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 LEADER: WP5 – Safety and Transient Analysis Overview and status of WP5 activities as of November 2012 E. Bubelis, M. Schikorr, W. Hering (KIT)

2. Institute for Neutron Physics and Reactor Technology (INR) 22012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WPs

3. Institute for Neutron Physics and Reactor Technology (INR) 32012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 – Time Table & Milestones Milestones M05/D11 (WPL & T5.3, PSI) - Identification of representative DBC and DEC accident initiators for the ETDR (May 2012) M07/D15 (WPL / T5.4, KIT-G) - Analysis of DBC events for the ETDR (September 2012) M10/D22 (WPL & T5.5, ENEA) - Analysis of DEC events for the ETDR (March 2013) - Milestones; - Draft reports ; - Final reports ; - WP5 meetings. WP5 meetings M01 - LFR updated reference configuration – 2012.03.06. (TEC 068-2011) ; M04 - ETDR reference configuration – 2012.05.21. (TEC 058-2012).

4. Institute for Neutron Physics and Reactor Technology (INR) 42012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 – Deliverables & Technical documents Deliverables D04 (T5.1, ANSALDO) - Safety approach for LFR plants (February 2011) D08 (T5.2, KIT-G) - Report on the results of analysis of DBC & DEC key transients for the LFR reference plant (October 2012) D11 (T5.3, PSI) - Identification of representative DBC and DEC accident initiators for the ETDR (May 2012) D15 (T5.4, KIT-G) - Report on the results of analysis of DBC events for the ETDR (September 2012) D18 (T5.6, EA) - Report on Containment Assessment for the ETDR (Rev. 0 – November 2012, Rev. 1 – March 2013) D22 (T5.5, ENEA) - Report on the results of analysis of DEC events for the ETDR (March 2013) Technical Documents T68 (T5.2, KIT-G) - Plant data for the safety analysis of the industrial LFR (March 2012) T58 (T5.3, PSI) - Plant data for the safety analysis of the ETDR (May 2012) T64 (T5.6, EA) - Report on Source Term Assessment for the ETDR (Rev. 0 – November 2012, Rev. 1 - March 2013)

5. Institute for Neutron Physics and Reactor Technology (INR) 52012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 - Tasks Task 5.1 (Monica Frogheri (Monica.Frogheri@ann.ansaldo.it)) Global safety approach re-assessment (ANSALDO, KIT-G, SCKCEN) Task 5.2 (Evaldas Bubelis (Evaldas.Bubelis@kit.edu)) Re-evaluation of consequences of representative accident (key transient) initiators within the DBC and DEC for the LFR reference plant (KIT-G, CIRTEN, ENEA, PSI) Task 5.3 (Konstantin Mikityuk (konstantin.mikityuk@psi.ch)) Identification of representative DBC and DEC accident initiators of the ETDR (PSI, ANSALDO, CEA, ENEA, KIT-G, SCKCEN) Task 5.4 (Evaldas Bubelis (Evaldas.Bubelis@kit.edu)) Analyses of representative DBC events of the ETDR (KIT-G, CEA, CIRTEN, ENEA, JRC-IE, KTH, PSI) Task 5.5 (Giacomino Bandini (giacomino.bandini@enea.it)) Analyses of representative DEC events of the ETDR (ENEA, KIT-G, JRC-IE, KTH, NRG) Task 5.6 (Angela Cortes (acm@empre.es)) Containment and source term assessment for the ETDR (EA, KTH)

6. Institute for Neutron Physics and Reactor Technology (INR) 62012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The objective of this task is the re-evaluation of the global safety approach for the LFR reference plant and the development of the safety strategy and approach for the ETDR. The specification of the operational and safety limits, as well as the acceptance criteria for the safety performance in order to facilitate the licensing of the ETDR plant will be developed. Safety system recommendations of the GEN IV Risk and Safety Working Group (RSWG) need to be considered appropriately and differences to other projects as the CP-ESFR for the sodium cooled system should be clearly identified. Task 5.1 Global safety approach re-assessment (ANSALDO, KIT-G, SCKCEN) D04 (T5.1, ANSALDO) - Safety approach for LFR plants (February 2011)

