of core damage. It is possible to design fuel assemblies with fuel pins placed on a lattice larger than in the case of sodium and this allows a large coolant fraction as in the case of the water reactor. This results in a moderate pressure loss through the core of about 1 bar, in spite of the high density of lead, with associated improved heat removal by natural circulation and the possibility of an innovative reactor layout such as installing the primary pumps in the hot collector to improve several aspects affecting safety. Lead allows a high level of natural circulation of the coolant; this results in less stringent requirements for the timing of operations and simplification of the control and protection systems. In case of leakage of the reactor vessel, the free level of the coolant can be designed to maintain a level that ensures the coolant circulation through, and the safe heat removal from the core. Any leaked lead would solidify without significant chemical reactions affecting the operation or performance of surrounding equipment or structures.
No need for off site emergency response. With lead as a coolant, fuel dispersion dominates over fuel compaction, preventing severe re-criticality. In fact lead with its higher density than the ones of oxide fuel and of the low-density metal fuel, and its natural convection flow prevent fuel aggregation with subsequent formation of a secondary critical mass in the event of postulated fuel failure.
Proliferation Resistance and Physical Protection
Unattractive route for diversion of weapon-usable material. The use of a MOX fuel containing MA increases proliferation resistance.
Physical Protection. There is reduced need for robust protection against the risk of catastrophic events, initiated by acts of sabotage because there is a little risk of fire propagation and because of the passive safety functions. There are no credible scenarios of significant containment pressurization.
2-3. HEAVY LIQUID METAL R&D ACTIVITIES
LFR concept development, from the U.S. LFR Program, focused on the development of a small transportable reactor system for international deployment, to the European ELSY (European Lead-Cooled System) project, aimed at showing the possibility of realization and operation of a safe and competitive fast lead-cooled critical reactor.
Wide importance is given to the activities in support of the development of systems for the transmutation of radioactive waste, as can be noticed from the European Union (EU) projects MEGAPIE-TEST (MEGAwatt PIlot Experiment), which has been a first decisive step to design and operate a liquid metal spallation target at PSI, Villigen, with a high-energy proton beam SINQ and EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in Accelerator Driven Systems), which concurs to demonstrate the feasibility of an ADS-type dedicated transmuter.
The main objective of EUROTRANS is, in fact, to carry out a first advanced design of a 100 MWth experimental facility demonstrating the technical feasibility of Transmutation in an Accelerator Driven System (XT-ADS), as well as to accomplish a generic conceptual design (several 100 MWth) of the European Facility for Industrial Transmutation EFIT (realisation in the long-term)
This step-wise approach is termed as European Transmutation Demonstration (ETD) approach, as reported in figure 4 .
In particular, in the frame of EUROTRANS, the Domain DEMETRA “Development and assessment of structural materials and heavy liquid Metal technologies for Transmutation systems” is focused on the HLM technologies and materials.
The activities foreseen is the domain involve the specification and fabrication of reference materials, their characterisation in HLM, and irradiation.
In parallel, project dedicated to address specific issues have been and still are carried out, as, for example, considering the EU framework programmes, TECLA (TEChnologies for Lead Alloys), dedicated to lead technologies, aimed at assessing the use of lead alloys both as a spallation target and as a coolant for an ADS, ASCHLIM (ASsessment for computational Codes in Heavy LIquid Metal flows), aimed at sharing experience in the field of computational fluid dynamics codes applicable to HLM and their benchmarking, SPIRE, devoted to addressed the irradiation effects specific to spallation target environment on basic in-service
properties, VELLA (Virtual European Lead LAboratory) , aimed at integrating the existing European HLM infrastructures, developing synergies and complementarities among the laboratories and the research groups across the EU
Figure 4. Development Scheme of the European Transmutation Demonstration (ETD) – progression from FP5 to FP6.
In this framework, several R&D priorities (and planned efforts) have been defined, focusing on the following major topics:
system design; fuels development; HLM technology materials; component development; demonstration. 2-3.1 MATERIALS
Because the development of new material is a very time consuming process, for the short term deployment it is necessary to make use as much as possible of available materials limiting the activity to their qualification in the new environment.
To establish reactor feasibility, it is necessary to provide a technologically viable structural material capable of withstanding the challenging operating conditions of a LFR.
Austenitic steels, due to the large database available for such kind of materials, especially those of low-carbon grade, are candidates for components operating at relatively low temperatures and low irradiation fluence as is the case of the reactor vessel.
Ferritic-martensitic steels appear to be candidate materials for fuel cladding and structures which are under high irradiation flux.
In the frame of the ADS activity in Europe, there is an ongoing experimental campaign of corrosion tests in stagnant and flowing LBE under different oxygen activities, in which ENEA is strongly involved.
Considering the corrosion behaviour, the main results obtained till now on the reference materials can be summarized, roughly, as follows.
