Conference Agenda

Shaping the Future of Nuclear Energy: Thermal Hydraulics, Key to Robust Defense-in-Depth and Severe Accident Management

Véronique Rouyer

OECD Nuclear Energy Agency, France

Thermal Hydraulics Innovation Enables Global Nuclear Energy Resurgence

W. David Pointer

Oak Ridge National Laboratory, United States of America

Regulation and Research for Demonstrating and Deploying Advanced Nuclear Systems

Toyoshi Fuketa

Nuclear Damage Compensation and Decommissioning Facilitation Corporation, Japan

From Models to Agents: Rethinking Safety Analysis in the Age of AI

Nam Dinh

North Carolina State University, United States of America

Artificial Intelligence and Nuclear Power: Developments and Challenges

Kyun-Tae Kim

Korea Institute of Nuclear Safety, Korea, Republic of

Advancing LWR Core Thermal Hydraulics Through Disruptive Innovation

Jean-Marie Le Corre

Westinghouse Electric Company, Sweden

Identification of Boiling Regime Based on Hydrodynamic Behavior and Heat Transfer Characteristics of Single Droplet-Heated Wall Collision

Ahyeong Cho, Hyungdae Kim

Kyung Hee University, Korea, Republic of

Modeling of Distribution Parameter and Drift Velocity for Microgravity Two-phase Flow

Xu Han^1,2, Aijie Li^1,2, Shixun He^1,2, Xiaobo Zeng^1,2, Di Wu^1,2, Jie Cheng^1,2, Jianjun Wang^1,2

1: College of Nuclear Science and Technology, Harbin Engineering University, China, People's Republic of; 2: Heilongjiang Provincial Key Laboratory of Nuclear Power System and Equipment, Harbin Engineering University, China, People's Republic of

Geometric Effects of Inverted U-bend on Two-phase Transport

Zhengting Quan, Alicja Stoppel, Seungjin Kim

Purdue University, United States of America

A Simulant of R134a-Ethanol Flow for Investigating Steam-water Annular Flow under High-pressure and High-temperature Conditions

Huacheng Zhang¹, Yutaro Umehara¹, Shoji Mori¹, Naoki Horiguchi², Hiroyuki Yoshida²

1: Kyushu University, Japan; 2: Japan Atomic Energy Agency, Japan

One-group Interfacial Area Transport in Vertical Two-phase Flow with Inverted U-bend

Zhengting Quan, Seungjin Kim

Purdue University, United States of America

Mechanistic Model of Heat Transfer Coefficient in AnnularTwo-Phase Flow

Roman W. Morse¹, Gustavo Aguiar¹, Kristofer M. Dressler², Jean-Marie Le Corre³, Evan T. Hurlburt⁴, Matteo Bucci¹

1: Massachusetts Institute of Technology, United States of America; 2: University of Wisconsin-Madison, United States of America; 3: Westinghouse Electric Sweden, Sweden; 4: Naval Nuclear Lab, United States of America

CFD Analysis of the Thermal-hydraulic Behavior during Loss of Coolant Accident (LOCA) Using ATLS-CUBE Test Facility Results

Mubashir Hassan¹, Nourah Alyammahi², Yacine Addad¹

1: Khalifa University of Science and Technology, United Arab Emirate; 2: Federal Authority for Nuclear Regulation (FANR), United Arab Emirate

RELAP5 and TRACE Simulations of Reflood Experiments at the RBHT Facility

Luís De Juan, Jordi Freixa

Universitat Politècnica de Catalunya, Spain

Modelling and Validation of Mixed Convection Flows in the SUPERCAVNA Facility using CFD Tools

Clotaire Geffray, Alain Genty

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

Investigating TRACE Thermal Hydraulics Code through Sensitivity Analysis and Inverse Uncertainty Quantification: A Case Study at KAIST's Condensation Test Facility

Samah A. Albdour, Yacine Addad

Khalifa University, United Arab Emirates

High-detailed CFD-Investigation of the MOTEL Facility for the Analysis of Cross-flow Experiments

Michael Böttcher, Victor Sanchez-Espinosa

Karlsruhe Institute of Technology, Germany

Investigation of Bolgiano-Obukhov Scaling in Mixed Convective Flow with a LES Model of the GaTE Facility

Theodore Ming-Rei Chu, Lane B. Carasik

Virginia Commonwealth University, United States of America

An Overview of Ongoing and Planned R&D Activities in the Field of Experimental Two-phase Thermal-hydraulics at CEA IRESNE

Jimmy Martin, Philippe Aubert, Stéphane Bantiche, Manon Bottin, Jérôme Brinster, Florence Bruna, Julien Carré, Sidonie Clément, Christine Coquelet-Pascal, Jean-Philippe Descamps, Fabrice François, Sébastien François, David Guénadou, Anthony Laurent, Matthieu Manzoni, Bastien Menc, Damien Moulin, Rémy Savoyet, Ludovic Sou, Mathilde Talluau, Eric Verloo