7. Institute for Neutron Physics and Reactor Technology (INR) 72012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The main objective of this task is to re-evaluate the consequences of accident (key transient) initiators within the DBC and DEC for the updated LFR reference configuration, as defined by WP3, in order to confirm the safety performance of the plant and to resolve previously identified potential safety issues. Task 5.2 Re-evaluation of consequences of representative accident (key transient) initiators within the DBC and DEC for the LFR reference plant (KIT-G, CIRTEN, ENEA, PSI) D08 (T5.2, KIT-G) - Report on the results of analysis of DBC & DEC key transients for the LFR reference plant (October 2012)

8. Institute for Neutron Physics and Reactor Technology (INR) 82012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.2. DBC & DEC transients analyzed for ELFR T-1 : PLOF, DHR-1 or DHR-2 available, reactor trip T-2 : ULOF, SCS in forced convection T-3 : ULOHS, PPs active, DHR-1 or DHR-2 available T-4 : UTOP, study max possible reactivity insertion w/o core melting T-5 : ULOF+ULOHS, DHR-1 or DHR-2 available T-6 : OVC, FW temp drop from 335 oC to 200 oC in 1 sec, reactor trip T-7 : SLB, reactor trip T-8 : SA blockage, determine max acceptable SA flow reduction factor T-9 : SGTR (limited scope, low priority, based on experiments)

9. Institute for Neutron Physics and Reactor Technology (INR) 92012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.2. Conclusions (1) Unprotected transients (ULOF; ULOHS and ULOF + ULOHS): Due to the enhanced natural convection capability in the primary circuit, in case of ULOF the maximum temperatures reached in the primary system are low enough to assure the integrity of the clad and the vessel in the short term, providing sufficient grace time for corrective operator action. The main potential safety issue is the maximum reactor vessel wall temperature that might exceed 700 °C within ~12 min. The integrity of the clad and the vessel seems not guaranteed in the medium/long term, because of the high temperatures reached in the primary system. An optimization of the neutronic core design, in order to reduce the positive coolant expansion reactivity feedback could provide additional grace time.

10. Institute for Neutron Physics and Reactor Technology (INR) 102012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Reactivity insertion: These transients envelope positive reactivity insertions of the Design Basis events such as fuel handling errors, control rods withdrawal or seismic core compaction. For reactivity insertion of 200 pcm in 10 sec time interval at EOC conditions, peak fuel pin cladding survives and fuel melting is not observed, even in the center of the peak fuel pins (pellets). For reactivity insertion of 260 pcm in 10 sec time interval at EOC conditions, peak fuel pin cladding survives, however fuel melting should be expected in the center of the peak fuel pins (pellets). LEADER WP5 Task 5.2. Conclusions (2)

11. Institute for Neutron Physics and Reactor Technology (INR) 112012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe FA flow blockage: For flow blockages of < 75%, no pin failures nor fuel melting is expected, even under unprotected conditions. For flow blockages > 75%, peak power pins clad failure shall be expected, but fuel melting is not expected even for blockage over 97.5%. However there is time (several hundreds seconds) to detect the flow blockage occurrence, by means of temperature measuring devices installed at each FA outlet. LEADER WP5 Task 5.2. Conclusions (3)

12. Institute for Neutron Physics and Reactor Technology (INR) 122012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe SGTR accident: Several limiting mechanisms and potentially important effects have been analyzed and suggest that: (i)the initial pressure shock wave poses no likely threat to in-vessel structures, except to few adjacent heat-exchange tubes; (ii)the sloshing-related fluid motion is well bounded in a domain beyond the heat exchanger; and yet (iii)the steam/water entrainment is expected to be comparatively limited due to the very large difference of density between steam and lead. The potential gradual pressurization of the vessel after SGTR due to inflow of the steam is limited by rupture disks to relief the resulting over-pressure. Moreover, a Venturi nozzle placed inside each spiral tube, mitigate the severity of SGTR interaction and reduce the potential effects on the entire reactor system. A dedicated scaled facility should be foreseen to experimentally analyze in depth the SGTR phenomena further as part of the future R&D activities. LEADER WP5 Task 5.2. Conclusions (4)

13. Institute for Neutron Physics and Reactor Technology (INR) 132012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe General: The safety analysis performed for the lead-cooled ELFR design demonstrated the forgiving nature of this plant design when compared to other similar plant designs, ascribable to the inherently, large thermal inertia of the lead-cooled primary system and optimization of safety relevant control, safety systems and components. LEADER WP5 Task 5.2. Conclusions (5)

