Department of Mechanical and Nuclear Engineering

Reactor Dynamics and Fuel Management Group

Comparative Analysis of PWR Core Wide and Hot Channel Calculations

M. Avramova K. Ivanov L. Hochreiter

The Pennsylvania State University

ANS Winter Meeting, Washington DC November 20, 2002 S. Balzus R. Mueller

Framatome ANP GmbH, Germany

OUTLINE



Introduction



COBRA-TF Code



PWR Core Model



Code-to-Code Comparison



Conclusions 2

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

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 3

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

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 4

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

COBRA-TF Thermal-Hydraulic Code COBRA-TF Application Areas

PWR Primary System LOCA Analysis LWR Rod Bundle Accident Analysis Two-Fluids

COBRA-TF Modeling Features

Three-Fields Three-Dimensions Continuous Vapor Entrained Liquid Drops Continuous Liquid 5

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

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 Tubes Heated Flat Plates Hollow Tubes Solid Cylinders Flat Plates 6

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

COBRA-TF PWR Core Modeling – Background COBRA-TF PWR Core Modeling – Stand Alone and Coupled Core Wide Analysis

Hot Channel Analysis

Steady State Anticipated Transients - Flow Reduction - Power Rise - Pressure Reduction TRAC-PF1/NEM/COBRA-TF Rod Ejection Accident (REA)

TMI-1 Rod Ejection

Main-Steam-Line-Break (MSLB)

TMI-1 MSLB (Exercise 2)

7

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

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 Parameter

Fuel rod outside diameter (in) Guide tube outside diameter (in) Instrumentation tube outside diameter (in) Fuel rod pitch (in) Fuel assembly pitch (in) Fuel assembly dimensions Gap between fuel assemblies (in) Number of fuel rods Number of guide tubes Number of instrumentation rods Fuel active length (in)

8

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

0.424

0.539

0.424

0.556

7.803

14x14 0.020

179 16 1 95.00

PWR Core Model

Subchannel layout of the macro-cell

15 17 8 1 14 10 3 5 2 7 9 6 5 13 

The macro-cell is comprised subchannels 1 through 7 of

6 16 7 

The the subchannels limiting rod surrounding have been modeled exactly as subchannels 1 through 4



Surrounding this area are lumped in channels 5, 6, and 7 9

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

PWR Core Model

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 Tubes Inlet Enthalpy (Subchannels 1-7)



Outlet BC - Outlet Pressure 10

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

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 11

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

Code-to-Code Comparisons

STEADY STATE The codes demonstrate excellent agreement steady-state results with The axial distributions of the mass flow rate, calculated by the two codes differ by only about 1% (on average) Liquid Enthalpy Steady-State Liquid Mass Flowrate Steady-State 660 640 620 600 580 560 540 520 0 Channel # 3 COBRA 3-CP COBRA-TF 0.670

0.660

0.650

0.640

0.630

0.620

20 40 60 Axial Location (in) 80 100 0 20 40 60 Axial Location (in) 12

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

Channel # 3 COBRA 3-CP COBRA-TF 80 100

Code-to-Code Comparisons

STEADY STATE The codes predict a similar DNBR COBRA 3-CP tends to predict a MDNBR at higher elevation 6 5 4 3 2 10 9 8 7 0 DNBR Steady State 20 40 60 Axial Location (in) 80 Channel # 3 COBRA 3-CP COBRA-TF 100 COBRA-TF constant “F” factor COBRA 3-CP - dynamically computed “F” factor 13

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

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 14

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

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) 15

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

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 Minimum DNBR Channel # 3 10 8 6 4 COBRA 3-CP COBRA-TF 2 0 2 4 6 Time (seconds) 8 10 16

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

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 17

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

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 18

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