1. Melt Pool Behavior and Coolability in the Lower Head of a Light Water Reactor - Progress in WP5-2 of SARNET2 Weimin Ma Division of Nuclear Power Safety Royal Institute of Technology (KTH) The 5th European Review Meeting on Severe Accident Research (ERMSAR-2012) Cologne, Germany, March 21-23, 2012 KTH: W.M. Ma, W. Villanueva, P. Kudinov; KIT: A. Miassoedov, X. Gaus-Liu RUB: M. Hoffmann; PSI: B. Jaeckel, J. Birchley; IVS: P. Matejovic, M. Barnak

2. 2  WP5-2 of SARNET2: - Remelting of debris, melt pool formation and coolability (MPF)  This topical area of research is concerned with corium behaviour and coolability in the lower head.  The objectives are: To create and enhance the database on the corium simulant behavior in the lower head. To develop and validate the models and computational codes for simulation of in-vessel melt pool and vessel behavior. To assess the influence of potential SAM measures (e.g. CRGT cooling and ex-vessel cooling) on in-vessel corium coolability and retention. The Scope

3. 3  Dual approach: Experimental and analytical.  Experimental work The LIVE program (KIT) is to provide further experiments with emphasis on the pool build-up and behavior in a well-scaled lower head using corium simulant. Experiments at the CNU facility (CEA) dedicated to study two-phase flow (and CHF) around a heated hemispherical geometry of a PWR. The data will be unique for the external vessel cooling in conditions more representative for the reactor case in 3D geometry (1:10 angular sector). Approach

4. 4  Analytical work Approach (contd.) USTUTT (IKE) – Stuttgart University, Germany INRNE – Institute for Nuclear Research and Nuclear Energy, Bulgaria TUS – Technical University of Sofia, Bulgaria IVS – Inzinierska Vypoctova Spolocnost, Slovakia PSI – Paul Scherrer Institut, Switzerland RUB – Ruhr-Universität Bochum, Germany

5. 5  Paper 2.1: Progress in Research on Melt Pool Behavior and Coolability in the Lower Head of a Light Water Reactor. Some results of LIVE-2D tests Calculations of fuel degradation tests on the ACRR. Thermo-mechanical analysis of BWR vessel failure under CRGT cooling Thermal load analysis related to the IVR of VVER440  Paper 2.5: Analysis and Interpretation of the LIVE-L6 Experiment To investigate the thermal-hydraulic behavior of a stratified melt pool in the reactor vessel lower head.  Paper 2.13: The Experimental Results of LIVE-L8B – Debris Melting Process in a Simulated PWR Lower Head. Contents

6. LIVE-2D Test Facility LIVE-2D – Study of both transient and steady-state behavior of the 2D molten pool – 12 cm thick slice, Ø1 m, wall thickness 25 mm – NaNO 3 -KNO 3 melts (T max ~ 370 °C) – Phase diagrams similar to corium – Supporting tests to compare with SIMECO and LIVE-3D Main investigated thermal-hydraulic parameters: – Melt temperature distribution – Heat flux distribution through the wall – Heat transfer from the upper layer to the vessel wall

7. 7 Test Cooling conditions Time (hh:mm) Power (W) Remarks LidVessel Wall L00A1Water 0-27:157316 Power is decreasing with each step 27:15-54:276275 54:27-96:154206 L01Water 0-30:157311 Order of Power levels: High, low, high 30:15-70:503691 70:50-99:157316 L02Water 0-28:453694 Order of Power levels: Low, high, low 28:45-71:157315 71:15-99:153706 L03_01InsulatedWater 0-50:132145 Cooling right from the start 50:13-95:131112 L03_AInsulatedWater 0-50:252150 Cooling of vessel wall only with delay after pouring in 50:25-95:251118 L04InsulatedWater0-46:282155 At beginning of test 50% debris already in the vessel LIVE-2D Tests

8. 8 Results of LIVE-2D (1) Comparison of the upward heat transfer in LIVE-2D tests with earlier experiments

9. 9 Results of LIVE-2D (2) Comparison of the upward and downward heat transfer in LIVE-2D tests with SEMICO experiments

10. 10 Analysis of Tests on ACRR  ACCR: Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL).  DF-4 was an in-pile experiment to investigate early-phase degradation of BWR fuel canisters and control blades.  MP-2 test performed was to investigate late-phase melt procession in LWRs.  DF-4 test was simulated by different versions of MELCOR, and the results for the degradation were essentially similar, but MELCOR 1.8.5 has kinetic models for eutectic interactions.  MP-2 test was simulated by the ATHLET-CD code with and without late phase module MEWA, and the results show that the code with the MEWA module has better prediction.

