MILITARY INSTITUTE OF SCIENCE AND TECHNOLOGY
B.Sc. Engg. Research Proposal
1. Name of the Student: Sakib Hossain Shium
Roll No: 202228039
Semester: 07
Session: 2022-2023
2. Present Address: Osmany Hall, Mirpur-12, Dhaka-1216
Email: sakibhossain2003@gmail.com
Telephone: 01308674739
3. Name of the Department: Department of Nuclear Science and Engineering
1. Name of the Student: Motiur Rahman
Roll No: 202228030
Semester: 07
Session: 2022-2023
2. Present Address: Osmany Hall, Mirpur-12, Dhaka-1216
Email: motiur30roky@gmail.com
Telephone: 01309398266
3. Name of the Department: Department of Nuclear Science and Engineering
1. Name of the Student: Md. Taiubur Rahman
Roll No: 202228027
Semester: 07
Session: 2022-2023
2. Present Address: Osmany Hall, Mirpur-12, Dhaka-1216
Email:
Telephone: 01816060981
3. Name of the Department: Department of Nuclear Science and Engineering
4. Thesis Title: Thermo-Mechanical Analysis of PWR T-Junctions Under Loss-of-Flow Events
Abstract
Pressurized Water Reactor (PWR) piping systems are subjected to extreme thermal and mechanical
loads during abnormal conditions such as Loss-of-Flow (LOF) events. One of the most vulnerable
regions is the T-junction, where high thermal gradients, stagnation, and flow recirculation can
occur. These conditions can initiate local stress concentration, thermal fatigue, or even creep
rupture, compromising reactor safety. This study aims to develop a computational framework to
assess the thermo-mechanical behavior of PWR T-junctions under transient LOF scenarios. A
coupled analysis using CFD and finite element tools will be performed, supported by uncertainty
analysis in Python. The outcomes will help identify structural risks and contribute toward safer Tjunction design under transient reactor conditions.
1. Introduction and Background
The structural integrity of primary loop piping in Pressurized Water Reactors (PWRs) is critical to
overall reactor safety. T-junction points, where main coolant lines connect to branch pipes, are
particularly susceptible to complex thermal and mechanical loads. These loads become more
pronounced during off-normal scenarios such as Loss of Flow (LOF) events. Under LOF
conditions, coolant velocity may reduce drastically, leading to the formation of thermal
stratification zones, reversed flow phenomena, and steep thermal gradients. These factors induce
significant stresses and thermal fatigue in pipe walls, posing a risk to long-term structural integrity
and aging management [1, 6].
Previous studies have primarily addressed LOF behavior at the system level using codes like
RELAP5 or MELCOR [10], or examined containment stress responses under Loss of Coolant
Accident (LOCA) scenarios [9]. However, there remains a lack of detailed investigations into
localized thermo-mechanical stress development at T-junctions during LOF events. Experimental
and computational fluid dynamics (CFD) studies have demonstrated the presence of nonisothermal vortex flows and complex velocity/temperature distributions at T-junctions [5, 7], yet
their integration with structural finite element models for real-time stress prediction remains
limited.
Moreover, uncertainty quantification has rarely been incorporated into these analyses. Techniques
such as Latin Hypercube Sampling and Wilks’ formula, which are capable of capturing variability
in input parameters and assessing probabilistic failure margins, are often overlooked [4]. This gap
limits the reliability of safety assessments under real-world conditions.
To address these challenges, this research aims to develop a high-fidelity, coupled CFD–finite
element simulation framework using advanced multiphysics platforms like ANSYS Fluent and
COMSOL Multiphysics [2, 3]. The model will evaluate real-time temperature and stress evolution
at T-junctions during LOF conditions, incorporating probabilistic uncertainty analysis to quantify
safety margins. This effort will directly support the IAEA’s recommendations on proactive aging
management and structural integrity evaluation for safety-class piping systems [1, 8].
2. Objectives

To simulate transient flow and heat transfer behavior within a PWR T-junction during a
Loss-of-Flow event using CFD.

To analyze the resulting thermal and mechanical stresses using finite element analysis.

To couple the CFD-generated thermal boundary conditions with structural simulations for
accurate stress prediction.

To assess the influence of uncertainty in operational and material parameters on structural
performance using Python-based sampling methods.
3. Methodology
Step 1: Geometry and Meshing

Develop a detailed 3D model of a standard PWR T-junction (e.g., hot leg to surge line or
cold leg branching).

