Boundary conditions for FLICA3 and COBRA 3-CP benchmark

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(No Transcript)
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OUTLINE

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

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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
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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

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COBRA-TF Modeling Features
COBRA-TF Thermal-Hydraulic Code
COBRA-TF Application Areas
PWR Primary System LOCA Analysis
LWR Rod Bundle Accident Analysis
Two-Fluids
Three-Dimensions
Three-Fields
Continuous Vapor
Continuous Liquid
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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
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COBRA-TF PWR Core Modeling Background
COBRA-TF PWR Core Modeling Stand Alone and
Coupled
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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
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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

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PWR Core Model
Layout of the ¼ core model
Subchannel 9
Subchannel 8
Instrumentation Tubes
Macro-cell (Subchannels 1-7)
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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

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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)
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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
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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
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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)

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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
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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

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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