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KIT – University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association Institute for Neutron Physics and Reactor.

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Presentation on theme: "KIT – University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association Institute for Neutron Physics and Reactor."— Presentation transcript:

1 KIT – University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association Institute for Neutron Physics and Reactor Technology www.kit.edu Safety analysis for ELFR and ALFRED E. Bubelis, M. Schikorr (KIT-G), G. Bandini (ENEA)

2 Institute for Neutron Physics and Reactor Technology (INR) 22012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Table of contents  Introduction  Safety analysis carried out for ELSY reactor. Main conclusions from safety analysis for ELSY  Safety analysis for ELFR. Preliminary results  Safety analysis for ALFRED

3 Institute for Neutron Physics and Reactor Technology (INR) 32012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Introduction EU FP-7 LEADER project is a continuation of the previous EU FP-6 ELSY project. Detailed safety analysis for ELSY reactor was carried out at that time within EU FP-6 ELSY project. Conclusions and recommendations from ELSY project were taken into account while designing ELFR and ALFRED reactors within the EU FP-7 LEADER project. The main aim of the LEADER project is to find improvements for all the sensitive issues raised during the ELSY project, in particular for the desings of the industrial ELFR reactor and the smaller prototype ALFRED reactor (LFR technology demonstrator).

4 Institute for Neutron Physics and Reactor Technology (INR) 42012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis carried out for ELSY reactor. Main conclusions from safety analysis for ELSY

5 Institute for Neutron Physics and Reactor Technology (INR) 52012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis carried out for ELSY reactor – DBC (Design Basis Condition) transients (denoted by P -..)

6 Institute for Neutron Physics and Reactor Technology (INR) 62012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis carried out for ELSY reactor – DEC (Design Extension Condition) transients (denoted by U -..)

7 Institute for Neutron Physics and Reactor Technology (INR) 72012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis carried out for ELSY reactor - DEC

8 Institute for Neutron Physics and Reactor Technology (INR) 82012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Conclusions from safety analysis for ELSY As related to the DBC (Design Basis Condition) accidents, the transients examined proved that the ELSY plant can withstand without problems a rather wide range of accidental events. With the exception of P-5b (PLOF+PLOH with no DHR systems available), ELSY plant has proved to be able to enter a safe shutdown phase after every DBC accident analyzed. ELSY system is very forgiving, and even under worst conditions (PLOF+PLOH with no DHR), there is an extended time margin for a possible operator intervention, preventing further damages.

9 Institute for Neutron Physics and Reactor Technology (INR) 92012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Conclusions from safety analysis for ELSY As related to the DEC (Design Extension Condition) accidents, the following issues were raised: 1) The transient of concern: the unprotected operation with complete loss of all heat sinks with and/or without the operation of the primary pumps. 2) Power spikes (during UTOP) exceeding a factor 1.6 nominal power are largely buffered as reactivity feedback effects decrease the power quickly towards nominal conditions. Only under extreme conditions limited, local clad failures are to be expected. 3) In the case of a large area inlet blockage of the open FA structure, failure of fuel cladding should be anticipated.

10 Institute for Neutron Physics and Reactor Technology (INR) 102012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Conclusions from safety analysis for ELSY 4) Another potential safety issue is the maximum reactor vessel wall temperature in case of total loss of secondary circuits (ULOH, ULOF+ULOH and large secondary break transients) that might exceed 700 °C within half an hour during such an event. 5) A potential concern identified was the very tight operational control requirements to be imposed on the secondary coolant conditions in order to assure prevention of freezing of the lead coolant at the coldest location of the primary loop, namely at the outlet of the primary side of the main heat exchanger.

