1. Safety Analysis Results of the DEC Transients of ALFRED LEADER Lead-cooled European Advanced DEmonstration Reactor G. Bandini (ENEA), E. Bubelis, M. Schikorr (KIT), A. Lazaro, K. Tucek (JRC-IET) P. Kudinov, K. Kööp, M. Jeltsov (KTH), M. H. Stempnievicz (NRG), Z. Youpeng, K. Mikityuk (PSI) Technical Workshop to Review Safety and Design Aspects of ALFRED, ELFR and ELECTRA JRC-IET, Petten, 27-28 February 2013

2. 2 Outline  Introduction  The ALFRED reactor  DEC transients for ALFRED  DEC transient results  Conclusions

3. 3 Introduction  One of the main objectives of the LEADER EU project was the evaluation of the safety aspects of the lead-cooled demonstrator reactor ALFRED  Both Design Basis Conditions (DBC) and Design Extensions Conditions (DEC) have been considered in the safety analysis of ALFRED  The DEC accident scenarios are very low probability events which include the failure of prevention or mitigating systems  The main objective of DEC transient analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the plant  More representative DEC events for ALFRED have been analysed by several research organizations using different system codes

4. 4 ALFRED: Reactor block Vertical section Horizontal section  Pool-type reactor of 300 MWth power  171 fuel assemblies in the core  8 pump-bayonet tube SG connected to the 8 secondary circuits

5. 5 ALFRED: Secondary circuits DHR System (4 x 2 IC loops) In-water pool isolation condenser (IC) Valve SG Feedwater Steam From DHR system To DHR system Steam lines Feedwater lines

6. 6 Steady-state at nominal power (EOC) ParameterUnitALFREDRELAP5CATHARESIM-LFR Reactor thermal powerMW300 Total primary flow ratekg/s25980252502546025682 Total ΔP in the primary circuitbar1.5 ΔP through the corebar< 1.01.0 Core inlet temperature°C400 Upper plenum temperature°C400480 Max core outlet temperature (*)°C-483 487 Peak clad temperature°C~550508518514 Peak fuel temperature°C~2000199119852064 Feedwater temperature°C335 Feedwater flow ratekg/s192.8 196.6193.6 Steam temperature°C450 451450 Steam pressurebar180 (*) Hottest FA flow rate is ~120% of average FA flow rate

7. 7 Analysis of DEC transients Organizations and codes: ENEA (RELAP5, CATHARE), KIT (SIM-LFR), JRC-IET (TRACE, SIMMER), KTH (RELAP5), NRG (SPECTRA), PSI (TRACE/FRED) UNPROTECTED PROTECTED

8. 8 DEC: Unprotected transients Objective: Verify the intrinsic safety behaviour of the ALFRED plant and its response to more unlikely accidental events Analysed transients without reactor scram:  UTOP: Reactivity insertion of 25 pcm in 10 s (core compaction, core voiding following SGTR, etc.)  ULOF: Loss of all primary pumps  ULOHS: Loss of feedwater to all MHXs  ULOHS + ULOF: Loss of feedwater to all MHXs + loss of all primary pumps  Partial FA blockage  verify the maximum acceptable flow are blockage without fuel rod damage

9. 9 Reactivity feedbacks at EOC REACTIVITY COEFFICIENTUnitRef. TemperatureValue Control rod differential expansion (*)pcm/KT upper plenum-0.218 Coolant expansion (**)pcm/KAverage T-core-0.268 Axial clad expansionpcm/KAverage T-clad0.039 Axial wrapper tube expansionpcm/KAverage T-wrapper0.023 Radial clad expansionpcm/KAverage T-clad0.011 Radial wrapper tube expansionpcm/KAverage T-wrapper0.003 Diagrid radial core expansionpcm/KT-core inlet-0.152 Pad radial core expansionpcm/KT-core outlet-0.430 Axial fuel expansion: freepcm/KAverage T-fuel-0.155 Axial fuel expansion: linkedpcm/KAverage T-clad-0.242 Doppler constantpcmAverage T-fuel-566 (*) Prompt response (the delayed response has been neglected) (**) Calculated on the whole height of the fuel assembly (the other feedbacks are calculated only in the fissile zone)