14. Institute for Neutron Physics and Reactor Technology (INR) 142012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The objective of this task is to identify representative accident initiators for Design Basis Conditions (DBC), Design Extension Conditions (DEC) including complex sequences, severe accidents and limiting events. In addition initiating events or accident sequences allocated to the “practically excluded” event category should be identified. On the basis of the design solution for the ETDR a simplified line-of-defense strategy will be applied to identify accident initiators. The identified event initiators will be categorized and the more representative for each category will be selected for safety analysis to be carried out in task 5.4 and 5.5. Impact of specific design solution especially for the heat conversion system will be taken into account appropriately. Task 5.3 Identification of representative DBC and DEC accident initiators of the ETDR (PSI, ANSALDO, CEA, ENEA, KIT-G, SCKCEN)

15. Institute for Neutron Physics and Reactor Technology (INR) 152012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Task 5.3 Identification of representative DBC and DEC accident initiators of the ETDR (PSI, ANSALDO, CEA, ENEA, KIT-G, SCKCEN) The objectives of this task are 1)to identify accident initiating events; 2)to categorize them (DBC, DEC, practically excluded) and 3)to select representative events from each category for safety analysis (for Tasks 5.4 and 5.5). M05/D11 (WPL & T5.3, PSI) - Identification of representative DBC and DEC accident initiators for the ETDR (May 2012) D11 (T5.3, PSI) - Identification of representative DBC and DEC accident initiators for the ETDR (May 2012) T58 (T5.3, PSI) - Plant data for the safety analysis of the ETDR (May 2012)

16. Institute for Neutron Physics and Reactor Technology (INR) 162012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The main objective of this task is to evaluate the consequences of representative accident initiators within the DBC. Specific attention will be given to the consequence evaluation of partial and/or total blockages in a fuel assembly and to consequences of steam generator tube ruptures (SGTR). An input data set representing the plant in sufficient detail will be established on the basis of a specific deliverable developed in strict collaboration with WP3. Several suitable computer codes (i.e. SIM-LFR, SAS-LFR, RELAP, TRACE, CFX) available among the partners will be used to analyze transients behavior within design basis conditions (DBC) including those requiring 3-dimensional neutron kinetics. The DBC accidents to be analyzed will be specified within Task 5.3. Consequences of SGTR will be analyzed with special attention. The SGTR could potentially lead to a rapid pressurization and shock waves in the heat exchanger (HX). By analyzing a section of a full bundle HX it will be possible to evaluate the generated pressure waves. The other important aspect is to investigate whether the steam bubble or bubbles can be dragged downwards towards the core inlet region. Task 5.4 Analyses of representative DBC events of the ETDR (KIT-G, CEA, CIRTEN, ENEA, JRC-IE, KTH, PSI) D15 (T5.4, KIT-G) - Report on the results of analysis of DBC events for the ETDR (Sept 2012)

17. Institute for Neutron Physics and Reactor Technology (INR) 172012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.4.

18. Institute for Neutron Physics and Reactor Technology (INR) 182012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.4. X

19. Institute for Neutron Physics and Reactor Technology (INR) 192012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.4.

20. Institute for Neutron Physics and Reactor Technology (INR) 202012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe DBC transients analyzed for ALFRED using SIM-LFR code TR-2 : Spurious withdrawal of the most reactive control rod TR-3 : Reactivity insertion (100 pcm) due to fuel loading error TD-1 : Spurious reactor trip TD-3 : Loss of AC power (PLOOP) TD-7 : Loss of all primary pumps (PLOF) TD-8 : Partial flow blockage of the hottest FA, with reactor trip delay TO-1 : FW temp drop (335 o C → 300 o C), reactor trip TO-4 : FW flow increase by 20%, reactor trip LEADER WP5 Task 5.4. KIT

21. Institute for Neutron Physics and Reactor Technology (INR) 212012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Preliminary conclusions General: The transients examined proved that the ALFRED plant can withstand without problems a rather wide range of accidental events. ALFRED plant has proved to be able to enter a safe shutdown phase after every DBC accident analyzed. ALFRED plant is very forgiving, and even under worst conditions (OVC - FW temperature decrease down to 300 o C), there is an extended time margin for a possible operator intervention, thus preventing damages to the plant. LEADER WP5 Task 5.4. KIT

22. Institute for Neutron Physics and Reactor Technology (INR) 222012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 22 Analyzed Transients Main events and reactor scram threshold LEADER WP5 Task 5.4. ENEA & CEA

