Conference Agenda

Mass and Energy Release Analysis for Postulated Main Steam Line Break Accident in APR1000 Using SPACE-ME Code

Jisu Kim, Sung Yong Kim, Ye Eun An, Sang Jin Lee, Seong Ho Jee, Min Shin Jung, Ung Soo Kim

KEPCO Engineering & Construction Company, Inc., Korea, Republic of

Experimental Study on Spray Activation in a Containment Atmosphere with Superheated Steam Conditions during a Design Basis Accident

Myeong-Seon Chae¹, Domenico Paladino¹, Ralf Kapulla¹, Ghendour Nabil¹, Yago Rivera¹, Simon Suter¹, Christoph Hug¹, Mira Jreissaty², Jean-Luc Vacher²

1: Paul Scherrer Institut (PSI), Switzerland; 2: Électricité de France (EDF), France

Experimental Test Results of the for ISP-52: NEA/ETHARINUS Project Contribution to DEC-A Safety Assessment

Martina Adorni¹, Anis Bousbia Salah², Simon Philipp Shollenberger³, Lars Dennhardt³, Omar Sharief Al-Yahia⁴, Alessandro Del Nevo⁵

1: OECD Nuclear Energy Agency (NEA), France; 2: BEL V, Belgium; 3: Framatome, Germany; 4: PSI, Switzerland; 5: ENEA, Italy

Simulations of a Station Black-Out with Extended Special Emergency Safety Systems and Automatic Partial Cooldown

Davide Papini, Peter Sarkadi, Reinhold Horstmann, Michael Hauck, Rainer Kaulbarsch

NPP Gösgen (KKG), Switzerland

VIKTORIA Experiments in the Frame of R&D Project on Sump Filtration during a Loss of Coolant Accident : Effect of the Type of Filter, the Mass of Fiber and the Presence of Zinc

Georges Repetto¹, Jean-Francois Trigeol¹, Vojtech Soltesz²

1: Autorité de Sûreté Nucléaire et de Radioprotection, France; 2: VUEZ A.S, Slovakia

Safety Analyses of Events in Shutdown Modes of VVER-1000

Pavel Kral¹, Radim Honcarenko², Zbynek Vymetalik²

1: UJV Rez, Czech Republic; 2: CEZ, Czech Republic

Scalability of Validation Data for Safety Evaluation of Water-cooled Nuclear Reactors

Chul-Hwa Song, Kyoung-Ho Kang, Ki-Yong Choi

Korea Atomic Energy Research Institute, Korea, Republic of

Large Scale Best-Estimate Plus Uncertainty Analysis of LOFT L2-5 Experiment

David Aumiller, Gabriel Kooreman, Michael Meholic

NNL, United States of America

The Role of BEPU Methodology in Nuclear Safety Demonstration

Julien Beaucourt, Fabrice Fouet, Pierre Ruyer, Franck Dubois

ASNR (Autorité de Sûreté Nucléaire et de Radioprotection), France

A Global Dialogue on Broadening Industrial Applications of BEPU: Outcomes of a Panel Session at the BEPU-2024 Conference

Jinzhao Zhang¹, Martina Adorni², Alberto Ghione³, Lucia Sargentini³, Vincent Larget⁴, Robert Martin⁵, Dong Gu Kang⁶, Alessandro Petruzzi⁷

1: TRACTEBEL, Belgium; 2: OECD/NEA, France; 3: CEA, France; 4: EDF, France; 5: USNRC, United States of America; 6: KINS, Korea, Republic of; 7: NINE, Italy

Research on Thermal Safety of Intensive Spent Fuel Dry Storage Facility for Heavy Water Reactor

Zhen Xu, Bing Ren, Zhan Liu, Qin Ye

Shanghai Nuclear Engineering Research & Institute CO.LTD, China, People's Republic of

