Conference Agenda

Personal Experiences in the Two Kinds of Experiments to Confirm Reliability of Passive Systems for Nuclear Reactors

Hideo Nakamura

Japan Atomic Energy Agency, Japan

Fission Products Scrubbing in Water Ponds during Severe Accidents: An Efficient Means of Attenuating Source Term

Luis Herranz

CIEMAT, Spain

Heavy Liquid-Metal Pool Thermal-Hydraulic Experiments and Simulations

Katrien D. A. Van Tichelen¹, Steven Keijers¹, Fabio Mirelli¹, Julio Pacio¹, Rui Serrano Diogo¹, Mohammed Muaaz Hussain², Fajar Sri Lestari Pangukir², Dirk Cornelis Visser², Ferry Roelofs², Lilla Koloszar³, Silvania Lopes³, Philippe Planquart³

1: SCK CEN, Belgian Nuclear Research Centre, Belgium; 2: NRG Pallas, the Netherlands; 3: von Karman Institute for Fluid Dynamics, Belgium

Numerical Investigation of Boiling Phenomena and Vibration Instabilities in Helical Cruciform Fuel for Water-cooled SMRs

Youyeon Choi, Koroush Shirvan

Massachusetts Institute of Technology, United States of America

Performance Evaluation and Validation of Load Following in the NuScale Concept Based URI-SMR Experimental Facility

Joo Hyung Seo, In Cheol Bang

UNIST, Korea, Republic of

Preliminary Assessment of Heat Transfer Performance in Helical CoilSteam Generators

Yu Jung Choi, Jaehwan Park, Kyoung-uk Lee

KHNP, Korea, Republic of

Predictability Evaluation of SPACE Code for the Condensation Model in the Nearly Horizontal Tube

Hae Min Park, Jong Hyuk Lee

KAERI, Korea, Republic of

Experimental Study and Optimized Design of Printed-circuit Heat Exchanger with Straight Channel from Laminar to Turbulent Conditions for Recuperators in Gas Brayton Cycles

Sungkun Chung¹, Gibbeum Lee², Jae Eun Cha², HangJin Jo¹

1: Pohang University of Science and Technology (POSTECH), Korea, Republic of; 2: Korea Atomic Energy Research Institute (KAERI), Korea, Republic of

Experimental Study on the Influence of Epoxy Patterns of Bi-conductive Surfaces on Pool Boiling

Longchang Tang, Wei Xu, Yan Liu, Xiaojing Liu

Shanghai Jiao Tong University, China, People's Republic of

Experimental Study of the Nucleate Flow Boiling to Convective Flow Boiling Transition

Carlos Alberto Dorao, Maria Fernandino

Norwegian University of Science and Technology, Norway

Quantitative Heat Flux Prediction from Images Using Convolutional Neural Networks with Multi-Branch Convolutions

Mengqi Wu¹, Zhen Zhang¹, Yongyong Wu¹, Nan Gui¹, Xingtuan Yang¹, Jiyuan Tu^1,2, Shengyao Jiang¹

1: Tsinghua University, China, People's Republic of; 2: RMIT University, Australia

Study of Subcooled Flow Boiling at Low Pressure Conditions Using Eulerian-Eulerian Multiphase Flow Model Coupled with Force Balance Model

Nikhil Chitnavis¹, Harish Pothukuchi^1,2, B.S.V. Patnaik¹

1: Indian Institute of Technology Madras, India; 2: Indian Institute of Technology Jammu, India

Microlayer and Contact Line Dynamics under Different Heat Fluxes at Nucleate Boiling

Corentin Le Houedec¹, Boštjan Zajec¹, Cassiano Tecchio¹, Benjamin Cariteau¹, Pere Roca i Cabarrocas², Pavel Bulkin², Vadim S. Nikolayev³

1: Université Paris-Saclay, CEA, STMF, France; 2: Institut Polytechnique de Paris, Ecole Polytechnique, LPICM, CNRS, France; 3: Université Paris-Saclay, CEA, SPEC, CNRS, France

Two-phase Critical Flow Experiments at LUT University

Lauri Pyy, Juhani Hyvärinen, Markku Puustinen, Giteshkumar Patel

LUT University, Finland

Integral Effect Test and SPACE-CAP Code Calculation for the Transient in the RCS and Containment during Small Break LOCA

Byoung-Uhn Bae, Jae Bong Lee, Yusun Park, Seok Cho, Jun-Yeong Jung, Kyoung-Ho Kang

Korea Atomic Energy Research Institute, Korea, Republic of

Natural Circulation Interruption Phenomena during Asymmetric Cooldown Operation in ATALS Test Facility

Seok Cho, Byoung-Uhn Bae, Yu-Sun Park, Jae-Bong Lee, Jong-Rok Kim, Kyoung-Ho Kang

Korea Atomic Energy Research Institute, Korea, Republic of

System Behavior Analysis of PATRIOT at SBO Scenario: A Scaled-Down IET Facility Using R134a Refrigerant

Ik Jae Jin¹, Do Yeong Lim², In Cheol Bang¹

1: Ulsan National Institute of Science and Technology (UNIST), Korea, Republic of; 2: Texas A&M University, United States of America

Performance of SMART100 Passive Safety System Validated in Thermal-Hydraulic Integrated Effect Test Using SMART-ITL

