Balestra, P.

Bandini, G.

Ederli, S.

Parisi, C.

Perez-Martin, S.

Pfrang, W.

English

KITopen

KIT –  University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association  www.kit.edu NURETH-16 International Topical Meeting on Nuclear Reactor Thermal Hydraulics August 30-September 4, 2015, Chicago Single and two-phase sodium flow analysis for two TUCOP CABRI tests using the ASTEC-Na code Sara Perez-Martin, Werner Pfrang (KIT, Germany), Giacomino Bandini, Stefano Ederli, Paolo Balestra, Carlo Parisi (ENEA, Italy)  2 NURETH-16 August 30-September 4, 2015, Chicago Introduction  High safety standards need a robust validated computational simulation software able to predict fuel pin behaviour, thermodynamics, thermo-hydraulics and the whole accident sequence.   Code system development and validation require adequate sets of experimental data covering the various interfering physical processes and phenomena taking place during an accidental event.   The JASMIN project of the 7th European Framework Programme aims at developing a new computer code system, ASTEC-Na, capable to evaluate the consequences of protected and unprotected accidents in Sodium-cooled Fast Reactors.   JASMIN project is devoted to modelling and validation of:  Sodium thermal-hydraulics  Fuel pin thermo-mechanical behaviour  Source term  Neutron physics  3 NURETH-16 August 30-September 4, 2015, Chicago Introduction   In JASMIN project ASTEC-Na sodium thermal-hydraulics models will be validated using:   SCARABEE and CABRI in-pile experiments  Natural circulation test conducted in PHENIX reactor  Tests to be carried out in the KASOLA sodium loop (ULOF, ULOHS, natural circulation)   ASTEC-Na models will also be benchmarked against other safety codes: SAS-SFR, CATHARE, SIMMER, RELAP5-3D, RELAP5-Na.   Participants of the present TUCOP tests simulations (E8 and EFM1) are:  ENEA (Italy) using ASTEC-Na, CATHARE and RELAP5-3D  KIT (Germany) using SAS-SFR  4 NURETH-16 August 30-September 4, 2015, Chicago Outline   ASTEC-Na Thermal-Hydraulics Models    Transient Under Cooling Over Power (TUCOP=LOF+TOP) Tests Analyses   E8 Test: 22.1 s LOF phase (21.6 s 1phase + 0.5 s 2phase)   EFM1 Test: 30.7 s LOF phase (22.7 s 1phase + 8.0 s 2phase)    Conclusions and Future Work 5 NURETH-16 August 30-September 4, 2015, Chicago ASTEC-Na thermal-hydraulics models  ASTEC-Na based on ASTEC code system (IRSN and GRS), extensively validated in European projects for LWR.    CESAR is the thermal-hydraulics module of ASTEC-Na.   Adapted to SFR by implementing Na physical properties and updating heat and mass fluxes for liquid-vapor phases and Na-wall interphases.   Provides boundary conditions to: fuel pin behavior, neutronics, fission products and  aerosol transport , ...   Description of the CESAR models:  Liquid-vapor heat and mass exchanges based on the kinetic theory of gases.   Flows at the interphase: vapor condensation when hitting the wall and spontaneous liquid evaporation.   Sodium flashing: pressure is lower than the liquid saturation one,  liquid bulk boiling.  Bulk condensation: vapor pressure is greater than the vapor saturation pressure.  Convective mechanism between Na phases,  significant in volumes with massively non-condensable gases.  Newton’s law of cooling where HTC is estimated as the inverse of individual thermal resistances of phases:   pure conduction for the dispersed phase (bubbles or droplets)  convection for the continuous phase   6 NURETH-16 August 30-September 4, 2015, Chicago ASTEC-Na thermal-hydraulics models  Wall-fluid heat exchange phenomena modeled: convection, nucleate boiling, film boiling, thermal radiation and droplet projection (heat flux from the droplets in the quench front).            For convection several options are given to estimate Nusselt number.  Nucleate boiling is modeled with the Forster & Zuber correlation.  Critical Heat Flux Temperature TCHF obtained from Thom’s correlation.  Critical Heat Flux qCHF from Zuber’s correlation corrected for liquid subcooling.  Minimum Stable Film Temperature TMSF calculated using Berenson’s correlation.  