ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like Lucas Albright a,b, Dr. Polina Wilhelm b, Dr. Tatjana Jevremovic a,c a Nuclear Engineering Program b Helmholtz-ZentrumDresden-Rossendorf (HZDR) c InternationalAtomic Energy Agency (IAEA) IAEA Technical Meeting on the Status and Evaluation of Severe Accident Simulation Codes for Water Cooled Reactors 9-12 October 2017, Vienna, Austria
Objectives Characterize the progression of severe accidents in the Westinghouse AP1000 and identify key grace periods relevant to core degradation More generally gain insights into the progression of severe accidents in advanced LWRs Observe the effects of the large cavity above the upper core plate on in-vessel phenomena and core degradation Page 2
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 3
Background CORA, QUENCH, Phébus experimental programs Key findings: Heat produced by Zr alloy oxidation can exceed the decay heat produced in the core Interactions between core materials such as the formation of eutectic mixtures and low-melting point alloys Open literature studies on the efficacy of the AP1000 passive safety systems in various accident scenarios (SBLOCA, LBLOCA, and SBO) Currently no systematic characterization of the progression of severe accidents in the AP1000 that identifies key grace periods relevant for core degradation Page 4
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 5
In-vessel Core Degradation Thermal hydraulics changes in temperature, pressure, and mass flow rates in both the primary and secondary systems Oxidation of core materials highly exothermic Zr oxidation ballooning and rupture of the cladding Page 6
In-vessel Core Degradation Relocation of core materials Control rod absorbers (AIC, ~1070 K ) Eutectic mixtures and lower melting point alloy formation Loss of rod-like geometry occurs near 2500 ± 200 K after the dissolution of UO 2 Magma pool and crust formation Relocation to the lower plenum Lower head failure Page 7
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 8
AP1000 Basics Pressurized Water Reactor (PWR) Two-loop (3 legs/loop) 3400 MWt, 1000 MWe T.L. Schulz. 2006. Westinghouse AP1000 advanced passive plant. Nuclear Engineering and Design. 236. p. 1547 1557. Passive Safety Systems High pressure injection systems Low pressure injection systems Automatic depressurization system (ADS) Page 9
Comparison to Other PWR Technologies Parameter AP1000 Design EPR Design KONVOI Design Reactor Power (MWt) 3400 4500 3850 Net Electric Output (MWe) 1117 ~1630 1365 Fuel Type UO 2 (or MOX) UO 2 or MOX UO 2 or MOX Fuel Assembly Type 17 x 17 (24) 17 x 17 (24) 18 x 18 (24) Rod Pitch (m) 0.0126 0.0126 0.0127 Active Core Height (m) 4.3 4.2 3.9 Fuel Assemblies (#) 157 241 193 Rod Cluster Control 53 89 61 Assemblies (#) Gray Rod Assemblies (#) 16 0 0 Total Flow Rate(kg/s) 15177 22225 18800 Inlet Temperature (K) 553.82 568.85 564.15 Outlet Temperature (K) 594.26 603.05 597.65 Reactor Operating Pressure (MPa) 15.5 15.5 15.8 Page 10
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 11
ASTEC V2.1.1.1 Accident Source Term Evaluation Code (ASTEC) Developed and maintained solely by IRSN (France) Integral code with modular structure State-of-the-art modeling of severe accident phenomena including 2D magma modelling Page 12
In-Vessel Phenomena ASTEC Modules CESAR Module Thermal hydraulic response in the primary and secondary system Simplified vessel description ICARE Module In-vessel core degradation phenomena Lower head failure Detailed vessel description Page 13
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 14
Model Development Design, material, and geometry data available in open literature Generic data used when AP1000-specifics not available Two models developed, a standard, or half-vessel model and a full vessel model Generic time-dependent decay heat Page 15
Model Development Page 16