Hwang Bae, Jin-Hwa Yang, Byong Guk Jeon, Sung Uk Ryu, Hyun-sik Park, Sung-Jae Yi, Yoon Gon Bang

KAERI, Korea, Republic of

Safety Analysis Update at Swiss PWR to Comply with the Revised ENSI-A01 Guideline

Davide Papini¹, Peter Sarkadi¹, Michael Hauck¹, Rainer Kaulbarsch¹, Andrey Fjodorow², Annemarie Müller², Oleksander Bykov², Lars Rossner², Steffen Dauderstädt²

1: NPP Gösgen (KKG), Switzerland; 2: Framatome GmbH, Germany

An Approach to Oscillatory Behaviors in Helical Coil Using a Code Framework

Ceyda Eldem¹, HangJin Jo^1,2

1: Division of Advanced Nuclear Engineering, POSTECH, Korea, Republic of; 2: Department of Mechanical Engineering, POSTECH, Korea, Republic of

NOUS - A Python based Initialization Software for Coupled TH-NK Calculation

Róbert Dániel Rádi¹, Dávid Zagyi¹, József Bánáti¹, Viktor Holl³, István Pós², Zoltán Kálya²

1: Paks II Ltd., Hungary; 2: NESP 2000 Ltd., Hungary; 3: MVM Paks Nuclear Power Plant Ltd., Hungary

Simulation of Representativeness of Airborne Radioactive Materials Sampling in Scaled Model of Nuclear Power Plant Stack

Yixin Huang¹, Jinghui Gao², Lili Tong¹, Xuewu Cao¹

1: Shanghai Jiao Tong University, China, People's Republic of; 2: Ling’ao Nuclear Power Co., Ltd., China, People's Republic of

miscibleSpeciesTransport: A New OpenFOAM-based Framework for Studying Interphase Tritium Transfer in Molten Salt Systems

Benjamin James Li, Guanyu Su

University of California Berkeley, United States of America

POD-Based Reduced Order Modeling of Molten-Salt CFD Simulations

Charlie Owen¹, Ethan Rozak¹, Kevin Clarno¹, Cole Gentry¹, Yassin Hassan², Joseph Seo²

1: University of Texas at Austin, United States of America; 2: Texas A&M University, United States of America

Conjugate Heat Transfer Analysis of the Stable Salt Reactor Fuel Pin Using Spectral Element Simulations

Jun Fang¹, Dillon Shaver¹, Ling Zou¹, Thanh Hua¹, Luke Godfrey², Rory O’Sullivan²

1: Argonne National Laboratory, United States of America; 2: Moltex Energy, Canada

Development of Phase Transition Model a Salt Spill Behavior Analysis Using Moving Particle Semi-implcit Method

Yun Sik Cho, Ye Hwan Chun, JinHo Song, Sung Joong Kim

Hanyang Univ., Korea, Republic of

Development and Validation of Two-Phase Flow Models in MOOSE for Molten Salt Reactor Application

Mengnan Li, Mauricio Tano, Piyush Sabharwall, Joshua E. Hansel

Idaho National Laboratory, United States of America

Design and Experimental Test of an Annular Linear Induction Pump for Driving Lead - Bismuth Eutectic

Shi Leitai, Li Da, Li Huaqi, Chen Sen, Tian Xiaoyan, Luo Xiaofei, Zhu Lei, Chen Lixin

Northwest Institute of Nuclear Technology, China, People's Republic of

CFD Analysis and Optimization of Hydrostatic Bearing Design for Primary Pumps in MYRRHA with Heavy Liquid Metal Coolant

Mikhail Iarmonov¹, Kurt Engelen¹, Kirill Makhov¹, Julio Pacio¹, Dieter Fauconnier²

1: SCK CEN, Belgium; 2: Ghent University, Belgium

Multi-scale Coupled Simulation of E-SCAPE at Steady Operation Conditions

Mohammed Muaaz Mohamed Dilshad Hussain¹, Fajar Sri Lestari Pangukir¹, Dirk Cornelis Visser¹, Julio Pacio², Marek Michal Stempniewicz¹, Ferry Roelofs¹

1: NRG PALLAS, The Netherlands; 2: SCK CEN, Belgium

Experimental and Numerical Research on the T/H Characteristics of Pool-type Natural Circulation with Liquid Lead

Yang Liu¹, Lu Zhang²

1: Lanzhou University of Technology, China, People's Republic of; 2: Lanzhou University, China, People's Republic of

Development and Application of System-CFD Coupled Code on Lead-bismuth Fast Reactor

Sipeng Wang, Zehang Kang, Jin Chen, Liming Hu

Nanjing University of Aeronautics and Astronautics, China, People's Republic of

A Comparative Study of Large Language Model Agents for Data-Driven Critical Heat Flux Prediction

Yang Liu, Zaid Abulawi, Abhiram Garimidi, Lim Doyeong

Texas A&M University, United States of America

Aided Active Learning (AAL) for Enhanced Critical Heat Flux Prediction

Umme Mahbuba Nabila¹, Mohammed I. Radaideh¹, Logan Burnett¹, Linyu Lin², Majdi I. Radaideh¹

1: University of Michigan, United States of America; 2: Idaho National Laboratory, United States of America

Evaluation of the Machine Learning CHF Model Enhanced COBRA-TF Prediction Performance

Congshan Mao, Yue Jin

University of Missouri, United States of America

Prediction of Critical Heat Flux with Hybrid Machine Learning: Uncertainty Quantification and CTF Deployment