BEPU LBLOCA Analysis Including Zirlo, FeCrAl and Cr-coated Zry Cladding

David Canal¹, Cesar Queral¹, Emilio Castro-Gonzalez¹, Jorge Sanchez-Torrijos²

1: Universidad Politécnica de Madrid, Spain; 2: NFQ Advisory Services, Spain

Steam Line Break with Blowdown of Multiple Steam Generators

Joakim Holmström, Magnus Holmgren

Ringhals AB, Sweden

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

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

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

On the Grid Resolution and Accuracy of under-resolved Direct Numerical Simulations Using the Spectral-element Method

Akshat Mathur¹, Shubham Dalvi¹, Mathis Bode², Ed Komen¹

1: NRG, Netherlands, The; 2: Forschungszentrum Jülich, Germany

Investigation on the Geometrical Representation of an Infinite Array of Rod Bundle Subchannels Using Direct Numerical Simulations (DNS)

Akshat Mathur¹, Jacob Ashwin Abraham², Marc Gerritsma², Ed Komen¹

1: NRG, Netherlands, The; 2: Delft University of Technology, Netherlands, The

A DNS Study on the Effect of Idealised Surface Roughness for Supercritical Flows Inhorizontal and Vertical Channels

Kenneth Chinembiri¹, Shuisheng He¹, Wei Wang²

1: University of Sheffield, United Kingdom; 2: Science and Technology Facilites Council, United Kingdom

High-fidelity CFD Simulation of a Turbulent Mixed Convection Axial Flow in a Heated Rod Bundle Using Code_saturne

Vladimir Duffal¹, Amine Laieb¹, Erwan Le Coupanec¹, Rodrigo Vicente Cruz², Cédric Flageul³, Eric Lamballais³, Stéphane Pujet¹, Sofiane Benhamadouche¹

1: EDF R&D, France; 2: CMCC Foundation - Euro-Mediterranean Center on Climate Change, Italy; 3: Pprime Institute, CNRS – Univ. Poitiers – ISAE/Ensma, France

Simulations of External Flow Around Twisted Elliptical Tubes for Above Unity Prandtl Numbers and Low to Moderate Reynolds Numbers

Sierra A Tutwiler¹, Dillon R Shaver², Lane B Carasik¹

1: Virginia Commonwealth University, United States of America; 2: Argonne National Laboratory, United Stated of America

Development of a Boiling Model at Bubble Scale

Marine Amoros, Ferdinand Louis, Nicolas Goreaud, Alejandro Carlos Mourgues Codern

FRAMATOME SAS, France

Numerical Modeling Saturated Pool Boiling Condition with Ultrasonic Treatment

Luyao Tang, Mingchen Gao, Wei Xu, Xiaojing Liu

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

Vortex and Turbulence Statistics in Nucleate Pool Boiling

Sara Youssoufi¹, Amir Riaz², Elias Balaras¹

1: George Washington University, United States of America; 2: University of Maryland, United States of America

Development of an Advanced Wall Boiling Model to Predict Dryout in CANDU Flow Conditions

Robert Brewster¹, Emilio Baglietto¹, Krishna Podila², Geoff Waddington²

1: Massachusetts Institute of Technology, United States of America; 2: Canadian Nuclear Laboratories, Canada

Interface Capturing Simulation of Subcooled Turbulent Flow Boiling

Nicholas J. Mecham, Nam T. Dinh, Igor A. Bolotnov

North Carolina State University, United States of America

Turbulent Free Convection in a Pool Combining Bubble Dynamics, Surface Evaporation and Water Level Descent

Joauma Marichal¹, Pierre Ruyer², Yann Bartosiewicz¹

1: Université Catholique de Louvain, Belgium; 2: Autorité de Sûreté Nucléaire et de Radioprotection (ASNR), France

Experimental Study of Mass Transfer and Solubility of Boric Acid with Steam

Zehua Zhao¹, Guogang Bao², Lili Tong¹, Xuewu Cao¹

1: Shanghai Jiao Tong University, China, People's Republic of; 2: China Nuclear Power Technology Research Institute Co., Ltd, China, People's Republic of

Flow Analysis in a 3×3 Helical Cruciform Fuel Bundle Using Magnetic Resonance Velocimetry for CFD Validation

