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Analyses of representative DEC events of the ETDR

LEADER 4 th WP5 MEETING, Karlsruhe – 22 nd of November 2012. Analyses of representative DEC events of the ETDR. Kaspar Kööp, Marti Jeltsov Division of Nuclear Power Safety Royal Institute of Technology (KTH) Stockholm, Sweden. ETDR – ALFRED description. Pool-type 300 MWth

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Analyses of representative DEC events of the ETDR

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

  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. 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 KTH contribution

  4. 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) TR-4 – Description

  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

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

  10. T-DEC1 - loss of 8 pumps

  11. T-DEC1 - loss of 7 pumps

  12. 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 1500.5s T-DEC4 – Description

  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 • …

  15. SGTR

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