Aidan John Furlong¹, Xingang Zhao², Robert K. Salko³, Xu Wu¹

1: North Carolina State University, United States of America; 2: University of Tennessee, Knoxville, United States of America; 3: Oak Ridge National Laboratory, United States of America

A Data-Driven Approach to Critical Heat Flux: An ML-Based Method

Marcello Savini¹, Mattia Massone²

1: UNIBO/ENEA, Italy; 2: ENEA, Italy

Development of Risk Profile for Accident Sequences based on Source Term Analysis

Yunho Kim, Jaehyun Cho

Chung-Ang University, Korea, Republic of

ASNR’s Approaches to Thermal-hydraulics Support Studies for Probabilistic Safety Assessments for French Nuclear Power Plants and Other Facilities

Kimlee Bourgeois, Felix Thomann, Nadia Rahni, Guillaume Kioseyian, Gabriel Georgescu, Emmanuel Raimond

French Authority for Nuclear Safety and Radiation Protection (ASNR), France

Identification of Failure Domain Boundaries of Nuclear Passive Safety System Using Genetic Algorithm

Xicheng Wang, Aleksander Kipiela, Dmitry Grishchenko, Pavel Kudinov

Division of Nuclear Science and Engineering, Royal Institute of Technology (KTH), Sweden

Identifying and Prioritizing Knowledge Gaps for the Safe Deployment of Advanced Technology Small Modular Reactors

Cole Nathaniel Takasugi¹, Peter Yarsky¹, Piyush Sabharwall²

1: United States Nuclear Regulatory Commission, United States of America; 2: Idaho National Laboratory, United States of America

Experimental Study and Modeling of Post-CHF Heat Transfer in Support of LWR Safety Analysis and Licensing Review

Xiaodong Sun¹, Kyung Mo Kim^1,2, Qingqing Liu^1,3, Stephen Bajorek⁴, Joseph Kelly⁴

1: University of Michigan, United States of America; 2: Korea Institute of Energy Technology (KENTCH), Korea, Republic of; 3: Mississippi State University, United States of America; 4: The U.S. Nuclear Regulatory Commission, United States of America

Rapid Microlayer Depletion Induced by Bubble Coalescence in Nucleate Pool Boiling

Sai Raja Gopal Vadlamudi¹, Jinming Zhang¹, Uwe Hampel^1,2, Wei Ding¹

1: Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Germany; 2: Technische Universität Dresden, Germany

Direct Numerical Simulation of Single Bubble Dynamics and Associated Heat Transfer: Sensitivity Analysis on Wall Thermal Properties

Linkai Wei¹, Guillaume Bois¹, Vadim S. Nikolayev², Benoit Gwinner³, Solène Delgado³

1: Université Paris-Saclay, CEA, Service de Thermo-hydraulique et de Mécanique des Fluides, France; 2: Service de Physique de l’Etat Condensé, CEA, CNRS, Université Paris–Saclay, France; 3: Université Paris-Saclay, CEA, Service de recherche en Corrosion et Comportement des Matériaux, France

Dynamical System Scaling Application to Bubble Dynamics

Grant Justin Hendrickson, Qiao Wu

Oregon State University, United States of America

Study on Bubble Nucleation and Growth Behaviors Under Supersaturated Conditions in Nitrogen-pressurized Reactors

Xin Lyu, Junlian Yin, Dezhong Wang

Shanghai Jiao Tong University, China, People's Republic of

Investigating Bubble Motion in Downward Flow through Perforated Plates

Nakjun Choi¹, Donkoan Hwang¹, Siyeh Kim², Jeyeong Cheon², HangJin Jo^1,2

1: Division of Advanced Nuclear Engineering, POSTECH, Korea, Republic of; 2: Department of Mechanical Engineering, POSTECH, Korea, Republic of

Local Bubble Characteristics of Subcooled Boiling Flow in an Annular Channel under a Wide Range of Pressure Conditions

Jinyeong Bak, Dongmin Lim, Dongwon Jeong, Minhong Cho, Sanggyun Nam, Manh Long Doan, Byongjo Yun

Pusan National University, Korea, Republic of

Transient Rod Temperature Distributions on a Rod Bundle Near Critical Heat Flux Measured by Optical Fiber Sensors

Takahiro Arai, Riichiro Okawa, Atsushi Ui, Masahiro Furuya, Tsugumasa Iiyama, Shota Ueda, Kenetsu Shirakawa

Central Research Institute of Electric Power Industry, Japan

Development of Turbulent Mixing Parameter for Subchannel Analysis in Hexagonal Rod Bundles

Kannan Iyer, Roopesh Wadhera

Indian Institute of Technology Jammu, India

Void Fraction Measurement on Flow Boiling in 7x7 Rod Bundle based on the WMS with Rod Electrodes

Hao Xie, Jinbiao Xiong

Shanghai Jiao Tong University, China, People's Republic of

Measurement and Analysis of Interfacial Parameters in Two-Phase Flow in a 5×5 Rod Bundle Channel

Shouxu Qiao, Quanbo Li, Jiaxing Ren, Fangdong Wang, Hongyang Wei, Puzhen Gao, Ruifeng Tian, Sichao Tan

Harbin Engineering University, China, People's Republic of

Effect of Pulsating Flow on Evolution of the Velocity Boundary Layer in a 5×5 Rod Bundle Channel

