ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS Alessandro Alemberti Alessandro.Alemberti@ann.ansaldo.it TECHNICAL MEETING ON IMPACT OF FUKUSHIMA EVENT ON CURRENT AND FUTURE FAST REACTOR DESIGNS 19-23 MARCH 2012: HELMHOLTZ-ZENTRUM DRESDEN-ROSSENDORF DRESDEN, GERMANY
SUMMARY From ELSY to ELFR (The LEADER project) ELFR DESIGN LEAD: Advantages (how they are used in ELFR) Disadvantages (design provisions to limit.) SAFETY APPROACH MAIN DBA and DEC Results ELFR response to extreme natural events Conclusion
LFR development in Europe - The ELSY project In 2006 the EC partially funded, within the FP6 programme, the ELSY project (European Lead-cooled System) The ELSY project has been conducted by a large consortium of European organizations to demonstrate the feasibility of a competitive and safe Lead Fast Reactor using simple engineered features, fully complying with Generation IV goals ELSY project ended on February 2010, but LFR development continued with the LEADER project (Lead-cooled European Advanced Demonstration Reactor) started on April 2010 in the frame of EC FP7 Steam generator: FA Support System B C Safety Vessel B DHR deep cooler Reactor Vessel Core C
LFR development in Europe The LEADER project Advanced nuclear systems for increased sustainability FP7-Fission 2009 2.2.1 Conceptual Design for Lead Cooled Fast Reactor Systems (LEADER) 16 European Organization are participating to the project Ansaldo Nucleare is the Project Coordinator 3 year Project (2010-2013), started April 1 st 2010 LEADER Objectives Deep analysis of the hard points of ELSY design to identify improvements with the goal to reach an European Lead Fast Reactor (ELFR) configuration Definition of a new frozen ELFR configuration to be used as a reference plant Conceptual design of a scaled down facility respect to the reference plant the Advanced Lead Fast Reactor European Demonstrator - ALFRED (300 MWth)
LEADER first step: Analysis of ELSY Critical Issues Steam generator: Manufacturability FA Support System: Floating core DHR deep cooler: Tank injection, dip cooler, Steam to chimney Containment boundary inside Reactor Vessel Inner Vessel/ Reactor Vessel bottom interface: No lateral restraints FA / Inner Vessel bottom interface: No lateral restraints
LFR Design Evolution - from ELSY to ELFR STRATEGY: Maintain the good solutions, change the rest SG manufacturability faced with a LEADER specific task Expected advantage of open square FA not verified, back to the wrapped option of ELSY, FA support provided, easy FA continuous monitoring for flow blockage Bottom grid introduced, lateral restraint for core and shroud, FAs weighted down by Tungsten ballast Need to develop alternative DHRs, ICS maintained.
ELFR Reactor Block Reactor block vertical section Top view
ELFR - Fuel Assembly weighted down whit Tungsten ballast Top Spring Spike Core Core Outlet Ballast Ballast Core Outlet Core Spike
ELFR - Fuel Assembly, with Tungsten ballast, connected to Lower and Top Core Plate
ELFR - Inner Vessel, Core support and Fuel Assembly Top core plate Pin Core diagrid
ELFR Control and Shutdown systems 2 independent passive shutdown systems gravity driven system (only shutdown) passively inserted by a pneumatic system (by depressurization) from the top of the core. In case of failure of the pneumatic system, the safety rods are equipped with tungsten ballast that forces the absorber down by gravity with a lower velocity control/shutdown system inserted from below the active core zone using the strong buoyancy available in lead. The inadvertent ejection from the core active part is avoided by means of a mechanical device in the core upper part to stop the control rod, once inserted, at the end of the active zone.