11 Institute for Neutron Physics and Reactor Technology (INR) 112012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis for ELFR: Preliminary results

12 Institute for Neutron Physics and Reactor Technology (INR) 122012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ELFR - Reactor Configuration Ref.: L. Mansani, Ansaldo Nucleare

13 Institute for Neutron Physics and Reactor Technology (INR) 132012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ELFR - design options ItemOption Electrical Power (MWe)600 Primary CoolantPure Lead Primary SystemPool type, Compact Primary Coolant Circulation: Normal operation/ Emergency conditionsForced/ Natural Allowed maximum Lead velocity (m/s)2 Core Inlet Temperature (°C)400 Steam Generator Inlet Temperature (°C)480 Secondary Coolant CycleWater-Superheated Steam Feed-water Temperature (°C)335 Steam Pressure (MPa)18 Secondary system efficiency (%)  43 Reactor vesselAustenitic SS, Hung Safety VesselAnchored to reactor pit Inner Vessel (Core Barrel)Cylindrical, Integral with the core support grid, Removable Ref.: L. Mansani, Ansaldo Nucleare

14 Institute for Neutron Physics and Reactor Technology (INR) 142012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ELFR - design options ItemOption Steam generators Integrated in the reactor vessel and removable. ELSY Spiral tubes have been exploited. Helical or Bayonet tubes types are alternative options Primary pumpsMechanical pumps in the hot collector, Removable Fuel AssemblyClosed (with wrapper), Hexagonal, Weighed down when primary pumps are off, Forced in position by springs when primary pumps are on Fuel typeMOX Maximum discharged burnup, (MWd/kg-HM)100 Fuel Cycle, (y)2 Fuel resident time, (y)5 Fuel Clad MaterialT91 (coated) Maximum Clad Temperature in Normal Operation, (°C) 550 Maximum core pressure drop, (MPa)0.1 Ref.: L. Mansani, Ansaldo Nucleare

15 Institute for Neutron Physics and Reactor Technology (INR) 152012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ELFR - design options ItemOption Control/Shutdown System2 diverse and redundant systems of the same concept derived from CDT. 1 st System for ShutdownPneumatic Inserted Absorber Rods: shutdown system passively inserted by pneumatic (by depressurization) from the top of core. In case of loss of this system, a tungsten ballast will force the absorber down by gravity 2 nd System for Control and ShutdownBuoyancy Absorbers Rods: control/shutdown system passively inserted by buoyancy from the bottom of the core. Refuelling SystemNo refuelling machine stored inside the Reactor Vessel DHR System2 diverse and redundant systems (actively actuated, passively operated) DHR-14 Isolation Condenser systems (ICs) connected to 4 Steam Generators (SGs) DHR-24 Isolation Condenser systems (ICs) connected to 4 Dip Coolers (DCs) Seismic Dumping Devices2D isolator below reactor building Ref.: L. Mansani, Ansaldo Nucleare

16 Institute for Neutron Physics and Reactor Technology (INR) 162012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ELFR

17 Institute for Neutron Physics and Reactor Technology (INR) 172012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ELFR

18 Institute for Neutron Physics and Reactor Technology (INR) 182012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ELFR

19 Institute for Neutron Physics and Reactor Technology (INR) 192012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-1 : PLOF (BOC) PP trip at t = 0 s; One DHR-1 sub-system fails; Power removed by DHR-1 : 22.5 MW; Reactor scram at t = 3 s; Power decreases to decay heat level; Pb flowrate decreases to 21% at ~ 6 s, later stabilizes ~ 8-9 % nominal flow at ~ 120 s.

20 Institute for Neutron Physics and Reactor Technology (INR) 202012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-2 : ULOF (EOC) PP trip at t = 0 s. No reactor scram; Secondary cooling system fully functional (forced convection); Pb flowrate decreases to 21% at ~ 6 s, later stabilizes ~ 27-28 % nominal flow at ~ 60 s. Power decreases to ~78% Pnom; ULOF can be accommodated.

21 Institute for Neutron Physics and Reactor Technology (INR) 212012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-3 : ULOHS (BOC) Heat transport path through the MHX’s abruptly decreases to 0 % at t = 0 s; No reactor scram. One DHR-1 sub- system fails; Power removed by DHR-1 : 22.5 MW; Twall approaches ~700 o C at t ~700 s; Long term structural integrity of reactor vessel not guaranteed; Tclad reaches 1033 o C at t ~28 min and clad ruptures, releasing FPs into Pb.