Shouxu Qiao, Xin Li, Songsheng Gao, Shuangcheng Li, Xiaoyong Yu, Ruifeng Tian, Sichao Tan

Harbin Engineering University, China, People's Republic of

Validation of the Intermediate Heat Exchanger Modelling for Fast Reactors using the CLAUDINA Experimental Data

Clotaire Geffray

French Alternative Energies and Atomic Energy Commission (CEA), France

Numerical Benchmarking of Thermal-hydraulic System Codes on Challenging LFR Transient Scenarios

Nicola Forgione¹, Chiara Robazza¹, Barbara Calgaro³, Donella Pellini²

1: University of Pisa, Dipartimento di Ingegneria Civile ed Industriale (DICI), Italy; 2: newcleo S.p.A., Italy; 3: Framatome, Italy

Comparative Study of GOTHIC and CFD in Predicting Thermal Stratification and Mixing Phenomena of Liquid Metal in TALL-3D

Yun Feng^1,2, Xicheng Wang³, Ruifeng Tian^1,2, Sichao Tan^1,2, Dmitry Grishchenko³, Pavel Kudinov³

1: College of Nuclear Science and Technology, Harbin Engineering University, China, People's Republic of; 2: Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment, Harbin Engineering University, China, People's Republic of; 3: Royal Institute of Technology (KTH), Sweden

Numerical Stability Analysis of Semi-implicit Numerical Algorithm for Lead-bismuth-gas Two-component Two-fluid Model

Congyi Xia, Sixiang He, Li Ge, Jianqiang Shan

Xi'an Jiaotong University, China, People's Republic of

Evaluation of Different Mesh Generation Strategies for a Grid-spaced Fuel Bundle within the Framework of the LFR-T/H Benchmark

Maria Faruoli, Michela Lomio, Florentin Prud'homme, Lilla Koloszar

von Karman Institute for Fluid Dynamics, Belgium

Numerical Analysis on the Non-Isothermal Mixing of Liquid Metal in Y-Junctions with Large Eddy Simulation

Xinyang Guo, Yiwei Huang, Wencang Guo, Xianliang Lei, Haijun Wang, Haohao Zhao

Xi'an Jiaotong University, China, People's Republic of

Simulation of NACIE Benchmark Tests using NekRS

Dillon Shaver¹, Maximiliano Dalinger²

1: Argonne National Laboratory, United States of America; 2: Pennsylvania State University, United States of America

MultiMorph - An Euler-Euler based CFD Framework for Multiphase Flows Combining Resolved and Unresolved Structures

Fabian Schlegel, Richard Meller, Benjamin Krull, Ronald Lehnigk, Dirk Lucas

Helmholtz-Zentrum Dresden - Rossendorf (HZDR), Germany

CFD Analysis of Turbulence and Bubble Size Development in a Vertical Pipe Bubbly Flow under Partial Blockage Condition

Yixiang Liao

Helmholtz-Zentrum Dresden - Rossendorf e.V., Germany

Role of Heat Loss in CFD Simulations of Slow Hydrogen Deflagration

Julius Venckus, Mantas Povilaitis

Lithuanian Energy Institute, Lithuania

CFD Assessment of Pressure Drop in Fuel Elements, Effect of Turbulence Model, Comparison with Experiments

Jan-patrice Simoneau

EDF, France

Potential to Optimize Flow Patterns through the Core for the FRM II Conversion using CFD

Kaltrina Shehu, Ronja Schönecker, Christian Reiter

Forschungs-Neutronenquelle Heinz Meier-Leibnitz, Technical University of Munich, Germany

Multi-scale Capillary Dynamic Heat Transfer Characteristic of Artery Heat Pipes under Reactor Transient Thermal Load

Yugao Ma¹, Suyi Zhang², Yue Liu², Zhifang Qiu¹, Shuhua Ding¹

1: Nuclear Power Institute of China, China, People's Republic of; 2: Chengdu University of Technology, China, People's Republic of

Evaluation of the Thermal-hydraulic Behaviour of a Micro Reactor under Steady-state and DLOFC Conditions

Charl G Du Toit¹, Wayne Boyes², Johan Slabber²

1: North-West University, South Africa; 2: University of Pretoria, South Africa

CFD Thermal Analysis for Primary Heat Exchanger of eVinci™ Nuclear Test Reactor

Hong Xu, Yizhou Yan, Oriol Noguera Oliva

Westinghouse, United States of America

Coupled Neutronic and Thermal Simulations of the eVinci^TM Nuclear Test Reactor

Liping Cao, Vefa Kucukboyaci, Mykola Boychenko, Austin Plivelich, Hong Xu, Yizhou Yan

Westinghouse Electric Company LLC, United States of America

Long Duration Testing of High Performance Sodium Heat Pipe

Zachary Don Sellers, Ilyas Yilgor, Jeremy Hartvigsen, Piyush Sabharwall

Idaho National Laboratory, United States of America

Power Transient Testing of High Performance Sodium Filled Heat Pipe

Zachary Don Sellers, Ilyas Yilgor, Jeremy Hartvigsen, Piyush Sabharwall

Idaho National Laboratory, United States of America

Numerical Investigation on Key Conditions for Metal Containment Vessel of i-SMR during Loss of Coolant Accident

