Department of Mechanical and Nuclear Engineering Reactor Dynamics and Fuel Management Group Comparative Analysis of PWR Core Wide and Hot Channel Calculations ANS Winter Meeting, Washington DC November 20, 2002 M. AvramovaS. Balzus K. IvanovR. Mueller L. Hochreiter The Pennsylvania State University Framatome ANP GmbH, Germany

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 2 OUTLINE  Introduction  COBRA-TF Code  PWR Core Model  Code-to-Code Comparison  Conclusions

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 3 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 INTRODUCTION PSU COBRA-TF Simulations COBRA-TF Simulations Framatome ANP COBRA 3-CP Simulations COBRA 3-CP Simulations

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 4 INTRODUCTION COBRA-TF Code - developed to provide best-estimate thermal-hydraulic analysis of LWR vessel for design basis accidents and anticipated transients 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

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 5 COBRA-TF Thermal-Hydraulic Code COBRA-TF Application Areas COBRA-TF Modeling Features Two-FluidsThree-DimensionsThree-Fields Continuous Vapor Continuous Liquid Entrained Liquid Drops PWR Primary System LOCA Analysis LWR Rod Bundle Accident Analysis

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 6 COBRA-TF Thermal-Hydraulic Code COBRA-TF Regimes Maps COBRA-TF VESSEL Structures Models Normal Flow Regime Hot Wall Regime Heat-Generating Structures Unheated Structures Nuclear Fuel Rods Heated Tubes Heated Flat Plates Hollow Tubes Solid Cylinders Flat Plates

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 7 COBRA-TF PWR Core Modeling – Background COBRA-TF PWR Core Modeling – Stand Alone and Coupled Core Wide Analysis Steady State Anticipated Transients - Flow Reduction - Power Rise - Pressure Reduction Hot Channel Analysis TRAC-PF1/NEM/COBRA-TF Rod Ejection Accident (REA) TMI-1 Rod Ejection Main-Steam-Line-Break (MSLB) TMI-1 MSLB (Exercise 2)

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 8 PWR Core Model The Simulated PWR Core Contains x14 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

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

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 10 PWR Core Model Macro-cell (Subchannels 1-7) Subchannel 8Instrumentation Tubes Subchannel 9 Layout of the ¼ core model  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 1.55 Axial Power Profile  Non-uniform Radial Power Profile  Inlet BC - Inlet Flow Rate and Inlet Enthalpy Inlet Enthalpy  Outlet BC - Outlet Pressure

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 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: 1.The same correlation for the rod friction factor used in the COBRA 3-CP code was introduced in COBRA-TF 2.The W3 Critical Heat Flux correlation was also added to the code

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 12 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) Channel # 3

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 13 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 Channel # 3

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

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

ANS Winter Meeting, Washington DC, November 20, 2002, Comparative Analysis of PWR Core Wide and Hot Channel Calculations 16 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 Channel # 3

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

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