23. Institute for Neutron Physics and Reactor Technology (INR) 232012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 23 Preliminary conclusions  In all analyzed DBC accidental transients, the protection system by reactor scram and by the prompt start-up of the DHR-1 for core decay heat removal is able to bring the plant in safe conditions in the short and long term.  The core temperatures always remains well below the safety limits.  No significant vessel wall temperature increase is predicted.  The time to reach lead freezing at MHX outlet after start-up of DHR-1 system strongly depends on the assumptions taken on cold pool mixing  but in any case there is large grace time for countermeasures by operator actions (PLOOP, PLOF, OVC). LEADER WP5 Task 5.4. ENEA & CEA

24. Institute for Neutron Physics and Reactor Technology (INR) 242012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 2015-9-18 LEADER Technical Meeting November 20 - 23, 2012 24 No significant discrepancies between simulation results from TRACE/FRED (PSI), SIM (KIT) and RELAP (ENEA) for steady state and unprotected transients (UTOP, ULOF and ULOHS) Working limits of fuel, clad material and coolant were not met for all transients Potential risk of liquid Pb freezing for SLB and OVC transients LEADER WP5 Task 5.4. PSI Preliminary conclusions

25. Institute for Neutron Physics and Reactor Technology (INR) 252012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe TD-5 Loss of one primary pump: RELAP5 preliminary nodalization (50 % work done)  Steady state Needed improvements TRB-1 Steam system piping break at SG outlet: SIMMER III preliminary model (80 % work done)  Preliminary obtained results Needed improvements LEADER WP5 Task 5.4. CIRTEN (UNIPI) Current status

26. Institute for Neutron Physics and Reactor Technology (INR) 262012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.4. JRC-IET Current status  Steady-State results  Transient analysis  ULOF  UTOP  Further developments in the model  Modeling of core by-pass  More accurate modeling of the core outlet  Modeling of heat transfer through the oxide layer on the clad and SG surfaces  Further modifications in the code  Aimed to solve issues with heat transfer in the MHX already identified

27. Institute for Neutron Physics and Reactor Technology (INR) 272012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The main objective of this task is to evaluate the impact of the core and plant design features on the Design Extended Conditions (DEC) including complex sequences. Several computer codes (i.e. SIM-LFR, SAS-LFR, RELAP, TRACE, CFX, SIMMER, SPECTRA) principally available among the partners will be applied to evaluate consequences of selected un-protected accidents scenarios such as Loss-of-Flow, Loss-of- Heat Sink, and Reactivity initiated accidents. Additionally consequences of SGTR events allocated to DEC are to be evaluated within this Task. However, priority will be given to those representative accident initiators identified and selected in Task 5.3. It will be analyzed qualitatively what specific problems might occur when part of the core materials should become mobile and become relocated within the reactor vessel. Task 5.5 Analyses of representative DEC events of the ETDR (ENEA, KIT-G, JRC-IE, KTH, NRG) D22 (T5.5, ENEA) - Report on the results of analysis of DEC events for the ETDR (March 2013)

28. Institute for Neutron Physics and Reactor Technology (INR) 282012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.5.

29. Institute for Neutron Physics and Reactor Technology (INR) 292012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.5.

30. Institute for Neutron Physics and Reactor Technology (INR) 302012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 30 Analyzed DEC Transients Main events and reactor scram threshold LEADER WP5 Task 5.5. ENEA & CEA

31. Institute for Neutron Physics and Reactor Technology (INR) 312012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 31 Preliminary conclusions  The analysis of DEC transients with RELAP5 and CATHARE codes has highlighted the very good intrinsic safety features of ALFRED design thanks to:  Good natural circulation characteristic,  Large thermal inertia, and  Significant negative reactivity feedbacks  In all analyzed transients there is no risk for significant core damage or risk for lead freezing (OVC)  large grace time is left to the operator to take the opportune corrective actions to bring the plant in safe conditions in the medium and long term. LEADER WP5 Task 5.5. ENEA & CEA