♦ At low temperatures (up to 550 °C), it has been noticed that the corrosion is
mitigated by in-situ growth of surface oxide layers on steels in LBE/Pb with sufficient concentration of oxygen. Maintaining the oxygen concentration above 10-6 wt %, the tests performed in both in stagnant and flowing LBE/Pb have shown that most martensitic and austenitic steels form protective oxide layers. Decreasing the oxygen level (less to ~10-7 wt%) both austenitic and ferritic/martensitic steels may suffer dissolution attack even at ~400 °C.
♦ At temperatures above 550 °C, up to 600 °C, within the oxygen control
range, the formation and quality of the oxide layers on martensitic steels are uncertain for durations up to a few hundred hours, and usually it can be noticed a failure after that duration. For austenitic steels, at the same temperature, the oxides are thin and not completely protective. Moreover, at high temperatures in Pb, oxidation kinetics may be accelerated too much and become detrimental.
While the ferritic/martensitic and austenitic stainless steels, appear a viable solution for the moderate temperature reactor concepts, they are clearly not well suitable to support higher temperature options.
Therefore, while studies are going on worldwide to complete the characterization of these materials, especially in terms of resistance to corrosion from lead or LBE, swelling, high-temperature creep strength and fabricability, special effort is dedicate to investigate, in parallel, new improved materials and corrosion protection barriers.
A possible alternative presently on the horizon to could be represented by the ODS steels, actually in an early stage of development but object of well targeted research activities dedicated to qualify their behaviour in the LFR environment. These kind of steel in the first tests have shown remarkable high-temperature creep properties but need further improvement regarding optimized micro structure and manufacturing issues.
Ceramics, refractory metals, or coated refractories may be necessary for significantly higher temperature (800°C) applications.
In parallel, special attention has to be paid to the development of advanced coatings for steel protection.
In particular, the characterization of FeCrAlY layers, which seem to be extremely promising, has to be preformed.
Extensive testing campaigns dedicated to study the corrosion behaviour of FeAl and FeCrAlY coatings are going on in the framework of different R&D projects, both in Pb and LBE environment, with satisfactory results up to now.
The test campaigns carried out up to now, FeCrAlY coated and GESA system treated 1.4970 steel at temperatures up to 600 °C have shown the absence of dissolution attack and no spallation of the surface layer.
The additional treatment of the coating with the GESA melting process lead to a smoothing of the surface, a removal of the pores and a “welding” of the coating to the bulk material.
FeCrAlY coated and GESA treated T91 samples will be tested in flowing Pb, at 500 °C, under controlled oxygen atmosphere in the framework of ELSY
Then, even if a low-temperature primary cycle is selected, a large programme of basic technology confirmation is necessary covering several aspects like materials specification and fabricability, materials characterisation in lead, materials characterisation under irradiation, advanced thermal-hydraulics, measurement techniques and system behaviour confirmation by means of large-scale integral tests.
2-3.2 LEADTECHNOLOGY
The main objectives of the Lead technology development are:
• the physical and chemical characterisation of the lead and compilation and validation of the necessary databases in the parameter range of interest; • the development and validation of a technique for lead
purification/conditioning before in vessel filling;
• the development and validation of a technique for in-reactor lead purification with reactor in operation and prevention/control of slag / aerosol formation;
• development and calibration of instrumentation operating in lead and under irradiation: oxygen sensors, thermocouples, pressure transducer, flow meters, strain gauges, neutron flux, velocity measurement devices;
• development of techniques for failed fuel detection; • development of techniques for in-core instrumentation;
• development of techniques and instrumentations for ISI (mainly main vessel operating in air from the outer wall and Steam Generator (SG) tubes operating steam side from inside the tubes).
• assessment of lead-fuel interaction phenomena;
• assessment of activation and fission products (mainly iodine, krypton, xenon) diffusion in lead and release process to the cover gas;
• lead, aerosol formation above the lead free level as function of the free level temperature and velocity-field and on the type of cover gas.
2-3.3 IRRADIATIONSTUDIES
The objectives are the characterization of the mechanical behaviour of the reference structural materials (including the oxide/coating corrosion barrier) under fast neutron spectrum exploring the temperature and dpa ranges as defined in the design.
First tests must be performed to address separate-effects behaviour (only fast flux environment) to develop models that will describe the irradiation phenomena and corrosion of the materials. Additional lead irradiation experiments
will be aimed at characterising the combined effect of lead coolant and fast neutron spectrum on the mechanical and corrosion behaviour of the material.
An experimental campaign of irradiation under fast neutron spectrum in pure lead is foreseen (LEXUR II experiment) in BOR 60 reactor, to be performed by ENEA and SCK-CEN in cooperation with Russian institutions: samples of reference materials will be irradiated in the environmental conditions typical of LFR systems, permitting to investigate important phenomena (as LME) on which data are still scarce.
2-3.4 THERMAL-HYDRAULICSINLEAD
Basic thermal hydraulic studies for the development of physical models and the validation of numerical tools, useful for the design and the safety analysis are foreseen. These studies are related mainly to the characterization of heat transfer coefficient, the assessment of the lead/water interaction and the measurement techniques.
Large-scale integral tests to characterise the behaviour of the main systems are necessary especially for the licensing process.