Core Radial Nodalization RPV Barrel Shroud Channel Boundary Gray Rod Cluster Assembly Rod Cluster Control Assembly Fuel Assembly Bypass Downcomer Cylindrical Macro-Component Radii (m) 2.24 2.02 1.75 1.70 1.55 1.52 Radial 0.0 Discretization (m) 0.36 0.65 0.95 1.22 1.52 1.70 2.02 ring1 ring2 ring3 ring4 ring5 bypass downcome Page 17
Core Axial Nodalization downcome bypass ring5 ring4 ring3 ring2 ring1 4.27 Height (m) 2.02 1.70 1.52 1.22 0.95 Radius (m) 0.65 0.36 0.0 Page 18
Core Degradation Page 19
Vessel and Primary Nodalization UPPERPLE CL1B1 CL1B2 CL1B3 CL1B4 CL1A4 CL1A3 CL1A2 CL1A1 CL2B1 CL2B2 CL2B3 CL2B4 HLB3 HLB3 HLB2 HLB1 CL2A4 CL2A3 CL2A2 CL2A1 HLA1 HLA2 HLA3 HLA4 Active Core downcome bypass ring5 ring4 ring3 ring2 ring1 LOWERPLE Page 20
Steady State Calculation Parameter ASTEC Design Percent Value Value Error(%) Reactor Power (MWt) 3400 3400 0.0 Core Inlet Flow (per loop, 3794.3 3794.3 0.0 kg/s) Core Inlet Temperature (K) 553.9 553.8 0.02 Core Inlet Pressure (MPa) 15.974 15.927 0.30 Core Outlet Flow (per loop, 7588.7 7590.9 0.03 kg/s) Core Outlet Temperature (K) 594.1 594.3 0.03 Core Outlet Pressure(MPa) 15.501 15.499 0.01 Bypass Flow (%) 5.9 5.9 0.0 Page 21
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 22
Accident Scenario Event Time Flow Temperature Pressure (s) (kg/s) (K) (Mpa) Start of Steady State Calculation -2000 3794.3 553.8 15.513 Start of Transient 0 3794.3 553.8 15.513 Calculation Loss of Flow Begins, SCRAM 100 3794.3 553.8 15.513 Diminished Flow 120 0.37943 553.8 15.513 Loss of Flow 8000 0.0 553.8 15.513 End of Transient 20000 0.0 553.8 15.513 Inlet Flow Parameters Page 23
Core Temperatures Average Fuel Temperatures Average Cladding Temperatures Page 24
Hydrogen and Magma Production Page 25
RPV Temperature Field: End-of-steady state Page 26
RPV Temperature Field: Lower Head Failure Page 27
Overview Background In-vessel Core Degradation Introduction to the Westinghouse AP1000 ASTEC V2.1 Overview AP1000-Like Model Development Accident Simulation and Results Summary Page 28
Summary A simplified, generic AP1000-LIKE model has been developed using the integral severe accident code ASTEC V2.1 The model is evaluated using an accident scenario without AMM, leading to core degradation, simulating loss of flow to the reactor vessel under high pressure conditions. According to current results lower head failure occurs nearly four hours after the reactor SCRAM Approximately 800 kg of H 2 are produced, and approximately 150 tons of molten material form and relocate to the RPV lower plenum Page 29
Summary The impact of modeling upper vessel structures can be large. Thus, description of the upper vessel structures should be applied with caution in future. These preliminary code results show the applicability of the model for simulation of scenarios involving core degradation. Page 30
Future Work Implementation of AP1000-specific decay heat curve and radial/axial power profiles Include advanced materials data for AP1000-specific materials in the model Complete the AP1000 primary and secondary circuits and include all active and passive safety systems Long term goal is to estimate key grace periods relevant for core degradation under a variety of accident scenarios and with different AMM considerations with the aim to accurately characterize the progression of severe accidents in the AP1000 Page 31
Acknowledgements This work was funded by the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) Summer Student Program. We are grateful for the valuable comments provided on contents of the paper by members of the ASTEC team at IRSN. Disclaimer This work was carried out at Helmholtz-Zentrum Dresden- Rossendorf by researchers from the and Helmholtz-Zentrum Dresden-Rossendorf. Westinghouse Electric Company and the US NRC are not responsible for the contents or results of this analysis. Page 32
Questions? Thank you for your attention Page 33
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Average Water Level Page 37
Maximum Fuel Temperature Page 38
H2 Production Rate Page 39
Generic Power Profiles Page 40
Generic Power After SCRAM Page 41