22 Institute for Neutron Physics and Reactor Technology (INR) 222012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-4 : UTOP (EOC) 260 pcm reactivity insertion in 10 s; Power peak ~2.42 Pnom; Maximum Tfuel reaches 2677 o C at the center of fuel pin and starts melting locally; Maximum Tclad reaches 719 o C, but does not rupture; Max possible react. insertion of 260 pcm in 10 sec will not lead to core melting at HFP.

23 Institute for Neutron Physics and Reactor Technology (INR) 232012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-5 : ULOF & ULOHS (EOC) PP trip & heat transport path through the MHX’s abruptly decreases to 0 % at t = 0 s. No reactor scram; One DHR-1 sub-system fails; Power removed by DHR-1: 22.5 MW; Tclad reaches ~880 o C at t ~12 min; min clad failure time ~3.5 min. Twall approaches ~700 o C at t ~1.5 h; Long term structural integrity of reactor vessel not guaranteed.

24 Institute for Neutron Physics and Reactor Technology (INR) 242012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-6 : OVC (EOC) Tfeedw drops to 300 o C in 1 s; Reactor scram at t = 2 s; Power removed by DHR-1: 30 MW; PP run normally – 100 % prim flow. Power decreases to decay heat level; At t ~ 0.63 h Pb temp at MHX outlet and core inlet reaches ~355 o C and stabilizes, being still high enough above Tfreezing for Pb.

25 Institute for Neutron Physics and Reactor Technology (INR) 252012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-6 : OVC; DHR-1 system schematics

26 Institute for Neutron Physics and Reactor Technology (INR) 262012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-7 : SLB (preliminary ENEA results, G. Bandini) All SG secondary lines are depressurized at t = 0 s; Reactor scram at t = 3 s on low secondary pressure signal; Immediate start-up of DHR-2 (4 W-DHR loops); No risk of LBE freezing at SG outlet in the initial phase of the transient.

27 Institute for Neutron Physics and Reactor Technology (INR) 272012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-7 : SLB (preliminary ENEA results, G. Bandini) The DHR-2 system (4 W-DHR loops) efficiently removes the decay power after t = 2000 s; The risk of LBE freezing at W-DHR outlet is attained after about 5 hours.

28 Institute for Neutron Physics and Reactor Technology (INR) 282012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop T-8 : SA blockage (EOC) For SA blockages of less then 75% blockage area, the critical ELFR is not expected to experience any pin failures even under unprotected conditions. Clad rupture of the peak power pins should be expected for SA blockages above ~ 75% blockage area. Fuel melting is not a problem for the critical ELFR. Fuel melting temperatures are not reached even in 97.5% SA blockage case.

29 Institute for Neutron Physics and Reactor Technology (INR) 292012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Preliminary conclusions from ELFR safety analysis 1) The transient of concern: the unprotected operation with complete loss of all heat sinks (ULOHS). In case of ULOHS transients clad temperature under BOC core conditions in ~ 0.5 h will exceed limits, leading to possible local cladding rupture (peak power pins). No whole core clad failure is however expected. 2) For reactivity insertion of 200 pcm in 10 sec time interval at EOC conditions, critical ELFR reactor peak fuel pin cladding survives this transient, fuel melting is not observed even in the center of the peak power fuel pins (pellets); For reactivity insertion of 260 pcm in 10 sec time interval (maximum possible reactivity insertion in 10 sec time interval), no core melting at HFP conditions at EOC is being observed. Critical ELFR reactor peak power fuel pin cladding survives this transient, however local fuel melting should be expected in the center of the peak power fuel pins (pellets).

30 Institute for Neutron Physics and Reactor Technology (INR) 302012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Preliminary conclusions from ELFR safety analysis 3) For SA blockages of less then 75% blockage area, the critical ELFR is not expected to experience any pin failures even under unprotected conditions. Clad rupture of the peak power pins should be expected for SA blockages above ~ 75% blockage area. Fuel melting is not a problem for the critical ELFR. Fuel melting temperatures are not reached even in 97.5% SA blockage case. 4) Another potential safety issue is the maximum reactor vessel wall temperature in case of total loss of secondary circuits (ULOHS, ULOF+ULOHS and large secondary break transients) that might exceed 700 °C within ~12 min during such an extremely unlikely transient event, thus presenting a concern of the mechanical integrity of the reactor vessel itself even before critical core (cladding) temperatures are reached.