Cheolwoong Kim¹, Jae Hyung Park¹, Hyo Jun An¹, Ok Hyeon Lee¹, Sang Ho Kim³, JinHo Song¹, Joon Eon Yang¹, Sung Joong Kim^1,2, Seokgyu Jeong³, Joon Young Bae¹, Jaehyun Ham³

1: Department of Nuclear Engineering, Hanyang University, Korea, Republic of; 2: Institute of Nano Science & Technology, Hanyang University, Korea, Republic of; 3: Korea Atomic Energy Research Institute, Korea, Republic of

Simulation and Analysis of Containment Behavior during Selected Severe Accident Transients in a Generic Konvoi-type PWR using COCOSYS

Miriam Müer¹, Michael Klauck², Stephan Kelm², Marco K. Koch¹

1: Ruhr-Universität Bochum (RUB), Germany; 2: Forschungszentrum Jülich GmbH (FZJ), Germany

Experiments on Spray System Efficiency and Performance under Severe Accident Conditions

Kevin Marc Dieter¹, Martin Freitag¹, Eike Wolfgang Schmidt¹, Benjamin von Laufenberg¹, Gert Langrock², Friedhelm Funke²

1: Becker Technologies GmbH, Germany; 2: Framatome GmbH, Germany

Impact of the Containment and Reactor Pool Modelling on the Evolution of a Severe Accident in a SMR using ASTEC 3.1

Mirco Di Giuli, Ali Mickael Swaidan, Shizhen Yu, Simon Verdebout, Paul Foucaud

Tractebel (ENGIE), Belgium

Sensitivity Analysis and Experimental Validation of the WISDOM Mecanistic Spray Model in Nuclear Reactor Containments

Antoine Ciechocki, Benoit Normand, Arnaud Chazottes

Université Paris-Saclay, CEA, France

Comparison between System Thermal Hydraulic and CFD Analyses of Atmospheric Mixing in the Dome of a Generic PWR Containment during a Severe Accident Transient

Fajar Sri Lestari Pangukir, Dirk Cornelis Visser, Marek Michal Stempniewicz, Nicolaas Bernardus Siccama, Sina Tajfirooz

NRG PALLAS, Netherlands, The

Design and Development of AI-Driven Autonomous Control Testbed for Small-Scale Advanced Nuclear Reactors

Doyeong Lim, Zavier Ndum Ndum, Yassin Hassan, Yang Liu

Texas A&M University, United States of America

Integrated CFD Model for Entire eVinci™ Test Reactor Thermal System

Hong Xu, Yizhou Yan, Oriol Noguera Oliva, Liping Cao

Westinghouse, United States of America

Development and Evaluation of a Machine Learning-based Drift-flux Model in Molten Salt Bubbly Flow

Tristan Reynaldo Villarreal, Ethan Rozak, Cole Gentry, Benjamin Collins, Kevin Clarno

University of Texas at Austin, United States of America

Preliminary CFD Benchmarking of Machine Learning Algorithms for Korean Virtual Small Modular Reactor

Minseo Lee¹, Joongoo Jeon¹, Sangam Khanal¹, Shilaj Baral¹, Chaehyeon Song², Minseop Song², Sunwoong Yang³, Namwoo Kang^3,4, Heesun Choi⁵, Yonggyun Yu⁵

1: Jeonbuk National University, Korea, Republic of; 2: Hanyang University, Korea, Republic of; 3: Korea Advanced Institute of Science and Technology, Korea, Republic of; 4: Narnia Labs, Korea, Republic of; 5: Korea Atomic Energy Research Institute, Korea, Republic of

Generative AI-based Agentic Framework for the Formulation of Phenomena Identification and Ranking Table for Advanced Reactor Application

Paridhi Athe, Nam Dinh

North Carolina State University, United States of America

Application of Intelligent Design Technologies for Nuclear Reactor Thermal-Hydraulics

Qi Lu^1,2, Qingyu Huang^1,2, Baotong Li³, Jian Deng^1,2, Wei Zeng^1,2, Zhifang Qiu^1,2, Jia Liu^1,2, Wenbin Han^1,2

1: State Key Laboratory of Advanced Nuclear Energy Technology, China, People's Republic of; 2: Nuclear Power Institute of China, China, People's Republic of; 3: Key Laboratory of Education Ministry for Modern Design & Rotor-Bearing System, Xi’an Jiaotong University, China, People's Republic of

Application of an FSI Approach based on Structural Reduced-order Model for the Analysis of Flow-induced Vibrations in Nuclear Power Plants

Angel Aleksandrov Papukchiev, Hemish Mistry

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Germany

High Resolution Fluid Structure Interaction Simulation of the GOKSTAD Tube Bundle

Trevor C. Franklin, Lane B. Carasik

Virginia Commonwealth University, United States of America

Flow-Induced Vibrations Simulations involving Structural Contact

Kevin Zwijsen, Ferry Roelofs

NRG PALLAS, Netherlands, The

Numerical Analysis of Flow-Induced Vibrations in Turbulent Two-Phase Cross-Flows Using a Two-Fluid Approach

Sina Tajfirooz, Kevin Zwijsen, Victor Habiyaremye, Edo Frederix, Ferry Roelofs

Nuclear Research and Consultancy Group (NRG), Netherlands, The

Vibration Analysis of a Rotating Propeller in Lead-Bismuth Eutectic through Fluid-Structure Interaction Simulation

