LEADER Project: Task 5.4 Analysis of Representative DBC Events of the ETDR with CATHARE G. Geffraye, D. Kadri – CEA/Grenoble G. Bandini - ENEA/Bologna LEADER 5 th WP5 Meeting JRC-IET, Petten, 26 February 2013

2 Outline  CATHARE code for LFRs  CATHARE modelling  Steady-state at EOC  Analysed DBC transients  Transient results  Conclusions

3 CATHARE code for LFRs  The CATHARE code is the result of a joint effort of CEA, EDF, Framatome ANP and IRSN  The CATHARE system code is commonly used for thermal-hydraulic transient analysis and best-estimate safety analysis of light water reactors. The CATHARE V2.5_2/mod5.1 can be also applied for safety analysis of gas-cooled and sodium-cooled fast reactors  The lead-bismuth eutectic (LBE) and lead thermo-physical properties have been recently implemented in the CATHARE code, in the frame of a bilateral collaboration between ENEA and CEA, for the safety analysis of lead-cooled fast reactors  Some validation works have been performed or are in progress at ENEA on the basis of:  The experiments conducted in the Korean LBE-cooled HELIOS loop in the frame of the OECD LACANES benchmark, including a comparison with RELAP5 code;  The experiments carried out in the LBE-cooled NACIE facility of ENEA/Brasimone;  The experiments carried out on the LBE-cooled TALL loop of KTH/Stockholm within the on-going activities of the THINS European project.  The whole core, primary system and secondary system discretization of the ALFRED reactor used in the CATHARE simulation has been harmonized with the one used in the RELAP5 simulation performed by ENEA, in order to better compare the code results  The systematic comparison of CATHARE results with the results of the more validated RELAP5 code, regarding both DBC and DEC transient analysis, has confirmed the good applicability of CATHARE for ALFRED transient simulations

4 ALFRED modelling ALFRED Nodalization scheme with CATHARE Primary circuit 2 Secondary loops (weight 4) 2 IC loops (weight 4) Cold pool modelled by two 1D axial elements

5 Steady-state at EOC ParameterUnitCATHARE Reactor thermal powerMW300 Total primary flowratekg/s25740 Active core flowratekg/s25460 Average FA flowratekg/s148.7 Hottest FA flowratekg/s176.8 Pressure loss through the primary circuitbar1.5 Pressure loss through the corebar1.0 Core inlet lead temperature°C400 Average FA outlet lead temperature°C480 Hottest FA outlet lead temperature°C483 Upper plenum lead temperature°C480 Average pin max clad temperature°C508 Hottest pin max clad temperature°C518 Average pin max fuel temperature°C1635 Hottest pin max fuel temperature°C1985 SG inlet lead temperature°C480 SG outlet lead temperature°C400 Total SG feedwater flowrate (8 SGs)kg/s196.6 SG feedwater temperature°C335 SG steam outlet temperature°C451 SG inlet pressurebar188 SG outlet pressurebar182 Steam line outlet pressurebar180

6 Main events and reactor scram threshold Analysed DBC transients

7 TD-1: Spurious reactor trip (1/2) Total reactivity and feedbacks ASSUMPTIONS:  Reactor scram at t = 0 s  Reactivity insertion of at least 8000 pcm in 1 s  Secondary circuits are available  constant feedwater flowrate  Core power reduced down to decay level at t = 0 s  Power removal by secondary circuits reduces with decreasing primary temperatures Core and MHX powers

8 Core temperatures  Initial temperature gradient on the fuel rod clad is about -8 °C/s  No risk for lead freezing since the feedwater temperature (335 °C) remains above the solidification point of lead (327 °C) TD-1: Spurious reactor trip (2/2) Primary lead temperatures

9 TD-3: Loss of AC power (1/2) Active core flowrate ASSUMPTIONS:  At t = 0 s  Reactor scram, primary pump coastdown, feedwater and turbine trip  At t = 1 s  DHR-1 system activation (4 IC loops  risk of lead freezing) Core temperatures  No initial core flowrate undershoot (lead free levels equalization)  No significant clad temperature peak in the initial phase of the transient