31 Institute for Neutron Physics and Reactor Technology (INR) 312012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Preliminary conclusions from ELFR safety analysis 5) Important issue for the critical ELFR is a tight and continuous operational control of the secondary coolant conditions (feedwater inlet temperature, feedwater flowrate) in order to assure prevention of freezing of the lead coolant at the coldest location of the primary loop, namely at the outlet of the primary side of the main heat exchanger. Any sudden decrease in feedwater temperature (either by loss of the secondary water preheater(s), or any other cold water injection paths), or inadvertent increase in feedwater flowrate under partial load condition could lead to the rapid decrease of the primary HX outlet lead temperature to close to the freezing point of Pb. In general: The safety analysis performed for the Pb-cooled ELFR design demonstrated the very forgiving nature of this plant design when compared to other similar GEN-IV plant designs, ascribable to the inherently large thermal inertia of the Pb-cooled primary system, the very high boiling temperature of lead coolant, and the optimization of safety relevant control, safety systems and components.

32 Institute for Neutron Physics and Reactor Technology (INR) 322012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Safety analysis for ALFRED

33 Institute for Neutron Physics and Reactor Technology (INR) 332012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ALFRED - Reactor Configuration Ref.: L. Mansani, Ansaldo Nucleare

34 Institute for Neutron Physics and Reactor Technology (INR) 342012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ALFRED - design parameters Ref.: ALFRED database CORE PARAMETERS Thermal power (reference for range)300 MW th Maximum pressure drop (in core)0.1 MPa (= 1 bar) Maximum inner vessel radius~150 cm FA Layout (open/closed or square/hexagonal) Closed hexagonal FUEL Type of fuel (including stoichiometric ratio) MOX (Pu,U)O 1.97 Max Pu enrich: 30% Thermal conductivityRef. [3] Max fuel temperature in nominal conditions (°C) ~2000 °C Peak burn-up~100 GWd/tHM Maximum plenum pressure~5 MPa Clad thickness~0.6 mm Gap~0.15 mm

35 Institute for Neutron Physics and Reactor Technology (INR) 352012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop ALFRED - design parameters Ref.: ALFRED database STRUCTURAL MATERIALS Cladding and grids material15-15Ti (coated) Wrapper materialT91 Max temperature of cladding in nominal conditions ~550 °C Max DpA 80-100 (clad) 2 (diagrid, RV and Internals) in 40y COOLANT Inlet temperature400 °C Outlet temperature480 °C Avg velocity (m/s) Max local velocity < 2 m/s < 3 m/s Clad/bulk heat transfer correlationRef. [4] CONTROL RODS MaterialB 4 C (90 at.% 10 B) Cold shutdown temperature SCRAM margin ~370 °C 1000 pcm

36 Institute for Neutron Physics and Reactor Technology (INR) 362012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ALFRED

37 Institute for Neutron Physics and Reactor Technology (INR) 372012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ALFRED

38 Institute for Neutron Physics and Reactor Technology (INR) 382012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ALFRED

39 Institute for Neutron Physics and Reactor Technology (INR) 392012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Transients list for ALFRED

40 Institute for Neutron Physics and Reactor Technology (INR) 402012.09.05. E. Bubelis, et.al. – 3rd LEADER Int. Workshop Progress with safety analysis for ALFRED A total of 20 transients, considered the most challenging for the ALFRED plant, have been selected and will be analyzed by Tasks 5.4 and 5.5 of the LEADER project. The safety analysis activities for ALFRED are starting in September 2012 and the results of the activities are foreseen to be available by the end of 2012. However, some preliminary transient analysis results for ALFRED will be presented by G. Bandini on September 7, 2012 in his presentation “Accident analysis overview”.

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