ELFR - Decay Heat Removal Systems strategy One non safety-grade system, the secondary system, is used for the normal decay heat removal following the reactor shutdown Two INDEPENDENT, DIVERSE, PASSIVE and REDUNDANT safety-grade Decay Heat Removal systems (DHR N1 and DHR N2): in case of unavailability of the secondary system the DHR N1 system is called and in the unlike event of unavailability of DHR1 the DHR N2 is used to evacuate the decay heat DHR N1: 4 Isolation Condenser Systems connected to 4 secondary systems DHR N2: 4 Water Decay Heat Removal System in the cold pool (dip cooler) Independence: Diversity: Redundancy: Passive: DHR Systems features two different systems with nothing in common two systems based on different physical principles three loops sufficient to remove decay heat Active actuation (but independent from grid), Passive operation
ELFR - Decay Heat Removal Systems DHR N1 ICS DHR N2 WDHRS dip cooler bundle continuous monitoring Loop closed using an HX inside a pool No release outside containment HX/Tank Water Helium Lead
ELFR Advantages and Safety characteristics Advantages of Lead and how they are used in the ELFR design Lead does not react with water or air Possibility to eliminate the intermediate loop; SGU installed inside the Reactor Vessel R&D needed on SGTR (no expected safety concern) No stringent requirements on reactor leak tightness Lead has very high boiling point and very low Partial Pressure Reduced core voiding risk (Lead boiling point is 1745 C, Pressure at 400 C 3. 10-5 Pa) Lead has a higher density than the oxide fuel No need for core catcher to face core melt (molten clad and fuel dispersion) No risk of re-criticality in case of core melt Lead is a low moderating medium and has low absorption cross-section. No compact Fuel Assemblies fuel rods p/d design not imposed by core designers Core pressure loss drastically reduced high natural circulation capability
ELFR only Advantages? Issues mainly operational High Lead melting point (~ 327 C) assure Lead T above 340 C Overcooling transient (secondary side) may cause Lead freezing Corrosion / erosion of structural materials due to lead environment. Slugging of primary coolant loop (lead-oxide accumulation) seismic risk due to large mass of lead in-service inspection of core support structures spent fuel management by remote handling Steam Generator Tube rupture inside the primary system Flow blockage and mitigation of core consequences
ELFR only Advantages? Issues countermeasures High Lead melting point (~ 327 C) assure Lead T above 340 C Active heating system, design and operating procedures Overcooling transient (secondary side) may cause Lead freezing FW requirement diversification and redundancy may be not a safety issue Corrosion / erosion of structural materials due to lead environment. Coatings, oxygen control, limit flow velocity Slugging of primary coolant loop (lead-oxide accumulation) Oxygen control or strategy at low oxygen content seismic risk due to large mass of lead 2-D seismic isolators, vessel hanged, specific design (EU FP7 SILER project) in-service inspection of core support structures Similar to other HLM reactors but high T, all components replaceable spent fuel management by remote handling Develop appropriate cooling system (probably active passive) Steam Generator Tube rupture inside the primary system Show no effect on core, provide top cover rupture disks to limit max pressure Flow blockage and mitigation of core consequences Hexagonal wrapped FAs outlet temperature continuous monitoring Full unprotected flow blockage causes cladding damages to a max of 7 FAs
ELFR SAFETY APPROACH safety is built-in-the-design rather than added on full implementation of the Defence in Depth principles Exhaustive: complete identification of initiating events Progressive: no major consequences from short sequences Tolerant: no cliff edge effects Forgiving: sufficient grace period and repair possible during accidental situations Well-balanced: no sequence contributes in an excessive way to damaged plant states risk-informed approach deterministic approach and probabilistic approach ISAM and in particular the Objective Provision Tree (OPT) adopted in the design process. for each level of DiD (normally level 1 to 5) and for each safety objective/function identification of: the possible challenges to the safety functions, the plausible mechanisms which can materialize these challenges, the provided provision(s) to prevent, control or mitigate the consequences
Summary of LFR safety analyses (DBA) Loss of Primary Flow (failure of one pump or loss of all primary pumps) Transient Over Power (at Hot full power and cold zero power conditions) Loss of Heat Sink Loss of Primary Flow + Heat Sink (Station Black-Out) Over-Cooling of Primary Side Loss of Primary Coolant Steam Generator Tube Rupture Loss of Primary Flow + Heat Sink without DHR systems available