Kamiel Ceusters^1,3, Mikhail Iarmonov¹, Kirill Makhov¹, Julio Pacio¹, Tom Verstraete², Joris Degroote³

1: Belgian Nuclear Research Centre (SCK CEN), Belgium; 2: Von Karman Institute for Fluid Dynamics (VKI), Belgium; 3: Ghent University, Belgium

Results from the New GOKSTAD Water Loop Facility for Fluid-Structure-Interaction Studies

Philippe Planquart, Muhsin Can Akkurt, Mathieu Delsipee, Lilla Koloszar

von Karman Institute, Belgium

Comparative Analysis of RELAP5 and STARCCM+ Simulations for Microchannel Heat Exchangers: A Case Study of the E-SMR Primary Heat Exchanger

Andrea Missaglia^1,4, Giorgio Khalil Youssef², Marco Caramello³, Emilio Baglietto⁴, Fabio Giannetti², Stefano Lorenzi¹, Marco Ricotti¹

1: Politecnico di Milano, Italy; 2: Sapienza University of Rome, Italy; 3: Ansaldo Nucleare, Italy; 4: Massachusetts Institute of Technology, United States of America

Coupled Multi-physics Simulation of Full-core Operation of Soluble-boron-free SMRs Using Sub-channel CFD and SERPENT

Bryan Tan¹, Yu Duan¹, Michael J Bluck¹, Matthew D Eaton¹, Bo Liu²

1: Imperial College London, United Kingdom; 2: Science and Technology Facilities Council, Daresbury Laboratory, United Kingdom

Numerical Evaluation of Parameters Influencing Mixing Characteristics under Boron Dilution Transient in a Scaled PWR Downcomer and Lower Plenum

Lilia Djebara, Hongyang Wei, Sichao Tan, Ruifeng Tian

Harbin Engineering University, China, People's Republic of

Development of the Coupled Code TRACE/PARCS/TWOPORFLOW for SMR Safety Analysis

Alejandro Campos-Muñoz¹, Victor Hugo Sanchez-Espinoza¹, Kanglong Zhang¹, Cesar Queral²

1: Karlsruhe Institute of Technology, Germany; 2: Universidad Politecnica de Madrid, Spain

Efficient Prediction of Turbulent Cross Flow in Helical Coil Steam Generators of SMR via Deep Learning–Driven Reduced-Order Models

Bumjin Cho, Minseop Song

Hanyang University, Korea, Republic of

Utilizing Thermal Inertia of Bedrock as a Passive Heat Sink for Small Modular Reactor

Aleksandr Pomogaev, Juhani Hyvärinen, Giteshkumar Patel, Juhani Vihavainen

Lappeenranta-Lahti University of Technology LUT, Finland

Numerical Simulation for Effects of Steam Mass Fraction on Condensation Heat Fluxes from Saturated Steam and Air Mixture on a Vertical Flat Plate

Michio Murase, Toshiya Takaki, Koji Miyoshi

Institute of Nuclear Safety System, Inc., Japan

Experimental Study on Vibration Characteristics of Heat Transfer Tube during Condensation

Haoyu Ruan^1,2, Dongyang Li^1,2, Shiyuan Jiang^1,2, Xue He^1,2, Ruifeng Tian^1,2, Sichao Tan^1,2

1: Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment, Harbin Engineering University, China, People's Republic of; 2: Key Laboratory of Nuclear Safety and Advanced Nuclear Energy Technology, Ministry of Industry and Information Technology, Harbin Engineering University, China, People's Republic of

Integrating Subcooled Flow Boiling and Film Condensation in CFD Modeling of Loop Thermosiphon

Quoc-Tuan Le, Byoung Jae Kim

Chungnam National University, Korea, Republic of

Potential Retention of Fission Products in a Two-phase Flow Problem: Focus on the Hydrodynamics in Pool Scrubbing

Fouzia Djeriouat^1,2, Catherine Marchetto¹, Philippe Nerisson¹, Maxime Chinaud², Jean Denis¹, Olivier Vauquelin²

1: Autorité de Sûreté Nucléaire et Radioprotection (ASNR), France; 2: Institut Universitaire des Systèmes Thermiques Industriels (IUSTI), France

Visulization Study of Film-wise Condesation Heat Transfer on Helical Tubes

Myeonggi Cha, HangJin Jo

POSTECH, Korea, Republic of

Validation Analyses of Natural Circulation and Condensation for MELCOR 2.2 based on PANDA Facility Experiment

Mateusz Michal Malicki, Terttaliisa Lind

Paul Scherrer Institut, Switzerland

Subcooled Flow Boiling Characteristics in a Hexagonal Fuel Sub-assembly with Plate Type Spacers Operating at Low Pressure Conditions

Ayush Kumar Rao, Harish Pothukuchi

Indian Institute of Technology Jammu, India

Further Verification of the High-precision Subchannel Program ATHAS-H

Kui Gao, Yixiang Zou, Jianqiang Shan

Xi'an Jiaotong University, China, People's Republic of

The Estimation of Uncertainty based on Various Fuel Burn-Up Condition in Loss of Coolant Accident Using SUPER Code

Seung-Chan Lee

Korea Hydro & Nuclear Power Co. Central Research Institute, Korea, Republic of

NCGDENSE Program: Advancing the Understanding of Non-Condensable Gases in Nuclear Reactor Coolant Systems

