Analyses of representative DEC events of the ETDR

1. LEADER WP5 MEETING, Petten – 26th of February 2013 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. KTH contribution • T-DEC1 – complete loss of forced flow + SCRAM fail • T-DEC4 – complete loss of forced flow, complete loss of SCS, DHR system operating

4. T-DEC1&4 RELAP5 model

5. T-DEC1 – Description • T-DEC1 – complete loss of forced flow + SCRAM fail • Pumps are tripped at 0s • Secondary side is operational, IC valves closed

6. T-DEC1 - loss of 8 pumps

7. T-DEC1 - loss of 8 pumps

8. T-DEC1 - loss of 8 pumps

9. T-DEC1 - loss of 8 pumps

10. T-DEC4 – Description • T-DEC4 – complete loss of forced flow + complete loss of secondary cooling system + SCRAM fail • Pumps and SCS are tripped at 0s, IC valves opened at 1s

11. T-DEC4 – loss of flow + loss of SCS + IC valves open

12. T-DEC4 – loss of flow + loss of SCS + IC valves open

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

14. T-DEC4 – loss of flow + loss of SCS + IC valves open

15. T-DEC1&4 RELAP5 model

16. T-DEC4 with 1 working pump

17. T-DEC4 with 1 working pump

18. T-DEC4 with 1 working pump

19. KTH contribution • TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction)

20. TR-4 – Description • TR-4 – a transient event due to reactivity insertion (enveloping SGTR, flow blockage, core compaction) 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)

21. ALFRED design • ALFRED – Advanced Lead Fast Reactor European Demonstrator • Power – 125 MWel (300 MWth, ~41% efficiency) • 8 steam generators with 8 axial pumps • 2 independent DHRs using SG tubes

22. Motivation • Core – submerged in the bottom middle part of the pool • 8 SGs – located inside the primary circuit radially around the core • One co-axial pump per SG • High P secondary vs low P primary system • SG tube rupture and leakage events in PWRs suggest that it can be a concern in LFRs • Risk of core voiding in case of SGTL due to: • Proximity of SGs to core • Nature (shape) of the flow path • Leak-Before-Break (LBB) (< liter/day in PWRs) • SGTL can hinder licensing due to: • Potential severe consequences (core damage) • Lack of operational experience (uncertainty in frequency of SGTL)

23. Objectives • To assess the risk related to SGTL • To identify the scenarios of core voiding • To quantify the likelyhood that a steam bubble is transported from a SG to the core • Identify the uncertainties in SGTL • Quantifiy these uncertainties • To estimate void accumulation rates in the core • To estimate the consequent effect to power (or to neutron flux) with a neutronic code

24. Scenarios • SGT leakage is considered during normal operation • If steam bubbles are dragged to the core then there are 3 distinct scenarios of safety concern: • Homogeneous voiding of the coolant (continuous leak, small bubbles) • Bubbles stuck in spacers (mid-size bubbles) • Slugs of void entering the core (formed in stagnation zones) Depending on the scenario... • RIA • Local damage (burn-out) of the fuel • Overpressurization of the vessel ...can happen.

25. Uncertainties • Aleatory and epistemic uncertainties assessed in the SGTL scenarios and phenomenology: • Crack size and morphology • Bubble size distribution • Leak rate through the crack • Bubble drag correlation • Bubble size distribution • Leak rate depends on crack size, morphology and pressure differences between two sides • ”Leak-before-break” K. Terasaka et al. (2011) A. V. Beznosov et al. (2005)

26. Drag coefficient Non-linear drag coefficientm, , as a function of bubble diameter Validation is done by comparing with predictions from Stokes and Mendelsons law Bubbles were modeled: • As Lagrangian particles • Constant density • Drag coefficient • With/without turbulent dispersion Tomiyama et al. correlation chosen:

27. Approach to estimate core voiding Bubbles are injected at - 3 different height levels in SG (bottom, middle, top) - at the exit of the core - at the pump outlet to estimate probabilities: P1 – that bubble enters core P2 – that bubble proceeds to pump P3 – that bubble stays in the primary loop Steam accumulation rate in the primary system: and in the core (assumed bubbles get stuck there): Cold free surface Hot free surface 1 3 P2 2 P3 P1

28. ALFRED CAD model

29. ALFRED CFD model • 45° CFD model is being created • Simplified modeling of complex (SG, lower/upper grids) components (porous media, momentum source etc) • Consists of 7 regions a • Downcomer • Lower grid • Lower inactive region • Active region • Upper inactive region • Pump channel • Steam generator

30. Example of ELSY modeling SG Pump Down- comer Core

31. Example of ELSY results dbubble=0.4 mm dbubble=0.2 mm dbubble=0.5 mm dbubble=1.0 mm Very small bubbles are dragged to core (<0.4mm), whereas middle size ones are not

32. Summary • Steam generator tube leakage accident is addressed • Motivation, scenarios, uncertainties • Nominal operational primary flow conditions will be modeled with a CFD code Star-CCM+ • Neutronics part of the analysis will 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

33. Thank you!