10. UTOP transient (1/4)  Insertion of 250 pcm in 10 s without reactor scram  No feedwater control on secondary side  Codes used: TRACE, SIM-LFR, RELAP5, CATHARE, TRACE/FRED, SPECTRA 10 Total reactivity and feedbacks Core and MHX powers RELAP5 Results Total Inserted Doppler Fuel exp. Core power MHX power  Maximum net reactivity insertion of 85 pcm  Initial core power peak of 680 MW

11. UTOP transient (2/4) 11 Core temperatures Max clad and fuel temperatures RELAP5 Results Core outlet MHX inlet MHX outlet Core inlet Max fuel Max clad  Maximum clad temperature remains below 650 °C  Maximum fuel temperature of ~ 2930 °C at t = 50 s (hottest pin, middle core plane, fuel pellet centre) exceeds the MOX melting point ( ~ 2700 °C)  only local fuel melting Core outlet Max clad Core inlet

12. UTOP transient (3/4) 12  Differences in fuel expansion reactivity feedback (free/linked effects) and fuel rod gap dynamic modelling  Only local fuel melting in the hottest pin is confirmed by all codes Peak power and max fuel temperatures:  RELAP5: 679 MW and 2930 °C  SIM-LFR: 656 MW and 2996 °C  CATHARE: 735 MW and 2866 °C  TRACE/FRED: 642 MW and 2779 °C

13. UTOP transient (4/4) 13 SIM-LFR: Minimum clad failure time >> 1.0E+7 s  Different heat transfer correlations used by RELAP5 and CATHARE for fuel rod bundle  Maximum clad temperature is below 650 °C

14. 14 ULOF transient (1/4)  All primary pumps coastdown without reactor scram  No feedwater control on secondary side  Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA Active core flowrateCore and MHX powers RELAP5 Results Core power MHX power  Natural circulation in the primary circuit stabilizes at 23% of nominal value  Core power reduces down to about 200 MW due to negative reactivity feedbacks

15. 15 ULOF transient (2/4) Core temperatures RELAP5 Results  Initial clad peak temperature of 764 °C  Max clad temp. stabilizes below 650 °C  Positive Doppler and fuel exp. effects are mainly counterbalanced by negative radial core exp. (Pad + Diag.), control rods and coolant exp. effects Total reactivity and feedbacks Max clad Max lead Core inlet Max clad Core inlet Max fuel Doppler Fuel exp. C. Rods Pad + Diag. Cool. exp.

16. 16 ULOF transient (3/4)  Slight deviations in the initial core flow rate transient, but good agreement in stabilized natural circulation flow rate in the primary circuit  Core power at t = 200 s is slightly under predicted by SIM-LFR (P = 177 MW) and TRACE/FRED (P = 180 MW) with respect to RELAP5 (P = 195 MW) and CATHARE (P = 198 MW)

17. 17 ULOF transient (4/4) SIM-LFR: Minimum clad failure time > 1.0E+5 s  The initial clad peak temperature is calculated in the range 730° C– 764°C  Maximum clad temperature predicted by the codes at t = 200 s is around 650 °C  No clad failure is expected under ULOF in the short and long term  No vessel wall temperature increase (Tw < 400 ° C during ULOF transient)

18. 18 ULOHS transient (1/4)  Loss of feedwater to all MHXs without reactor scram  Startup of DHR-1 (3 out of 4 IC loops are in service)  Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA  Core power progressively reduces down towards decay level  removed by DHR-1  Maximum clad and vessel temperatures rise up to ~ 700 °C after about one hour Core and MHX powers Core and vessel temperatures Core power MHX power Max vessel Max clad CATHARE Results

19. 19 ULOHS transient (2/4) Total reactivity and feedbacksCore temperatures  Fuel temperature reduces down close to clad temperature  Positive Doppler and fuel and clad expansion effects are mainly counterbalanced by negative radial core expansion (Pad + Diag.), coolant expansion and control rods effects Max clad Core inlet Max fuel Doppler Fuel exp. Clad exp. C. Rods Cool. exp. Pad + Diag. CATHARE Results

20. 20 ULOHS transient (3/4)  Vessel wall temperature is over predicted by RELAP5 and CATHARE (no heat losses from the external wall surface) with respect to SIM-LFR  Maximum vessel temperature rises over about 650 °C in 30 minutes  no vessel failure is expected in the medium term  vessel integrity is not guaranteed in the long term