11.  Control Rod Guide Tube (CRGT) cooling is being considered as a potenetial SAM measure for Nordic BWRs.  Given a thermal and mechanical load on the lower head, what is the mode of failure of the vessel wall?  Is in-vessel retention possible by means of combined CRGT cooling and ex-vessel cooling? Analysis of Vessel Failure Modes

12. Temperature H = 0.7 m DB H = 1.9 m DB t = 4.44 h t = 3.47 h  Different amounts of melt relocated to the lower- head have been considered, corresponding to 0.7m, 1.1 m, 1.5m and 1.9 m deep pool.  PECM model was used to predict heat transfer and phase changes in the melt. Snapshots of Temperature

13. ANSYS 13 specifics Full transient analyses Strong structural-thermal coupling Plane223: 2D 8-node coupled-field element 850 quadrilateral elements and 2901 nodes Creep Analysis in ANSYS

14. 0.7 m deep melt pool The maximum displacement is located at the bottom of the RPV t = 5.0 h ‘Ballooning’ Mode of Vessel Failure t = 4.94 h 1.1 m deep melt poo1

15. 1.5 m deep melt pool ‘Localized’ Mode of Vessel Failure 1.9 m deep melt poo1 t = 3.89 h t = 3.61 h Creep intensifies in one location (localized creep). Displacement of the lower part of the vessel is relatively uniform.

16. CHF limit External cooling reduces the heat fluxes on the outer wall H = 1.9 m Simulations taking into account the external vessel cooling on the melt heat transfer. Cooling of the vessel from outside effectively reduces temperature on the inner side of the vessel wall. Effect of External Cooling

17. Melt Pool Convection in VVER-440  In-vessel retention of corium in the lower head with external cooling is considered for VVER-440.  The molten pool in the lower head was simulated by the stand- alone DIVA module of ASTEC V1.3 rev.2 code  Boundary condition – constant heat transfer coefficient and coolant temperature – was applied on the outer surface in order to simulate the external reactor vessel cooling.  Assumptions of base case:100% of UO 2, 50 tons of stainless steel (SS) and 19 tons of zirconium were relocated into the lower head.

18. Sensitivity Study (1)  Effect of molten pool configuration on the thermal load.

19.  Effect of stainless steel mass on the thermal load. Sensitivity Study (2)

20.  Effect of decay heat distribution between oxidic and metallic layers on the thermal load. Sensitivity Study (3)

21. Concluding Remarks  Progress on melt pool and lower head behaviour related to WP5-2 of SARNET2 was reported.  The LIVE-2D experiment shows a good agreement for downward heat transfer with SIMECO, but upward heat transfer is considerably higher than in the SIMECO.  The MELCOR analysis of the ACRR DF-4 test indicates only minor oxidation of B 4 C because of the lack of any treatment in MELCOR for oxidation of the molten mixture.  The ACRR MP-2 test was better simulated by the ATHLET-CD code with the MEWA module.  The coupled thermo-mechanical creep analysis of a BWR lower head reveals that if only CRGT cooling is activated, the cases with 0.7 m and 1.1 m melt pool depths result in a ballooning type of vessel failure while the cases with 1.5 m and 1.9 m melt pool depths result in a localized creep. If external vessel cooling plus CRGT cooling is implemented, in-vessel retention of the melt is possible.  Heat transfer of molten pool in the lower head of a VVER-440 reactor was studied by the ASTEC V1.3 code. Sensitivity study shows for various molten pool configuration, stainless steel mass in molten pool and decay heat distribution, the heat flux is well below the cooling limit (CHF).