An Integral system experiment is foreseen in ENEA, where will be reproduced the circulation of a sector of the primary lead coolant in the reactor pool. The normal steady state conditions operating transients (e.g. pumping system start-up and core power changes) and incidental transients (e.g. transient from forced to natural circulation) will be experimented.
The experimental interpretation will be made with thermal–hydraulic calculations and the data will support modelling for transient and safety analysis.
2-3.5 COMPONENTDEVELOPMENT
Challenging components for LFR development are the Primary Pumps and the Steam Generators (SG).
Validation of the functional sizing and of materials (mainly the pump impeller) operating in erosion/corrosion environment (lead at high speed) is necessary.
Coating of the impeller is probably a suitable technical solution, but no experience exists at present in Europe.
Alternative promising materials should be tested like the MAXTHAL 312 (Ti3SiC2). The natural circulation capability of the primary system eliminates the need of safety-grade pumps which makes easier the development and even makes possible to replace the pumps as soon as new material and technologies are made available.
Even the SG is a component with important innovative features, namely the operation with lead, the supercritical cycle, and the installation in the Reactor Vessel.
The integration of the SGs inside the vessel is a key-feature for economics, and, consequently, the demonstration of the system capability to tolerate a SG tube rupture accident is a requirement for safety.
The qualification of a high-conductivity material for SG tubes like T91 is beneficial for SG compactness. The SG tube rupture has to be tested at a sufficiently large scale of the SG tube bundle to validate the codes and to optimize the SG geometrical configuration for effective damping of the pressure waves without damaging the reactor core.
For their safety role an extensive test campaign has to be performed for the validation of control rods mechanisms operating in lead and for the DHR system and associated components.
Qualification is necessary also for fuel handling machines.
2-3.6 DEMONSTRATIONFACILITY
The results that are expected for the year 2009, particularly about the economic viability and the competitiveness of ELSY for central station power generation beyond the well known advantages of the fast spectrum should create the conditions for initiating a significant development program including the necessary step of design and construction of a Demo.
From that date the research plan could be focused on the design of a Demo of about 100-200 MWe that could serve the ELSY projects, validating the lead technology and the overall system behaviour.
The Demo will prove the general strategy to use as much as possible simple solutions, available MOX fuel, classical materials and to operate the system at low temperature, and particularly important for ELSY, to reduce the technological risks to a minimum.
Once the correct operation of an “Easy” Demo is demonstrated, more ambitious options (e.g. advanced fuel and new materials for components) will be addressed relying also on the possibility to replace the in-vessel components.
Full power operation of the Demo around the year 2018 could justify to initiate at that date the construction of the industrial prototypes of ELSY, the design of which should be carried out in parallel to the construction of the Demo.
The correct operation of the prototype will create, in turn, the conditions for the international industrial deployment around the year 2025 as foreseen in the GEN IV Roadmap of a reactor for electricity generation complying with all GEN IV requirements.
It is expected that focusing the international efforts, an earlier industrial deployment will be possible in the 2020 time frame whereas more advanced solutions operating at higher temperature for hydrogen production or large-scale plants with CO2 cycle are much more challenging and will be made available not before the years 2035 [2] .
2-4. REFERENCES
[1] GIF, 2002, “A Technology Roadmap for Generation IV Nuclear Energy Systems”, issues by the U.S. DOE Nuclear Energy Research advisory Committee and the Generation IV International Forum, Washington, DC, United States.
[2] EC Project ELSY, 2006, Contract number FP – 036439
[3] L. Cinotti, C. Fazio, J. Knebel, S. Monti, H. Ait Abderrahim, C. Smith, K. Suh "Lead-Cooled Fast Reactor”, May 16, 2006, FISA 2006, Kirchberg, Luxembourg
[4] L. Cinotti et al., “The eXperimental Accelerator Driven System (XADS) Designs in the EURATOM”. 5th Framework Program – Journal of Nuclear Materials 335 (2004) 148-155.
[5] Report ENEA HS-A-R-016, “Report on Gas Enhanced Circulation Experiments and Final Analysis (TECLA D41)”, G. Benamati, G. Bertacci, N. Elmi, G. Scaddozzo, January 2005;
[6] Tarantino M., Report ENEA HS-F-R-001 “Gas Enhanced Circulation Experiments On Heavy Liquid Metal System”, 2007.
[7] A.Aiello, A Azzati, G. Benamati, A. Gessi, B. Long, G. Scadozzo, “Corrosion behaviour of steels in flowing LBE at low and high oxygen concentration”,Journal of Nuclear Materials 335 (2004) 169-173.
[8] C. Fazio, A. Alamo, A. Almazouzi, D. Gomez-Briceno, F. Groeschel, F. Roelofs, P. Turroni and J. U. Knebel, “Assessment of Reference structural materials, heavy liquid metal technology and thermal-hydraulics for European waste transmutation ADS”, Proceedings of GLOBAL, Tsukuba, Japan, Oct 9-13, 2005
[9] EC Integrated Project EUROTRANS, 2005, Contract number FI6W-CT-2004-516520