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

Experiment Investigation of Condensation Effect on an Integral Facility

Zijiang Yang, Di Liu, Zhanfei Qi, Kemei Cao

Shanghai Nuclear Engineering and Design Corporation, China, People's Republic of

Validation on Condensation Heat Transfer Models of SPACE and MARS-KS based on Condensation Experiment Facility for Small Modular Reactor Passive Safety System

Hyo Jun An¹, Jae Hyung Park¹, Taeseok Kim², Geunyoung Byeon¹, Yujin Kim¹, JinHo Song¹, Sung Joong Kim^1,3

1: Department of Nuclear Engineering, Hanyang University, Korea, Republic of; 2: Jeju National University, Korea, Republic of; 3: Institute of Nano Science and Technology, Hanyang University, Korea, Republic of

Erosion of a Stratified Containment Atmosphere by a Vertical Jet after Interacting with a Horizontal Disk

Domenico Paladino¹, Ralf Kapulla¹, Guillaume Mignot², Sidharth Paranjape³, Simon Suter¹, Myeong-Seon Chae¹, Yago Rivera¹, Nabil Ghendour¹, Christiph Hug¹

1: Paul Scherrer Institut, Switzerland; 2: Oregon State University, United States of America; 3: OST – Ostschweizer Fachhochschule, Switzerland

Experimental Study on Flow-induced Vibration of Plate-type Fuel Assembly

Shuai Liu, Bing Ren, Jie Ding, Jie Zhang, Jingya Sun, Yu Dang, Fujun Gan, Heting Sun, Chaozhu Zhang

Shanghai Nuclear Engineering Research and Design Institute Co.Ltd., China, People's Republic of

Uncertainty Quantification of Calculated Fuel Temperature for the AGR-5/6/7 Irradiation Experiment

Cam Binh Pham, Grant Hawkes, Nancy Lybeck, Courtney Otani, Paul Demkowicz

Idaho National Labaratory, United States of America

Modeling of the Molten Salt Fast Reactor Transient Behavior based on the Modified Point Reactor Kinetics Model

Thanh Hung Nguyen, Yitung Chen

University of Nevada, United States of America

Development for Integrated System-Level Analysis Capabilities in SAM for Molten Salt Reactors

Rui Hu¹, Travis Mui¹, Robert Salko², Tingzhou Fei¹, Daniel O’Grady¹, Ling Zou¹, Huihua Yang³, Wei Ji³

1: Argonne National Laboratory, United States of America; 2: Oak Ridge National Laboratory, United States of America; 3: Rensselaer Polytechnic Institute, United States of America

Experimental Methodology for Forced Convection Heat Transfer in Molten Salt with Volume Internal Heat Source

Shanwu Wang¹, Hui Deng¹, Pinyan Huang¹, Hongxiang Yu¹, Ying Cao², Chong Zhou¹, Yang Zou¹

1: Shanghai Institute of Applied Physics, Chinese Academy of Sciences, China, People's Republic of; 2: ShanghaiTech University, China, People's Republic of

Influence of an Internal Heat Source on Turbulent Wall cooling Heat Transfer in a Circular Pipe

Dong-Hyuk Park, Bum-Jin Chung

KyungHee University, Korea, Republic of

Effect of Uncertainties in the MSRE Model Part 1: Salt Properties, SPECTRA / SUE Analysis

Marek Stempniewicz, Eric de Geus, Dirk Visser, Christian Fedon, Ferry Roelofs, Fajar Pangukir

NRG, Netherlands, The

Effect of Uncertainties in the MSRE Model Part 2: Delayed Neutron Precursors, SPECTRA / SUE Analysis

Marek Stempniewicz, Eric de Geus, Dirk Visser, Christian Fedon, Ferry Roelofs, Fajar Pangukir

NRG, Netherlands, The

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

Validation of NekRS for Flow-splitting in a Wire-wrapped Fuel Pin Bundle

Dillon Shaver¹, Adam Kraus¹, Aleksandr Obabko¹, Sterling Harper², Monica Pham², SuJong Yoon²

1: Argonne National Laboratory, United States of America; 2: TerraPower, LLC, United States of America

Chemical Kinetic Uncertainty Quantification in Hydrogen Combustion Computational Fluid Dynamics Simulation for ENACCEF2 Experiment

Kosuke Motegi, Toshinori Matsumoto, Hiroyuki Shiotsu

Japan Atomic Energy Agency, Japan

Validation of a CFD Model for Steam Condensation in the Presence of a Noncondensable Gas under Conjugate Heat Transfer Conditions

