The Role of Computational Fluid Dynamics & Safety
System Codes for Nuclear Reactor Predictions

Yassin A Hassan
Department of Nuclear Engineering
Texas A&M University

Panel Session - The Role of Thermal Science for Nuclear Disaster Resilience
The 15th International Heat Transfer Conference
Kyoto, Japan, August 10-15, 2014

Why pursue advanced modelingand simulation capabilities?
 When integrated with theory and experiment, modeling &simulation enhances opportunities for new insights into thecomplex phenomena occurring in the nuclear reactor

 Advanced modeling & simulation offers the ability to improve theperformance and safety of nuclear energy; computational ofCFD/system codes provides new capabilities & tools fordoing so These advancements can be utilized in smart simulators – will
impact the decisions of management reactor accidents.

Experimentation – Extremely Difficult & Expensive
Measurement Techniques, Sensors, Instrumentation
not yet Developed- Measurement Volume Inaccessible, Small for
Intrusive/Non-intrusive measurement- Intermittent/Unsteady, Multicomponent,
Multiphase Flow
- Hostile Environment: High Temperature,
Contaminants (Dirt), radioactive
CFD Appears to be a Logical Scheme to ComplementExperimentation
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Introduction

The last decade has seen an increasing use of
three-dimensional CFD codes to predict steadystate and transient flows in nuclear reactors. (ACFD hurricane hits nuclear industry!!!).
The reason for the increased use of multidimensional
CFD methods is that a number of importantphenomena such as pressurized thermal shocks, boron
mixing and thermal striping cannot be predicted bytraditional one-dimensional system codes with therequired accuracy and spatial resolution.
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CFD codes contain empirical models for
simulating turbulence, heat transfer, multiphase flows, and chemical reactions. Suchmodels must be validated before they canbe used with sufficient confidence in
Nuclear Reactor Safety applications.

5Beyond Design Basis Accident• Accidents occurring outside the NPP designbasis• It may or not involve core degradation• Severe Accident event: a BDBA involving coredegradation (melting of reactor core)Accident Management• Actions taken during evolution of events
leading to a BDBA
– Prevent evolution towards severe accident plantconditions– Mitigate consequences of a severe accident– Achieve long term safe stable stateSevere accident phenomena• In-vessel– Core heat-up and uncovering– Fuel degradation– Core debris relocation in lower head
– RPV failure– Etc.

• Ex-vessel– Primary containment pressurization– Primary containment / core debris interaction– Fission product transport– Hydrogen distribution– Primary containment failure– Etc.

BDBA Code General Classification• Mechanistic– Best estimate phenomenological models as muchas possible– Also commonly known as Detailed Codes• Parametric– Combination of phenomenological and user
defined parametric models– Also commonly known as Fast Running Codes
Note: the codes commonly referred as to Special or
Dedicated Codes belong to either of the above.

Thermal-Hydraulic Codes

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Detailed Codes• detailed information of performance of different plantcomponents and systems during BDBA and severe accident,core degradation and/or containment performance, fission
product transport, etc.
• ATHLET-CD (Analysis of Thermal-Hydraulics of Leaks andTransients – Core Degradation): GRS, Germany.
• ICARE/CATHARE (Code Avancé de THermohydraulique pourles Accidents de Réacteurs à Eau): IRSN, France.
• RELAP/SCDAP (Reactor Excursion and Leak Analysis Program /Severe Core Damage Analysis Package): USNRC/INEEL, USA.
• COCOSYS (Containment Code System): GRS, Germany.
• SAMPSON (Severe Accident analysis code with Mechanistic,
Parallelized Simulations Oriented towards Nuclear field),
NUPEC, Japan.
• CONTAIN: USNRC/SNL, USA.
• VICTORIA: USNRC/SNL, USA.