Sharandeet Singh, Giteshkumar Patel, Lauri Pyy, Juhani Hyvärinen

LUT University, Finland

BWR-6 LOCA Modeling with TRACE

Konstantin Nikitin, Ivor Clifford, Hakim Ferroukhi

PSI, Switzerland

Evaluation of the Effect of Flow Channel Deformation with Ballooning of Multiple Fuel Rods in Bundle during LBLOCA

Yunseok Lee, Taewan Kim

Incheon National University, Korea, Republic of

Validation of the Thermal Stratification Behavior for the Sodium-cooled Fast Reactor

Fatih Sinan Sarikurt, Selim Kuran, Mike Steer

TerraPower, United States of America

Sensitivity Analysis of Neutron Cross-Sections and Its Impact on Neutronics-Thermal Hydraulics Coupling in Advanced Sodium-Cooled Fast Reactors

Oyeon Kum

Institute for Energy Conversion and Safety System, Korea, Republic of

Investigation of Surface Temperature Measurement Discrepancies of Capillary Held Fiber Optic Sensors for Experiments on Cladding Failure Propagation in Liquid Metal Fast Reactors Using Water and Air

Sophia RS Jones¹, Jacob Sorensen¹, Guillaume Mignot¹, Aydin Karahan², Samual Miller³, Wade Marcum¹

1: Oregon State University, United States of America; 2: Argonne National Laboratory, United States of America; 3: TerraPower LLC, United States of America

Two-phase Flow Simulations of Gas Entrainment

Ziad Hamidouche¹, Simon Li², Fabienne Delage²

1: Commissariat a L'energie Atomique, Centre de Saclay, France; 2: Commissariat a L'energie Atomique, Centre de Cadarache, France

Numerical Study on Circumferential Temperature of Fuel Rods in Rods Bundles of Liquid Metal Cooled Fast Reactors

Jiahao Zhao¹, Bin Han¹, Xiaoliang Zhu¹, Bao-Wen Yang²

1: Southeast University, China, People's Republic of; 2: DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer, China, People's Republic of

Development of a Coarse-grid CFD Model within the Nuclear Reactor Fuel Assembly

Jeggathishwaran Panisilvam¹, Shuisheng He¹, Bo Liu², Jean-Marie Le Corre³

1: University of Sheffield, United Kingdom; 2: Science and Technology Facilities Council (STFC), United Kingdom; 3: Westinghouse Electric Sweden AB, Sweden

CFD Study of Spacer-grid Induced Turbulence in Typical Pressurized Water Reactor

Tu Guang Tan¹, Nasr Mahmoud², Md Abdur Razzak², Sicong Xiao¹

1: Singapore Nuclear Research and Safety Initiative, Singapore; 2: Temasek Laboratories, Singapore

CFD for Nuclear Safety Studies: Challenges and Related Activities within WGAMA CFD-Task Group

Pierre Ruyer¹, Stephan Kelm², Jan-Patrice Simoneau³, Mingjun Wang⁴, Martina Adorni⁵

1: IRSN, France; 2: Forschung Zentrum Juelich, Germany; 3: EDF, France; 4: Xi'an Jiaotong University, China, People's Republic of; 5: NEA, France

Development and Applications of the Computational Fluid Dynamics Code for Nuclear Reactors-WINGS-CFD

Wei Zeng, Zonglan Wei

Nuclear Power Institute of China, China, People's Republic of

Coarse-mesh CFD Simulations of Rod Bundle Flows

Ezequiel Oscar Fogliatto, Ivor David Clifford

PSI Center for Nuclear Engineering and Sciences, Switzerland

Condition-Adaptive Anisotropic Eddy Viscosity Turbulent Model for Subchannels in Rod Bundles with Specific Mixing Vane Grid

Siwei Qi¹, Bin Han¹, Xiaoliang Zhu¹, Bao-Wen Yang², Jiahao Zhao¹

1: Southeast University, China, People's Republic of; 2: DEQD, China, People's Republic of

High-Pressure Experimental Analysis of Thermal Effects on Near-Wake Turbulence and Energy Distribution for Flow Over a Heated Sphere

Reynaldo Chavez¹, N. K. Anand¹, Yassin Hassan²

1: Department of Mechanical Engineering, Texas A&M University, United States of America; 2: Department of Nuclear Engineering, Texas A&M University, United States of America

Recent Improvements in Pronghorn for Advanced Reactor Modeling

Ramiro Oscar Freile, Mauricio Tano Retamales, Peter German, David Reger

Idaho National Laboratory, United States of America

GAMMA+ Modelling Method on the High Temperature Test Facility Benchmark Problems

Sung Nam Lee, Nam-il Tak, Hong-sik Lim, Chan Soo Kim

KAERI, Korea, Republic of

Towards High-Precision Optical Measurement in Large Pebble Beds for CFD-Grade Experiments: Error Factors and First PIV Results

Zachary Welker¹, Annalisa Manera^2,3, Victor Petrov^1,2,3

1: University of Michigan, United States of America; 2: ETH Zurich, Switzerland; 3: Paul Scherrer Institut, Switzerland

Research of Bypass Flows in Vertical Gaps between Side Reflectors in HTR-PM Using Thermal Hydraulic System Code GCR

Zhao Wu¹, Yingjie Hua¹, Pan Wu¹, Jianqiang Shan¹, Wei Liu²

1: Xi’an Jiaotong University, China, People's Republic of; 2: Huaneng Nuclear Energy Technology Research Institute, China, People's Republic of