Radiation exchange based on the grey-body approximation (300 W/m2K HTC value in the projection region).  Convection and radiation for wall-vapor heat transfer (max. convection assumed between nat. and forc. reg.)  Dittus-Boelter correlation used for forced regimes. 7 NURETH-16 August 30-September 4, 2015, Chicago RELAP5-3D, CATHARE, SAS-SFR codes  RELAP5-3D (Idaho National Laboratory)  Thermal-hydraulic transient analysis of LWR but extended to SFR.  Wide variety of hydraulic and thermal transients (nuclear and non-nuclear systems).  Mixtures of vapor, liquid, non-condensable gases and non-volatile solutes.  Seban-Shimazaki correlation is used for convective heat transfer with sodium coolant.    CATHARE (CEA, EDF, AREVA and IRSN)  Originally conceived for PWR safety studies and recently extended to other nuclear reactors, such as SFR.  Flexible modular structure for thermal-hydraulics: simple experimental test facilities to nuclear power plants.   The Spukinski heat transfer correlation is used for sodium.    SAS-SFR (KIT/INR, CEA, IRSN and JAEA)  Deterministic analysis for steady state power operation and accident conditions in SFR during  the initiation phase.   Extensive qualification of steady-state fuel irradiation, transient fuel deformation, primary coolant system heat transport, sodium boiling model, cladding melting and motion, fuel failure behavior in voided and un-voided regions.   Multi-channel model with fuel assemblies represented by a single pin. Multiple-bubble slug ejection model.  8 NURETH-16 August 30-September 4, 2015, Chicago Transient Under Cooling Over Power (TUCOP)   E8 test Annular MOX fuel with 4.6 at. % burn-up 316 SS cladding PPN linear power 593 W/cm TOP triggered in a partially voided channel 22.1 s LOF phase (21.6 s 1ph + 0.5 s 2ph)    EFM1 test Annular MOX fuel with 6.4 at. % burn-up 15-15 Ti cladding PPN linear power 487 W/cm TOP triggered in a voided channel 30.7 s LOF phase (22.7 s 1ph + 8 s 2ph) 9 NURETH-16 August 30-September 4, 2015, Chicago E8 and EFM1 Code Modelling  ASTEC-Na modelling: fissile column, lower and upper fertile blankets, fission gas plena and lower and upper structures of the test section.  CATHARE, RELAP5-3D and SAS-SFR consider the same regions as ASTEC-Na.   RELAP5-3D only simulates the structure of the test section above the fuel pin.  E8 test EFM1 test Imposed total flow rate with regulated pumpBypassImposed pressure vs. timeSection heights with hot fissile, assumed in the calculation, are between brackets0.0-20.1-83.4-87.4-88.414.063.320.175.8 (76.6)21.836.3 (35.5)75.8 (76.6)97.6 (98.4)133.930.347.4164.2211.610.0 221.6height (cm) elevation (cm)Level 0 corresponds to the BOTTOM of FISSILE COLUMN (BFC)Imposed total flow rate with regulated pumpBypassImposed pressure vs. timeSection heights with hot fissile, assumed in the calculation, are between brackets0-2.0-79.3-83.3-84.314.077.375.0 (76.2)53.6 (52.4)75.0 (76.2)78.0 (79.2)131.653.530.4185.1215.510.0 225.5height (cm) elevation (cm)Level 0 corresponds to the BOTTOM of FISSILE COLUMN (BFC)2310 NURETH-16 August 30-September 4, 2015, Chicago E8 and EFM1 Initial Conditions  GERMINAL code provides pin characteristics to ASTEC-Na, CATHARE and RELAP5-3D after the PHENIX power operation simulation.  SAS-SFR is able to simulate the power operation providing pin geometry, fission gas retention, fuel and clad thermal and mechanical characteristics.    Experiment ASTEC-Na CATHARE RELAP5-3D SAS-SFR E8 Fuel Clad Gap Width at PPN (µm) - 66.12 66.10 66.12 0.00 Inner Fuel Radius at PPN (mm) - 1.023 1.023 1.023 1.144 Total channel power (W) 36,709 36,600 36,600 36,600 38,830 Total power produced in the fuel (%) 96.6 100 100 100 98.0 Peak Linear Rating (W/cm) 593 611 612 626 608 EFM1 Fuel Clad Gap Width (µm) - 80.30 80.30 80.30 0.00 Inner Fuel Radius (mm) - 1.036 1.036 1.036 1.234 Total channel power (W) 31,050 31,050 31,050 31,050 31,530 Total power produced in the fuel (%) 98.3 100 100 100 97.3 Peak Linear Rating (W/cm) 491 504 504 513 491 Exp. uncert. ~4% E8 EFM1 11 NURETH-16 August 30-September 4, 2015, Chicago E8 test results   Linear temperature increase Pin distortion: diff. cross sections  Estimated TC uncertainty 1.6 % LBO at TFC 20.7 s BBO total at 21.6 s (0.9 s)    12 NURETH-16 August 30-September 4, 2015, Chicago E8 test results  Axial profiles of coolant temperature    Experiment ASTEC-Na CATHARE RELAP5-3D SAS-SFR Local boiling onset time (s) 20.7 - - - - Bulk boiling onset (s) 21.6 20.92 21.10 20.40 21.98 Boiling onset height (cm BFC) 75* 75.62 75.62 69.72 79.17 Saturation temperature (K) 1250 1247.2 1246.3 1248.0 1251.0 Pressure at TFC (bar) - 2.17 2.19 2.19 2.18 Clad dry-out time (s) Not observ. 