Parametric Codes• Close to real time simulations. Applications to support Level 2PSA, validation of accident management strategies, andsevere accident prevention and mitigation actions.
• ASTEC (Accident Source Term Evaluation Code): IRSN/GRS,France/Germany.
• MELCOR (Methods for Estimation of Leakages andConsequences of Releases): USNRC/SNL, USA.
• MAAP (Modular Accident Analysis Program): EPRI, USA.

Dedicated Codes• Deal with a single specific phenomenon. Several of them arebased on CFD techniques.
• MC3D. Thermo-hydraulic multiphase flow code used forsteam explosion due to fuel-coolant interaction, IRSN, France.
• GASFLOW. Hydrogen transport, mixing and combustion incontainments, IKET/KIT, Germany.
• GOTHIC. 0 – 3D single and multiphase thermalhydraulics,NAI/EPRI, USA.
• TRIO. Single and two-phase thermalhydraulics, CEA, France.

MotivationImagesfrom
http://www.casl.gov/
Multiphase flow
measurements

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Capabilities of Virtual Environment for Reactor
Application (VERA) for modeling and simulation

Multiphysics / Multiscale DevelopmentApproach

CFD Analysis Procedure
1: Understand the Problem and Formulate the Simulation Strategy
2: Create the Geometry
3: Generate the Mesh
4: Establish the Boundary and Initial Conditions
5: Run the Solver, and Monitor for Convergence
6: Post-Process and Interpret the Results
7: Refine the Mesh98

8: Perform a Sensitivity Analysis

79: Document the Analysis

654321

DSTO-TR-1633
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The successive steps for establishing and solving the equations

V&V

Verification & Validation (V&V)
Verification•RealWorld

Code
Validation

Model
Qualification

••

V&Vof
CFD code
Computer
Code(AIAA guide, 1998)

Validation
Mathematical
Model

Discrete Numerical
Approximation

Code
Verification

Verification is the assessment of theaccuracy of the solution to acomputational model by comparison withknown solutions.
The relationship of the simulation to thereal world is not an issue.
Verification is primarily a mathematics
issue.

•••

Validation is the assessment of theaccuracy of a computational simulation bycomparison with experimental data.
The relationship between computation andthe real world, i.e., experimental data, isthe issue.
Validation is primarily a physics issue.

V&V in CFD
Csni-r2007-13

NRS problems requiring CFD with/without coupling to system codes (1/2)

V&V in CFD
NRS problems requiring CFD with/without coupling to system codes (2/2)

RELAP5-3D Nodalization Diagram (1D Model)

RELAP5-3D Nodalization Diagram (3D Model)

Realistic Cold and Hot
Legs Layout (from CADDrawings)

193 Fuel Channels
withcrossflow
individuallysimulated

MELCOR Nodalization DiagramAnalysis of Core Blockage Scenarios
Preliminary Analysis (1D Core)
Purpose:• Identify possible alternative flow paths for core cooling• Identify critical scenarios that may lead to core damage
Conditions:• Instantaneous full core and core bypass blocked at sump switchover• No core cross flow (1D model)

Larger breaks in coldleg may lead to coredamage

Systemenvelope
DBA

Calculational
envelope

Diagram diagram of system and calculation envelopes

BDBA
Severe Accidents
Envelope

Incorporate lessons learned fromrecent experience and previous works+ CFD/System Code Calculations +Experimental Data + ….. = BIG DATABIG DATASMART SIMULATOR

neural networks can be utilized to demonstrate thefeasibility of ANNs to predict the event progress
based on the available data
Mitigating Strategies for Beyond Design Basis Events• Change operational mode to avoid severe states• Alter the unfavorable events (How we get here?)• Interpretation and collection of the old andrecent data & benchmarks & Identify
uncertainty• Mitigate the consequences using simulators• Lower the cost

• Conduct many analyses to identify vulnerability• Support / use predictive multi-scale multiphysics modeling and simulation

Nuclear Power is Different Because Accidents areRare, But Can be High-Consequence

Plants Must Operate at a High Level of Safety for
60 years – perhaps longer
Most Difficult of all Industrial Activities to Explain
to the General Public

Success will be achieved
international collaborationby