Summary of LFR safety analyses (DBA) Results ACCIDENT Max Fuel temperature ( C) Loss of Primary Flow 2070 759 Transient Over Power (HFP) 2178 540 Transient Over Power (CZP) 619 405 Loss of Heat Sink 2062 550 Loss of Primary Flow + Heat Sink (Station Black-Out) 2062 664 Over-Cooling of Primary Side 2192 525 Loss of Primary Coolant 2062 553 Max Cladding temperature ( C) Loss of Primary Flow + Heat Sink without DHR systems available 2062 664
Summary of LFR safety analyses (DBA) Results The highest fuel temperature is reached during the Primary side Overcooling Transient after reactor trip this value suddenly decreases and there is no further concern for fuel melting. The highest clad temperature is observed in Loss of primary flow transient within the first seconds of transient (< 30 s) After 1 hour, all the temperatures are well below 500 C In general, there is no real safety concern for any of the DBC accidents. The system is very forgiving and even in worst conditions (concomitant Loss of Primary Flow and Heat sink, without DHR system availability), there is an extended margin in time The steam generator tube rupture event does not jeopardize the integrity of the plant and the reactivity effects related to core void formation are clearly enveloped by the current overpower transient analysis.
Summary of LFR safety analyses (DEC) There are only a few transients during which the core conditions could approach clad failure unprotected complete loss of all heat sinks with and/or without the operation of the primary pumps clad temperature under EOC core conditions will most likely exceed limits, leading to possible local cladding failure (peak pins). No whole core clad failure is however to be expected. Potential positive reactivity insertions are judged to be enveloped by +200 pcm insertion in 10 sec case for HFP and by +350 pcm insertion in 10 sec case for CZP conditions. The results of the calculations have shown that power spikes exceeding a factor 1.6 nominal power are buffered as reactivity feedback effects decrease the power quickly towards nominal conditions. In order to prevent freezing of the lead coolant, a strict control of the secondary coolant conditions (feedwater inlet temperature, feedwater flow rate) is needed. In general, the safety analysis demonstrates the extremely forgiving nature of the ELFR design, ascribable to the intrinsic properties of the coolant taken and used from the beginning in the design of the core and primary system
LFR RESPONSE to extreme natural events Response to earthquakes enhanced adoption of 2D seismic isolators (EU-FP7 SILER Project) Reactivity insertion due to core compaction anticipated adoption of wrapped hexagonal FAs, with lateral restrains. Two Decay Heat Removal Systems independent, redundant and diverse DHR systems completely passive. Only actuation (through valves alignment) is active, using local stored energy the Station Blackout does not represent a concern any initiating event is managed without AC power Even in case of Station Blackout without DHR systems available, the safety analyses demonstrated that the fuel and cladding temperatures do not represent a concern Complete core melt is practically impossible due to favorable intrinsic lead characteristics: high thermal inertia, very high boiling point, higher density with respect to oxide fuel with enhancement of the fuel dispersion respect to fuel compaction. In the very unlikely event of an extreme Fukushima like scenario (or beyond) leading to the loss of all heat sinks (both DHR and secondary systems), the heat can still be extracted injecting water in the reactor cavity between the reactor and safety vessels, while in case of reactor vessel breach the decay heat can still be removed by the same system that cools the concrete of cavity walls.
Conclusion In the framework of the LEADER project, the safety approach for a Leadcooled fast reactor has been defined and, in particular, all the possible challenges to the main safety functions and their mechanisms have been specified, in order to better define the needed provisions. On the basis of the above and taking into account the results of the safety analyses performed during previous project (ELSY), a reference configuration of the ELFR plant has been consolidated, by improving and updating the plant design features. In particular, the emerged safety concerns have been analyzed in the LEADER project and a new set of design options and safety provisions have been proposed. The combination of favourable Lead coolant inherent characteristics and plant design features, specifically developed to face identified challenges, resulted in a very robust and forgiving design, even in very extreme conditions, as a Fukushima-like scenario.
Thank you for your attention