Hyeongi Moon, Sejin Oh, Hangfei Dong, Simon Song, Minseop Song

Hanyang University, Korea, Republic of

Flow Structure and Temperature Fluctuation of Penetration Flow in a T-junction Branch Pipe with an Upstream Elbow

Koji Miyoshi

Institute of Nuclear Safety System, Inc., Japan

Multi-hour Steam Generator Transient Temperature Modelling for Stress Analysis Using Conjugate Heat Transfer CFD

Graham Macpherson¹, Chris Sigournay²

1: Frazer-Nash Consultancy, United Kingdom; 2: EDF Nuclear Services, United Kingdom

Study on the Effect of the Dissolved Gas in the Hydroaccumulator

Attila Guba, Aron Hegedus

HUN-REN Centre for Energy Research, Institute for Atomic Energy Research, Hungary

FeCrAl and Cr Coating ATF Performance in Accidental Sequences of VVER-1000 Reactors

Elena Redondo-Valero¹, Jorge Sanchez-Torrijos¹, Cesar Queral², David Canal², Emilio Castro-Gonzalez², Kevin Fernandez-Cosials², Victor Sanchez-Espinoza³

1: NFQ Advisory Services, Spain; 2: Universidad Politécnica de Madrid, Spain; 3: Karlsruhe Institute of Technology (KIT), Germany

Impacts of Internal Heating on Temperature Distribution in Channels

Lubomir Bures, Per Nilsson

Saltfoss Energy ApS, Denmark

Digital Twins and Separate Effects Loop to Support Operation of a Stable Salt Reactor

Thanh Hua¹, Nathaniel Hoyt¹, Yan Cao¹, Alexander Heifetz¹, Taeseung Lee¹, Konstantinos Prantikos¹, Benjamin Cutrer¹, Won-Sik Yang², Seungug Jae², Hangyu Lee², Luke Godfrey³, Tom Taylor³, Rory O’Sullivan³

1: Argonne National Laboratory, United States of America; 2: University of Michigan, United States of America; 3: Moltex Energy, Canada

Hydrogen Risk Analysis of Aqueous Homogeneous Reactor with Gas Recombination System

Ruicheng Zhong, Ziqiang Zou, Hongping Sun Sun, Yu Liu, Jiagen Wang, Yong Zhang, Zhifang Qiu, Wenpei Feng

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

Development of a High-Fidelity Radiative Heat Transfer Model for Assessing Thermal Radiation Influence in Molten Salt Reactors

Mahdi Aghili Nasr¹, Lorenzo Loi¹, Carolina Introini¹, Antonio Cammi^1,2

1: Politecnico di Milano, Italy; 2: Khalifa University, United Arab Emirates

Pressure Drop Experiment in Conical Pebble Bed

Minho Lee, Seth Cadell

Kairos Power, United States of America

Application of Subchannel Analysis to NACIE Pin Bundle

Francesco Lodi¹, Zhihao Ren², Wang Ze², Rong Zhang³, Haoyang Li³, Dalin Zhang³, Ivan Di Piazza¹, Roberto Da Vià¹, Fabio Giannetti⁴, Cristiano Ciurluini⁴, Gianfranco Caruso⁴, Luca Ammirabile⁵, Mirea Mladin⁶, Cristina Dobrinoiu⁶, Akhmed Baisov⁷, Artem Zalesov⁷, Andrey Vlasenko⁸, Anton Ekimov⁹, Andrei Moiseev⁹, Elena Biiushkina⁹, Konstantin Sergeenko⁹, Anton Moisseytsev¹⁰, Rachel Thomas¹⁰, Daniel J. O'Grady¹⁰, Jun Liao¹¹, Nicole Virgili¹², Vladimir Kriventsev¹²

1: ENEA, Italy; 2: CNPRI, Italy; 3: XJTU, China; 4: UniRoma La Sapienza, Italy; 5: EC-JRC; 6: RATEN ICN, Romania; 7: Gidropress, Russian Federation; 8: IBRAE RAN, Russian Federation; 9: NIKIET, Russian Federation; 10: ANL, United States of America; 11: Westinghouse, United States of America; 12: IAEA

