Comparative Analysis of PWR Core Wide and Hot Channel Calculations ANS Winter Meeting, Washington DC November 20, 2002

1. Comparative Analysis of PWR Core Wide and Hot Channel CalculationsANS Winter Meeting, Washington DCNovember 20, 2002 M. Avramova S. Balzus K. Ivanov R. Mueller L. Hochreiter The Pennsylvania State UniversityFramatome ANP GmbH, Germany

2. OUTLINE • Introduction • COBRA-TF Code • PWR Core Model • Code-to-Code Comparison • Conclusions

3. INTRODUCTION In the framework of joint research program between the Pennsylvania State University (PSU) and Framatome ANP the COBRA-TF best-estimate thermal-hydraulic code is being validated for LWR core analysis As a part of this program a PWR core wide and hot channel analysis problem was modeled using COBRA-TF and compared with COBRA 3-CP PSU COBRA-TF Simulations Framatome ANP COBRA 3-CP Simulations

4. INTRODUCTION COBRA-TF Code - developed to provide best-estimate thermal-hydraulic analysis of LWR vessel for design basis accidents and anticipated transients COBRA 3-CP - used at Framatome ANP as a thermal-hydraulic subchannel analysis and core design code

5. Entrained Liquid Drops COBRA-TF Thermal-Hydraulic Code COBRA-TF Modeling Features COBRA-TF Application Areas PWR Primary System LOCA Analysis LWR Rod Bundle Accident Analysis Two-Fluids Three-Fields Three-Dimensions Continuous Vapor Continuous Liquid

6. Heated Tubes Solid Cylinders COBRA-TF Thermal-Hydraulic Code COBRA-TF Regimes Maps Normal Flow Regime Hot Wall Regime COBRA-TF VESSEL Structures Models Heat-Generating Structures Unheated Structures Nuclear Fuel Rods Heated Flat Plates Hollow Tubes Flat Plates

7. Core Wide Analysis Steady State Anticipated Transients - Flow Reduction - Power Rise - Pressure Reduction Hot Channel Analysis Rod Ejection Accident (REA) TMI-1 Rod Ejection Main-Steam-Line-Break (MSLB) TMI-1 MSLB (Exercise 2) TRAC-PF1/NEM/COBRA-TF COBRA-TF PWR Core Modeling – Background COBRA-TF PWR Core Modeling – Stand Alone and Coupled

8. PWR Core Model • The Simulated PWR Core Contains 121 14x14 FA • The hot assembly is located at the center of the core • A quarter core model was chosen for the COBRA-TF model similar to the COBRA 3-CP model • The sub-channels surrounding the limiting rod were represented on a sub-channel basis • The remaining part of the quarter-core was modeled as lumped channels

9. PWR Core Model Subchannel layout of the macro-cell • The macro-cell is comprised of subchannels 1 through 7 • The subchannels surrounding the limiting rod have been modeled exactly as subchannels 1 through 4 • Surrounding this area are lumped in channels 5, 6, and 7

10. The remaining parts of the four fuel assemblies are modeled as channel 8 • The rest of the quarter core is modeled as channel 9 • 5 Spacer Grids (4 mixing spacers and 1 structural spacer ) • Chopped cosine with a peak value of 1.55Axial Power Profile • Non-uniform Radial Power Profile • Inlet BC - Inlet Flow Rate and • Inlet Enthalpy • Outlet BC - Outlet Pressure PWR Core Model Layout of the ¼ core model Subchannel 9 Subchannel 8 Instrumentation Tubes Macro-cell (Subchannels 1-7)

11. COBRA-TF Modifications • In order to define an identical basis for the comparative analysis two modifications were made to COBRA-TF as code features: • The same correlation for the rod friction factor used in the COBRA 3-CP code was introduced in COBRA-TF • The W3 Critical Heat Flux correlation was also added to the code

12. Channel # 3 Channel # 3 Code-to-Code Comparisons • STEADY STATE • The codes demonstrate steady-state results with excellent agreement • The axial distributions of the mass flow rate, calculated by the two codes differ by only about 1% (on average)

13. Channel # 3 Code-to-Code Comparisons • STEADY STATE • The codes predict a similar DNBR • COBRA 3-CP tends to predict a MDNBR at higher elevation • COBRA-TF - constant “F” factor • COBRA 3-CP - dynamically computed“F” factor

14. Transient Models • Main differences • COBRA 3-CP - the wall heat flux time history is specified as a boundary condition • COBRA-TF - the wall heat flux was calculated from the rod heat conduction solution in the code • Therefore in COBRA-TF the rod power was specified and during a transient the heat flux took into account the stored heat release

15. Transient Models • Solution • These differences between the two transient models for the wall heat flux are eliminated in the following way: • In the COBRA-TF input deck the fuel rods are modeled as tubes with very small thickness of the wall • In this case the generated heat in the fuel rods is neglected • Wall heat flux time history is specified as a boundary condition (in a similar way as in the COBRA 3-CP code)

16. Channel # 3 Code-to-Code Comparisons • 50% Loss of Flow Transient • The maximum heat flux to flow ratio is predicted at two seconds into the transient by both codes and as a result the minimum DNBR is reached at about two seconds into the transient for both code simulations

17. CONCLUSIONS • The PWR core-wide and hot channel analysis problem was modeled with both COBRA 3-CP and COBRA-TF computer codes • Identical modeling basis for rod friction has been defined and the COBRA 3-CP correlation has been implemented into the COBRA-TF source • In COBRA 3-CP the Critical Heat Flux is calculated using the W3 correlation and this correlation was added to the current version of COBRA-TF • Consistent transient surface heat flux boundary conditions were used such that more exact comparisons can be made between the two different code calculations

18. CONCLUSIONS – cont. • Results from the codes show a very good agreement for the initial steady-state conditions as well as for the simulated loss of flow transient • The only difference in the two calculations is the location of the minimum DNBR • This is explained by the fact that in COBRA-TF a constant Tong “F” factor (which accounts for a non-uniform axial power shape) is used while in COBRA 3-CP this “F” factor is dynamically computed