22.0 22.5 - - Clad melting onset time (s) Not observ. 22.7 22.8 - - 13 NURETH-16 August 30-September 4, 2015, Chicago E8 test results  Inlet and Outlet Flow rate after boiling onset          Voiding front  Flow reversal 14 NURETH-16 August 30-September 4, 2015, Chicago EFM1 test results Pin distortion Estimated TC uncertainty 1.6 % LBO at TFC 21.9 s BBO total at  22.7 s (0.8 s) Clad dry-out observed (3-3.7 s) Fission gas blow-out  Clad melting           15 NURETH-16 August 30-September 4, 2015, Chicago EFM1 test results  Axial profiles of coolant temperature    Experiment ASTEC-Na CATHARE RELAP5-3D SAS-SFR Local boiling onset time (s) 21.9 - - - - Bulk boiling onset time (s) 22.7 21.80 21.70 21.60 22.51 Boiling onset height (cm BFC) 75* 75.22 75.22 75.45 77.00 Saturation temperature (K) 1246 1242 1241 1252 1248 Pressure at TFC (bar) - 2.09 2.10 2.27 2.12 Clad dry-out time (s) 25.7-26.4 23.00 23.5 - 25.1 Clad melting onset time (s) - 24.70 24.7 - 26.3 16 NURETH-16 August 30-September 4, 2015, Chicago EFM1 test results  Inlet and Outlet Flow rate after boiling onset          Voiding front  17 NURETH-16 August 30-September 4, 2015, Chicago ASTEC-Na sensitivity analysis in EFM1 test  Modification in the number of axial meshes above fissile zone            Coolant Temperature Base Case (BC) Mesh   14 HTC     800 Thick   0.5   Sensitivity Case 1 (SC1) Mesh   54 HTC     800 Thick   0.5 18 NURETH-16 August 30-September 4, 2015, Chicago ASTEC-Na sensitivity analysis in EFM1 test  Inlet and Outlet Flow Rate         Voiding Front  19 NURETH-16 August 30-September 4, 2015, Chicago Conclusions  ASTEC-Na satisfactory results during the single phase of both E8 and EFM1 tests.  Small differences in boiling onset between ASTEC-Na and the experimental data.  Difficulty to consider the fuel pin distortion, local boiling inception.  ASTEC-Na inlet and outlet flow after boiling onset is inaccurate. Lower voiding interface is in agreement with the experimental data for E8 and slightly underestimated for EFM1. The upper front is underestimated for both tests.   Parametric studies on the consideration of radial heat losses and on the meshing above the fissile zone:  The finest meshing slows down the flow rate reduction after the onset of boiling according to experimental measurements  The heat loss and thermal inertia of heated section above the fissile zone has impact on the axial coolant temperature distribution in the upper part of the test section affecting after boiling onset also the voiding behaviour and coolant temperatures in the lower part.  Code-code comparison: differences attributed to models, radial heat losses and pressure calculation. 20 NURETH-16 August 30-September 4, 2015, Chicago Future work  Implementation of Lockhart-Martinelli model for two phase pressure drop based on the single phase: ∆𝑃2𝜑 = Φ2∆𝑃1𝜑 (Currently ASTEC-Na wall friction model is based on Blasius correlation with an averaged parameter, homogeneous mixture)   Implementation of various correlations for Φ2 estimation   Validation foreseen:  Steady-state boiling experiments: ISPRA Tubular Experiment and KNS-37 S-33  LOF transients: KNS-37 L-22 21 NURETH-16 August 30-September 4, 2015, Chicago Acknowledgement This work was financed by the JASMIN project within the 7th Framework Programme of the European Commission.       www.jasmin-fp7.eu  

SIngle and two-phase sodium flow analysis for two TUCOP CABRI tests using the ASTEC-Na code

https://publikationen.bibliothek.kit.edu/240102828/3816950

SIngle and two-phase sodium flow analysis for two TUCOP CABRI tests using the ASTEC-Na code

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