1 Kaspar Kööp, Marti Jeltsov Division of Nuclear Power Safety Royal Institute of Technology (KTH) Stockholm, Sweden LEADER 4 th WP5 MEETING, Karlsruhe – 22 nd of November 2012 Analyses of representative DEC events of the ETDR

2 ETDR – ALFRED description Pool-type 300 MWth Core pressure drop 1 bar Temperature –Core inlet 400 C –Core outlet 480 C Coolant velocity –Average 2 m/s –Maximum 3 m/s Lead void effect at EOC (only the fuel zones) –+2 $

3 KTH contribution Transients to be analyzed for Pb-cooled ALFRED Design (LEADER project) Case Number TransientDescription Burnup State Transients analyzed for Lb-cooled ALFRED Design BOCEOC ENEAKIT-GNRGJRC/IETKTH RELAP5SIM-LFRSPECTRA SIMMER / TRACE RELAP5 / CFD code DEC Transients TR-4 Reactivity insertion (enveloping SGTR, flow blockage, core compaction) Reactivity insertion (voiding of part of active region enveloping voids introduction due SGTR, core compaction, fuel blockage) = 250 pcm Reactor at hot full power (HFP) XXXXXX (*)X (**) TO-3 Reduction of FW temperature + all pumps stop Loss of one preheater (feedwater temperature reduction from 335oC down to 300oC) All primary pumps are stopped Reactor is tripped XXXXX TO-6 Increase of FW flowrate+ all pumps stop 20 % increase in feedwater flowrate All primary pumps are stopped Reactor is tripped XXXXX T-DEC1 Complete loss of forced flow + Reactor trip fails (total ULOF) All primary pumps are stopped Feedwater system available Reactor trip fails XXXXXX (*)X T-DEC3 Loss of SCS+ Reactor trip fails (ULOHS) All primary pumps are operating DHR system is operating Reactor trip fails XXXXXX (*) T-DEC4 Loss of off- site power (LOOP) + Reactor trip fails (ULOHS + ULOF) All primary pumps are stopped SCS is lost DHR system is operating Reactor trip fails XXXXX X T-DEC5 Partial blockage in the hottest fuel assembly Reactor trip fails The maximum acceptable flow reduction factor has to be determined XXXXXX T-DEC6SCS failure All primary pumps are operating DHR system totally fails Reactor is tripped XXXXX TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction) T-DEC1 – complete loss of forced flow + SCRAM fail T-DEC4 – complete loss of forced flow, complete loss of SCS, DHR system operating + SCRAM fail

4 TR-4 – Description TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction) We have shown in our previous works that using system TH codes it is not possible "...to investigate whether the steam bubble or bubbles can be dragged downwards towards the core inlet region.“ Steam Generator Tube Leakage (SGTL) is assumed to be the cause for reactivity insertion (voiding of part of active region We address the task on the transport of bubbles that have leaked in the SG to the primary coolant flow Reactor is at hot full power (HFP) Actions: –First thermal-hydraulic part (CFD analysis of bubble transport) –Neutronic part (SERPENT code to look at the consequences of different local core voiding that are typical for SG leakage)

5 TR-4 – Thermal-hydraulics approach Approach: –Develop (or ask from partners) 3D CAD model of primary system of ALFRED according to the latest design provided to LEADER partners. –Create 3D mesh of the primary system for CFD analysis. –Simulate primary coolant flow at normal (HFP) operation conditions with a 3D CFD code (Star-CCM+) –Simulate bubble transport from the SG to the core –Assumptions in modeling of bubble transport: Lagrangian framework Turbulent dispersion Uncertainty in: –bubble size distribution –different correlations for bubble drag in lead –locations of possible leakage from steam generator –leak rate –voiding scenarios –etc. –Assess void accumulation rate in the core accounting for the uncertainties given

6 TR-4 – Neutronics approach Neutronics part of the analysis is foreseen to be done using Serpent Monte Carlo code Input for neutronic calculation –void characteristics: accumulation rates voiding scenarios are input for neutronics calculation –geometry ALFRED model exists in the house

7 T-DEC1&4 ENEA’s RELAP5 model

8 Model steady state ParameterBy designPrevious inputNew input ( ) Nominal reactor power 300 MWth MWth MWth Number of SGs 8 Number of PPs 8 Power removed per SG 37.5 MWth MWth MWth Lead (primary system) Core inlet temperature 400 C K ( C) K ( C) Core outlet temperature 480 C K ( C) K ( C) Mass flow rate *8 kg/s25013 kg/s ( * 8) kg/s ( * 8) Secondary side Pressure 180 bar Imposed 188 bar feed water in Imposed 180 bar steam out Water inlet temperature 335 C K ( C) Steam outlet temperature 450 C k ( C) K ( C) Mass flow rate 24.1 kg/s/SG24.08 kg/s/SG Pressure drop over core0.98 bar

9 T-DEC1 – Description T-DEC1 – complete loss of forced flow + SCRAM fail Pumps are tripped at 1500s Secondary side is operational, IC valves closed

10 T-DEC1 - loss of 8 pumps

11 T-DEC1 - loss of 7 pumps

12 T-DEC4 – Description T-DEC4 – complete loss of forced flow + complete loss of secondary cooling system + SCRAM fail Pumps are tripped at 1500s SCS is tripped at 1500s IC valves opened at s

13 T-DEC4 – loss of flow + loss of SCS + IC valves open ?

14 Next steps Check T-DEC4 results Combine T-DEC1 and T-DEC4 –Only some pumps fail –Only some IC valves open –Possibility of pump/valve recovery Look for –Overcooling/overheating scenarios –High local velocity scenarios –…

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16 SGTR