Development of a Subchannel Thermal-Hydraulics Analysis Code for the Natrium^® Demonstration Reactor

Dustin Langewisch, Sterling Harper, Stephen Palzewicz, Ricky Dehner, SuJong Yoon

TerraPower, LLC, United States of America

Coupled Simulation for Pin-wise Reactor Core Using Subchannel Analysis Code CUPID with Neutron Kinetic and Fuel Performance Code

Yeonghun Lee¹, Kyuseok Sim¹, Youho Lee¹, Hyoung Kyu Cho¹, Jae Ryong Lee²

1: Seoul National University, Korea, Republic of; 2: Korea Atomic Energy Research Institute, Korea, Republic of

Validations of a New Subchannel Analysis Code for the Next Generation BWR Fuel Bundles - Void Fraction and Two-phase Pressure Drop in Single-tube and Rod Bundle -

Yuki Narushima, Takuji Nagayoshi, Kenichi Katono

Hitachi, Ltd., Japan

Minimum Film Thickness Estimations in BWR Fuel Assemblies by Applying a Method to Partition the Liquid Flow in the Annular Flow Regime with the Subchannel Code CTF

Daniel Vincent Mazzone^1,2, Omar Al-Yahia¹, Ivor Clifford¹, Annalisa Manera^1,2

1: Paul Scherrer Institute (PSI), Switzerland; 2: ETH Zürich, Switzerland

Implementation and Verification of the Plate-Type Fuel Heat Structure in the Sub-Channel Thermal-Hydraulics Code CTF

Wonkyo Cho, Agustin Abarca, Maria Avramova

North Carolina State University, United States of America

Deposition Characteristics of Particles in Printed-Circuit Heat Exchangers based on Dynamic Mesh Method

Qi Sun, Haifeng Liu, Zixuan Zhang, Xiaozhong Wang, Wenkui Ma, Gang Zhao, Xiaoyong Yang, Wei Peng

Tsinghua University, China, People's Republic of

Establishment of a Mechanism-data Hybrid Driving Model of Shell-and-tube Heat Exchanger based on MWORKS/Modelica

Chunyuan Niu, Minjun Peng, Xinheng Zhao

Harbin Engineering University, China, People's Republic of

Numerical Design Framework for the PCHE Heat Exchanger in the Supercritical CO2 Brayton Cycle

Jianxin Qiao

Southeastern University, China, People's Republic of

Surrogate-based Shape Optimization of Airfoil Fin PCHE based on the FFD Method

Tong Liu

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

Impact of Nodalization on Performance Analysis of Helical Steam Generator

Semin Joo, Junha Hwang, Taeyeon Min, Jeong Ik Lee

Korea Advanced Institute of Science and Technology, Korea, Republic of

IAEA's Current Efforts in Advancing Reactor and Nuclear Fuel Technologies

Anzhelika Khaperskaia, Alexei Miassoedov

IAEA, Austria

Numerical Modeling of CRUD Layer for Investigating Thermal Limit in PWR-Type Nuclear Power Plant

Ju Hun Jung, In Cheol Bang

Ulsan National Institute of Science and Technology (UNIST), Korea, Republic of

Study of Passive Safety Injection of Core Makeup Tank and Analysis of the Influencing Factors

Junxiao Yang, Chengcheng Deng, Yi Huang

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

CATHARE Code Assessment of Steam Line Break Scenario with Passive Auxiliary Feedwater Cooling

Anis Bousbia Salah

Bel V, Belgium

Thermal Behavior of a PHWR Channel with an Eccentric Pressure Tube in an Oxidizing Environment

Ketan Ajay¹, David Novog¹, Ravi Kumar², Deb Mukhopadhyay³

1: McMaster University, Canada; 2: Indian Institute of Technology Roorkee, India; 3: Bhabha Atomic Research Centre, India