Multiphysics Modeling of Radiation-Induced Changes in Graphite for High-Temperature Gas-Cooled Reactors

Youchan Kim¹, Yonghyun Na¹, Geunyong Choi², Jee Hyun Seong¹

1: KAIST, Korea, Republic of; 2: Hanyang University, Korea, Republic of

Efficiency of Electrostatic Precipitator for Removal of Iodine-Containing Particles in Severe Accident Scenarios

Satish Basnet, Atte Jäntti, Olli Sippula, Anna Lähde

University of Eastern Finland, Finland

What is Important and What is Less Important in the Studies of Containment Iodine Behavior at the Severe Accidents

Adolf Rydl¹, Jungsook Clara Wren²

1: INSET s.r.o., Czech Republic; 2: McMaster University, Canada

Experimental Study on the Removal of Iodine Vapor and Iodine-Based Aerosols Inside the Containment of Nuclear Power Plants

Jialong Li^1,2, Jiaxuan Tang^1,2, Luteng Zhang^1,2, Zaiyong Ma^1,2, Wan Sun^1,2, Longxiang Zhu^1,2, Liangming Pan^1,2

1: Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University, China, People's Republic of; 2: Department of Nuclear Engineering and Technology, Chongqing University, China, People's Republic of

Investigation of Source Term During Severe Accidents in Integral PWR SMRs

Govatsa Acharya, Dmitry Grishchenko, Pavel Kudinov

KTH Royal Institute of Technology, Sweden

Uncertainty and Sensitivity Analysis of Radioactive Source Terms from Intact Containment Category for Nuclear Power Plants Based on the BEPU Method

Jing Sun, Shengsheng Lin, Xuhao Zhang, Yaru Li

China Nuclear Power Engineering Co. LTD., China, People's Republic of

Multiphysics Coupling in Nuclear Reactors: A Physics-Informed Neural Network Framework

Hongtao Bi^1,2, Meiqi Song^1,2, Xiaojing Liu^1,2

1: Shanghai Digital Nuclear Reactor Technology Integration Innovation Center, China, People's Republic of; 2: Shanghai Jiao Tong University, China, People's Republic of

Efficient Reliability Assessment of EHRS in IRIS Reactor under LOCA using Physics-Informed Neural Networks (PINNs)

Xi Yuan, Lanxin Gong, Tianze Bai, Changhong Peng

University of Science and Technology of China, China, People's Republic of

Developing a Real-time Surrogate Model with POD-PINN approach for the Flow Field in the Main Coolant Pump of Lead-Cooled Fast Reactors

Yiwen Guo¹, Liangxing Li¹, Kailin Wang^1,2, Zutao Xiang¹, Jiabin Gui¹, Shang Shi¹, Zhenxin Lei¹, Xiangyang Xu¹, Xiaorong Jia¹

1: Xi’an Jiaotong University, China, People's Republic of; 2: Tokyo Institute of Technology, Japan

Sparsity-Promoting Regularization and Uncertainty Quantification for Heat Source Identification in Conjugate Heat Transfer System

Haeseong Kim¹, Sacit Cetiner^2,3, Matteo Bucci¹

1: Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, United States of America; 2: Idaho National Laboratory, United States of America; 3: INL/MIT Center for Reactor Instrumentation and Sensor Physics, MIT Nuclear Reactor Laboratory, United States of America

Toward Improved NekRS Case Setup Workflow with AI-powered Chatbot

Tri Nguyen¹, Luiz Aldeia Machado¹, Elia Merzari¹, Arsen Iskhakov²

1: Penn State University, United States of America; 2: Kansas State University, United States of America

High-fidelity Temperature Field Prediction for Upper Plenum and Hot Legs of PWR Using CFD and Non-intrusive Reduced-order Method

Yanjie Kang¹, Qingyu Huang², Yuan Zhou¹, Yuan Yuan¹

1: Sichuan University, China, People's Republic of; 2: Nuclear Power Institute of China, China, People's Republic of

Development of Transient Heat Pipe Reactor Modeling Capabilities in a Fully-implicit Solution Framework

Kaiyuan Zhu, Guojun Hu

University of Science and Technology of China, China, People's Republic of

Impact of Void Distribution Resolution on Neutronics in Plate-type Reactors with RMC

Ying He, Yongtang Wang, Guodong Liu, Shanfang Huang, Kan Wang

Tsinghua University, China, People's Republic of

Study on Brayton Cycle Start-up and Load Tracking Operation Characteristics of Heat Pipe Reactors

Xufan Han, Li Ge, Jianqiang Shan

Xi’an Jiaotong University, China, People's Republic of

Transient Response of Screen Wick Heat Pipes to Pulsed Dryout Conditions

Trevor Alan Melsheimer, Hansol Kim, Joseph Seo, Yassin Hassan

Texas A&M University, United States of America

Composite Wick Heat Pipe Design for High Power Experiments with Comparison to Past Experiments

Scout Bucks, Chris Balbier, Saya Lee

The Pennsylvania State Univeristy, United States of America

Parametric Experiment and Modeling Analysis of Geyser Boiling Phenomena in Sodium Heat Pipes

Zilin Su, Zeguang Li*, Huifu Wang, Dacai Zhang, Henglong Lin

Tsinghua University, China, People's Republic of