Abdelouahab Dehbi¹, Omar Al-Yahia¹, Lina Si Salah², Mohamed Aksouh²

1: Paul Scherrer Institut, Switzerland; 2: University of Science and Technology Houari Boumediene, Algeria

Validation of the Generalized Multiphase CFD Modelling Approach GENTOP Using Fluent

Thomas Höhne¹, Maycon Ferreira²

1: Helmholtz-Zentrum Dresden - Rossendorf (HZDR), Germany; 2: University of Almeria, Spain

RANS Validation of Two-Layer Scalar Diffusivity Model for High Schmidt Mass Transfer Problems

Kin Wing Wong, Ignas Mickus, Dmitry Grishchenko, Pavel Kudinov

KTH Royal Institute of Technology, Sweden

Validation of Multi-Phase CFD for Compact Steam Generator Application

Witiwat Jiragoontansiri, Koroush Shirvan

Massachusetts Institute of Technology, United States of America

New IAEA Network on Experimental Testing and Validation for Design and Safety Analysis Computer Codes for SMR and Advanced Reactor Designs: NEXSHARE

Xiaoyu Zhang¹, Eve-Lyne Pelletier¹, Martina Adorni², David Hummel³, Andrew Morreale⁴

1: International Atomic Energy Agency; 2: OECD Nuclear Energy Agency; 3: Generation IV International Forum; 4: Canadian Nuclear Laboratories, Canada

Main Outcomes of the McSAFER Project Devoted to Numerical and Experimental Investigations for the Safety Assessment of Water-cooled SMRs

Victor Hugo Sanchez Espinoza¹, Heikki Suikkanen², Ville Valtavirta³, Marek Bencik⁴, Sören Kliem⁵, Cesar Queral⁶, Barbara Vezzoni⁷, Alan Charles⁸, Paul van-Uffelen⁹, Marcus Seidl¹⁰, Christophe Schneidesch¹¹, Dmitry Grishchenko¹², Hector Lestani¹³

1: Karlsruhe Institute of Technology (KIT), Germany; 2: LUT University, Finland; 3: VTT, Finland; 4: UJV Rez a.s, Czech Republic; 5: Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Germany; 6: Universidad Politécnica de Madrid (UPM), Spain; 7: CEA, France; 8: Global Amentum, United States of America; 9: Joint Research Centre Karlsruhe, Germany; 10: PreussenElektra GmbH, Germany; 11: Tractebel Engineering S.A, Germany; 12: PreussenElektra GmbH, Sweden; 13: Comision Nacional de Energia Atomica (CNEA), Argentina

Analysis of Dynamic Response of Passive ECCS Valves Using MARS-KS Code Based Scheme, SEMICOM

Young Seok Bang, Youngjae Park, You Cho Choi, Seong-Su Jeon, Soon-Joon Hong

Future and Challenge Technology Co., Korea, Republic of

Ensuring Assessment of Safety Innovations for Light Water SMR: Experimental Testing, Code Validation, and Reliability Assessment in the Horizon Euratom EASI-SMR Project

Fulvio Mascari¹, Franck Morin², Petr Bizek³, Nicolas Sobecki⁴

1: ENEA, Italy; 2: CEA, France; 3: UJV, Czech Republic; 4: EDF, France

Convolutional Prototype Learning-based Open Set Recognition Fault Diagnosis Method for Nuclear Power Plant Faults

Jiangkuan Li¹, Shangcai Zheng², Jiaoshen Xu¹, Sichao Tan¹, Ruifeng Tian¹

1: Harbin Engineering University, China, People's Republic of; 2: China Nuclear Power Engineering Co., Ltd., China, People's Republic of

Effects of Heat Pipe Temperature Oscillation on the Operation of Supercritical CO2 Heat Pipe Cooled Reactor

Guanghui Jiao, Genglei Xia, Tao Zhou, Jianjun Wang

Harbin Engineering University, China, People's Republic of

Experimental Analysis of Non-Condensable Helium and Steam Distribution Due to Condensation in the CIGMA Facility Simulating the Reactor Building

Ari Hamdani, Shu Soma, Satoshi Abe, Yasuteru Sibamoto

Japan Atomic Energy Agency, Japan

Comparison between EDF MAAP5.04 and EPRI MAAP6 Codes on Hypothetical Severe Accidents in an Integral PWR

Jeremy Bittan, Nikolai Bakouta

Electricité de France, France

Severe Accident Progression Analyses of Loss-of-coolant Accidents with Different Break Sizes in a Typical Japanese Four-loop PWR Using MELCOR2.2

Junichi Kaneko, Wataru Kikuchi, Masashi Taira

Nuclear Regulation Authority, Japan

ASTEC V3: A Comprehensive Integral Code for Nuclear Safety Analysis and Research – Overview of Recent Applications and Perspectives