10 TD-3: Loss of AC power (2/2) Core decay, MHX and IC powers Primary lead temperatures  After the initial transient the natural circulation in the primary circuit stabilizes around 2% of nominal value  DHR power (7 MW) exceeds the decay power after about 15 minutes  After 3 hours the minimum lead temperature at MHX outlet is still far enough from the lead solidification point (mixing in the cold pool around MHXs is effective) Active core flowrate

11 TD-7: Loss of primary pumps (1/2) ASSUMPTIONS:  At t = 0 s  All primary pumps coastdown  Reactor scram at t = 3 s on second scram threshold (Hot FA ΔT > 1.2 nominal value)  At t = 4 s  DHR-1 system activation (3 IC loops  maximum temperatures) Active core flowrateCore temperatures  No initial core flowrate undershoot (lead free levels stabilization)  More significant clad temperature peak than in case of LOOP transient due to delayed reactor scram

TD-7: Loss of primary pumps (2/2) Active core flowrate Core decay, MHX and IC powers Primary lead temperatures  After the initial transient the natural circulation in the primary circuit stabilizes around 1.5% of nominal value  DHR power (5 MW) exceeds the decay power after about 45 minutes  No risk of lead freezing at MHX outlet in the short and medium term (mixing in the cold pool around MHXs is effective)

13 TO-1: FW temper. reduction (1/2) ASSUMPTIONS:  Loss of one preheater (FW temperature from 335 °C down to 300 °C in 1 s)  reactor scram at t = 2 s on low FW temperature  At t = 3 s  DHR-1 system activation (4 IC loops) Primary lead temperatures  DT through the core and the MHX reduces quickly down to few degrees  After some fluctuations the primary lead temperatures stabilizes around 425 °C

14 Core decay, MHX and IC powers Primary lead temperatures TO-1: FW temper. reduction (2/2)  No risk of lead freezing in the initial phase of the transient due to prompt reactor scram  After about 15 minutes the DHR power (7 MW) exceeds the decay power  No risk of lead freezing in the sort and medium term (mixing in the cold pool around MHXs is effective)

15 TO-1: FW flowrate +20% Core and MHX powers Primary lead temperatures ASSUMPTIONS:  Feedwater flowrate +20% in 25 s  No significant perturbations on both primary and secondary sides  The system reaches a new steady-state condition in about 10 minutes without exceeding reactor scram set-points  Slight increase in core power (+4%) leads to max fuel temperature increase of 40 °C

16 Maximum core temperatures Transient DescriptionCode SystemMax Temperatures [°C] FuelCladdingCoolant NominalSteady state, peak pin - ENEACATHARE TD-1Spurious reactor tripCATHARE TD-3Loss of AC powerCATHARE TD-7Loss of all primary pumps (PLOF)CATHARE TO-1Reduction of FW temperatureCATHARE TO-4Increase of FW flowrate by 20 %CATHARE

17 Conclusions In all analysed DBC accidental transients the protection system by reactor scram and prompt start-up of the DHR-1 system for core decay heat removal is able to bring the plant in safe conditions in the short and long term. The core temperatures (clad and fuel) always remain well below the safety limits and no significant vessel wall temperature increase is predicted. The time to reach lead freezing at the MHX outlet after start-up of DHR-1 system strongly depends on the assumptions taken on the lead mixing in the cold pool surrounding the MHX that involves the largest part of the primary lead mass inventory. In the CATHARE calculations the cold lead flowing out of the MHX mixes with hotter lead of the cold pool surrounding the MHX, before to move downward into the lower plenum towards the core inlet. Therefore, in the calculations of TD-1, TD-7 and TO-1 transients, the decrease of lead temperature in the primary system is significantly delayed by the coolant mixing in the cold pool, that increases noticeably the effective thermal inertia of the primary system. This cold pool mixing effect (not observed in the analysis with the RELAP5 code by ENEA) mainly explains the large difference between CATHARE and RELAP5 results, regarding the time needed to approach the risk of lead freezing following DHR-1 start-up.