21. 21 ULOHS transient (4/4) SIM-LFR: Minimum clad failure time > 1.0E+6 s  Maximum clad temperature stabilizes around 700 °C after one hour transient  No clad failure is calculated by SIM-LFR code in the short and long term

22. 22 ULOHS+ULOF transient (1/4)  Loss of feedwater to all MHXs and all primary pumps without reactor scram  Startup of DHR-1 (3 out of 4 IC loops are in service)  Codes used: SIM-LFR, RELAP5, CATHARE, SPECTRA Core flow rate and powerCore and vessel temperatures Max fuel Max lead, clad Core inlet, vessel Power Flow rate SIM-LFR Results  Sharp decrease of core power and flow rate in the initial transient phase and then their progressive decrease  Core flow rate/power ratio is ~ 1/3 of nominal value  Large ΔT through the core  Maximum clad temperature rises up to ~ 800 °C in 30 minutes

23. 23 ULOHS+ULOF transient (2/4)

24. 24 ULOHS+ULOF transient (3/4)  Similar evolution of core flow rate and core power is calculated by the codes  Calculated vessel wall temperature is in the range 440 °C - 520 °C after 30 min.  Vessel integrity is guaranteed in the medium term and likely also in the long term according to RELAP5 results

25. 25 ULOHS+ULOF transient (4/4) SIM-LFR: Minimum clad failure time > 1.0E+4 s  The maximum clad temperature is over predicted of 25° C – 30 °C by RELAP5 and CATHARE with respect to SIM-LFR  The minimum clad failure time predicted by SIM-LFR is of about 3 hours

26. 26 Partial FA blockage (RELAP5 results) Code used: RELAP5, SIM-LFR, SIMMER RELAP5 assumptions:  Total ΔP over the FA = 1.0 bar  ΔP at FA inlet = 0.22 bar  Flow area blockage at FA inlet  No heat exchange with surrounding FAs MAIN RESULTS:  75% FA flow area blockage  50% FA flowrate reduction  85% blockage  T-max clad = 700 °C  No clad melting if area blockage < 95%  Fuel melting if area blockage > 97.5%  50% inlet flow area blockage can be detected by TCs at FA outlet

27. 27 SCS Failure (1/2) (RELAP5 results) Secondary pressure Core and MHX powers Primary lead temperatures  Depressurization of all secondary circuits at t = 0 s (no availability of the DHR)  Reactor scram at t = 2 s on low secondary pressure  Initial MHX power increase up to 850 MW  no risk for lead freezing MHXs Core MHX outlet

28. 28 SCS Failure (2/2) (RELAP5 results) Core decay and MHX powersCore and vessel temperatures  No risk for lead freezing in the initial transient phase  Slow primary temperature increase due to large thermal inertia of the primary system  large grace time for the operator to take opportune corrective actions Core MHX

29. 29 Conclusions (1/2)  In all simulated transients there is a very large margin to coolant boiling since the coolant is always at least 900 °C below the lead boiling point (1740 °C)  Clad failure is not predicted in all simulated transients except for:  Undetected FA blockage greater than ~ 85% which might be excluded by design (many orifices at the FA inlet)  The very unlikely ULOHS+ULOF event, when the time-to-failure reduces down to few hours, but still leaving enough grace time for corrective operator actions  Fuel melting is excluded in all simulated transients except for local fuel melting in the hottest pins in case of UTOP transient  The vessel integrity seems guaranteed in the long term in all simulated transients except for the ULOHS transient, but even in this case there is enough grace time for corrective operator actions  No relevant safety issues have been identified for ALFRED in case of representative DEC events – In particular the ULOF transient can be accommodated without the need of corrective operator actions

30. 30 Conclusions (2/2)  The analysis of DEC transients with various codes has highlighted the very good intrinsic safety features of ALFRED design thanks to:  Benign characteristics of the coolant  Good natural convection in the primary circuit  Large thermal inertia to slow down the transients  Prevalent negative reactivity feedbacks to limit power excursions  In all analyzed unprotected transients there is no risk for significant core damage and then for transient evolution towards severe accidents  enough grace time is left to the operator to take the opportune corrective actions and bring the plant in safe conditions in the medium and long term