32. Institute for Neutron Physics and Reactor Technology (INR) 322012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe DEC transients analyzed for ALFRED using SIM-LFR code TR-4 : UTOP, 250 pcm in 10 sec, no reactor trip TO-3 : PLOF+FW temp drop (335 o C → 300 o C), reactor trip TO-6 : PLOF+FW flow increase by 20%, reactor trip, DHR-1 T-DEC1 : ULOF, SCS in operation, no reactor trip T-DEC3 : ULOHS, PPs operating, no reactor trip, DHR-1 T-DEC4 : ULOHS+ULOF, no reactor trip, DHR-1 T-DEC5 : SA blockage, determine max acceptable SA flow reduction factor, no reactor trip T-DEC6 : SCS failure, PPs operating, DHR-1 & DHR-2 fails, reactor trip LEADER WP5 Task 5.5. KIT

33. Institute for Neutron Physics and Reactor Technology (INR) 332012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Preliminary conclusions General: The analysis of DEC transients performed for the Pb-cooled ALFRED design demonstrated the forgiving nature of this plant design when compared to other similar plant designs, ascribable to the combination of: 1. the inherently, large thermal inertia of the Pb-cooled primary system, 2. the detailed focus on the optimization of all safety relevant systems, in particular emphasizing appropriate designs of all relevant control and safety systems (as well as components), and 3. optimizing the neutronic core characteristics of the ALFRED “core system” thereby assuring various reactivity feedback effects (fuel, diagrid, pads, Pb-coolant and CRs drivelines expansion reactivity) that effectively depress the reactor power under all adverse DEC transient conditions, 4. as the ALFRED core is specifically designed to accommodate the ULOF transient. LEADER WP5 Task 5.5. KIT

34. Institute for Neutron Physics and Reactor Technology (INR) 342012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Current status Simulations are not started yet. Preliminary results expected in January 2013 (???). LEADER WP5 Task 5.5. NRG

35. Institute for Neutron Physics and Reactor Technology (INR) 352012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.5. JRC-IET Current status  Steady-State results  Transient analysis  ULOF  UTOP  Further developments in the model  Modeling of core by-pass  More accurate modeling of the core outlet  Modeling of heat transfer through the oxide layer on the clad and SG surfaces  Further modifications in the code  Aimed to solve issues with heat transfer in the MHX already identified

36. Institute for Neutron Physics and Reactor Technology (INR) 362012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 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 LEADER WP5 Task 5.5. KTH

37. Institute for Neutron Physics and Reactor Technology (INR) 372012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 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 LEADER WP5 Task 5.5. KTH

38. Institute for Neutron Physics and Reactor Technology (INR) 382012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 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 LEADER WP5 Task 5.5. KTH

39. Institute for Neutron Physics and Reactor Technology (INR) 392012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe T-DEC1&4 ENEA’s RELAP5 model LEADER WP5 Task 5.5. KTH

40. Institute for Neutron Physics and Reactor Technology (INR) 402012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe 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  … Preliminary results expected in March 2013 (???). LEADER WP5 Task 5.5. KTH

41. Institute for Neutron Physics and Reactor Technology (INR) 412012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 The objective of this task is to provide a scoping assessment of the Source Term that may be available for release to the environment through containment leakage at some assumed rate. The Source Term assessment would include thermodynamic calculations to assess what fission products and which Polonium compounds may be formed. This task will also provide an assessment of the adequacy of the containment systems. The activity will consider the scenarios in which there may be thermal and pressure loadings of the containment due to the release of secondary coolant and fission product release from the primary system to the containment. Another activity will be that of the transport of airborne radio-nuclides in the containment and their potential release to the environment. Task 5.6 Containment and source term assessment for the ETDR (EA, KTH) T64 (T5.6, EA) - Report on Source Term Assessment for the ETDR (Rev. 0 – November 2012, Rev. 1 - March 2013) D18 (T5.6, EA) - Report on Containment Assessment for the ETDR (Rev. 0 – November 2012, Rev. 1 – March 2013)

42. Institute for Neutron Physics and Reactor Technology (INR) 422012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.6. KTH Source term assessment for the ETDR

43. Institute for Neutron Physics and Reactor Technology (INR) 432012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe LEADER WP5 Task 5.6. KTH Source term assessment for the ETDR

44. Institute for Neutron Physics and Reactor Technology (INR) 442012.11.23. E. Bubelis et. al. – LEADER PCC meeting, Karlsruhe Further actions Review the draft report (T64 ??? - Report on Source Term Assessment for the ETDR ) presented by KTH to EA, provide comments, check the data used by KTH for the source term assessment. Proceed with Task 5.6 activities and produce DEL018 - Report on Containment Assessment for the ETDR. LEADER WP5 Task 5.6. EA