Use ANSYS Design Modeler or COMSOL geometry tools for model creation.

Mesh the model with refinement at junction zones prone to high gradients.
Step 2: CFD Simulation – ANSYS Fluent or COMSOL Multiphysics


Perform a transient CFD simulation representing an LOF scenario:
o
Sudden drop in flow velocity
o
Rising wall temperatures
o
Buoyancy-driven flow and thermal stratification
Extract time-dependent temperature and pressure fields from pipe walls.
Step 3: Thermo-Mechanical FEA – ANSYS Mechanical or COMSOL

Import the CFD results as boundary conditions.

Apply material properties (e.g., SA-508 or SS-304) with nonlinear temperature-dependent
behavior.

Include:

o
Thermal expansion
o
Creep (if high temperatures persist)
o
Plastic deformation
Simulate stress evolution, deformation, and identify critical regions.
Step 4: Uncertainty Analysis – Python


Use Python + NumPy + pyDOE for Latin Hypercube Sampling (LHS) of:
o
Flow reduction rate
o
Initial wall temperature
o
Material property variation
Apply Wilks’ formula to define sample sizes for 95% confidence.

Automate simulation parameter variation (e.g., using scripts linked with COMSOL API or
ANSYS Workbench).
4. Expected Outcomes

A validated numerical model for analyzing thermo-mechanical effects in T-junctions
during LOF.

Identification of high-risk zones for thermal fatigue or creep failure.

Quantitative evaluation of stress margins under transient and uncertain conditions.

A reproducible methodology for coupled CFD-FEA analysis in nuclear safety research.

Probabilistic safety indicators to support design upgrades or mitigation strategies.
5. Timeline
6. Conclusion
This research addresses a critical need for localized thermo-mechanical evaluation of PWR Tjunctions under LOF conditions, a key aspect of reactor safety that is often overlooked in
traditional system-level analyses. By integrating CFD, FEA, and statistical methods, this study
will provide new insights into structural risk and inform better design and maintenance practices
for nuclear reactors. The proposed methodology also has potential applications in broader thermalhydraulic safety studies beyond the T-junction.
References
1. IAEA-TECDOC-981, Assessment and Management of Ageing of Major Nuclear Power
Plant Components Important to Safety: Piping, International Atomic Energy Agency, 1997.
2. ANSYS Fluent Theory Guide, Release 2023 R1. ANSYS Inc.
3. COMSOL Multiphysics Reference Manual, Version 6.0. COMSOL AB, 2022.
4. Peng, Xintong, et al. “Research on Thermo-Mechanical Response of Solid-State Core
Matrix in a Heat Pipe Cooled Reactor.” Energies, vol. 18, no. 6, 2025, p. 1423,
https://doi.org/10.3390/en18061423.
5. Abdulwahid, Mohammed A., Niranjan K. Injetib, and Sadoun Fahad Dakhil. “CFD
Simulations and Flow Analysis Through a T-Junction Pipe.” ResearchGate, 2010,
https://www.researchgate.net/publication/313502258_CFD_Simulations_and_Flow_
Analysis_Through_a_T-Junction_Pipe.
6. ASME Boiler and Pressure Vessel Code, Section III, Division 1 – Subsection NB, Rules
for Class 1 Components. American Society of Mechanical Engineers, 2021 Edition.
7. Baranova, Tatyana A., et al. “Non-Isothermal Vortex Flow in the T-Junction Pipe.”
Energies, vol. 14, no. 21, 2021, p. 7002, https://doi.org/10.3390/en14217002.
8. Lu, S. C., R. A. Larder, A. L. Chan, and S. M. Ma. “Evaluation of Stress Histories of
Reactor Coolant Loop Piping for Pipe Rupture Prediction.” Proceedings of SMiRT-6,
Lawrence Livermore National Laboratory, 1981.
9. Bang, Youngsuk, et al. “Estimation of Temperature-Induced RCS and SG Tube Creep
Rupture Probability Under High-Pressure Severe Accident Conditions.” Atomic Energy
Society of Japan, vol. 50, no. 5, 2012, pp. 855–872.
10. Sheykhi, Sh., S. Talebi, M. Soroush, and E. Masoumi. “Thermal-Hydraulic and Stress
Analysis of AP1000 Reactor Containment During LOCA in Dry Cooling Mode.” Journal
of Nuclear Engineering and Design, vol. 324, 2017, pp. 123–136.