Patrick Drai, Nathalie Girault, Laure Carénini, Laurent Laborde, Julie-Anne Zambaux, Laure Cloarec, Eric Delaume, Florian Fichot, Catherine Marchetto, Sidi Souvi, Olivia Coindreau, François Virot, Matthieu Jobelin, Guillaume Astier, Francois Kremer, Karine Chevalier-Jabet, Loic Bosland, Helene Bloch, Lionel Chailan, Raphael Monod, Michael Johnson, Laurent Foucher, Alexandre Bleyer, Alain Commande, Jean-Jacques Ingremeau, Marton Szogradi

Autorité de Sûreté Nucléaire et de Radioprotection, France

Analyzing Steam Explosions During Severe Accidents in Nordic BWRs with MELCOR-TEXAS Coupling

Govatsa Acharya, Ibrahim Batayneh, Dmitry Grishchenko, Pavel Kudinov

KTH Royal Institute of Technology, Sweden

Progress in Utilizing Macroscopic Models for the Simulation of Passive Systems in the Lumped Parameter Code AC²/COCOSYS

Ivan Bakalov, Claus Spengler, Nils Reinke, Siegfried Arndt

GRS, Germany

OECD/NEA Benchmark on Artificial Intelligence and Machine Learning for Critical Heat Flux Predictions – Summary of Phase 1 Results

Jean-Marie Le Corre¹, Gregory K. Delipei², Xu Wu², Xingang Zhao³, Oliver Buss⁴

1: Westinghouse Electric Sweden AB, Sweden; 2: North Carolina State University, United States of America; 3: University of Tennessee, United States of America; 4: Nuclear Energy Agency, France

AI-informed Parameter Selection for Critical Heat Flux Prediction Models: Revisiting the Role of Inlet Conditions

Evgeny Nikitin, Sören Kliem

Helmholtz-Zentrum Dresden-Rossendorf, Germany

Assessment of the State-of-the-Art AI Methods for Critical Heat Flux Prediction

Wen Zhou, Shuichiro Miwa, Hanyu Wang, Koji Okamoto

The University of Tokyo, Japan

Deep Learning for Critical Heat Flux Regression through an Increasing-complexity Approach

Gabriele Cassetta, Lucia Sargentini, Alberto Ghione, Guillaume Damblin

CEA, France

A Data-informed Continuous Machine Learning Approach for CHF Prediction

Meiqi Song^1,2, Fabian Wiltschko², Xiaojing Liu¹, Aurelian Florin Badea², Xu Cheng²

1: Shanghai Jiao Tong University (SJTU), China, People's Republic of; 2: Karlsruhe Institute of Technology (KIT),Germany

Study of the Operation of a Passive Heat Removal System on a Light Water Small Modular Reactor with the ASTEC V3.1.2 Code

Julie-Anne Zambaux¹, Zhexi Guo², Sicong Xiao²

1: Autorité de Sûreté Nucléaire et Radioprotection (ASNR), France; 2: Singapore Nuclear Research and Safety Institute (SNRSI), Singapore

Experimental and Numerical analysis on Safety Condenser Transient Performance based on P1.2 Experiments at the PKL Facility

Ignacio Gomez-Garcia-Torano¹, Maxime Vernassiere¹, Simon Schollenberger², Lars Dennhardt², Andreas Wielenberg³, Sebastian Buchhlotz³, Massimiliano Polidori⁴, František Lahovský⁵, Omar Sharief Al-Yahia⁶, Michaël Montout⁷

1: Commissariat à l'énergie atomique et aux énergies alternatives (CEA), France; 2: Framatome GmbH, Germany; 3: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Germany; 4: Agenzia Nazionale per le Nuove Tecnologie, l'Energia e lo Sviluppo Economico Sostenibile (ENEA), Italy; 5: Nuclear Research Institute Řež (UJV), Switzerland; 6: Paul Scherrer Institut (PSI), Switzerland; 7: Electricité de France (EdF), France

Reliability Assessment of the BWRX-300 Passive Isolation Condenser System: Addressing Uncertainties in Two-Phase Natural Circulation Flow Modeling

Graeme Trundle¹, Sergey Galushin², Sean Roshan¹, Sevostian Bechta¹, Michael Söderström³, Anders Olsson²

1: Royal Institute of Technology (KTH), Sweden; 2: Vysus Group, Sweden; 3: Vattenfall AB, Sweden

Numerical Study on Steady-State Characteristics of Two-Phase Loop Thermosiphon in a Novel Small Modular Reactor

Xubin Wu, Wentao Hao, Wenwen Zhang, Zhenlei Liu, Weihua Li, Xingtuan Yang

Institute of Nuclear and New Energy Technology, Tsinghua University, China, People's Republic of

Update on Standard Wall Modeled Large Eddy Simulation on a Few Validation Test-cases

Jean-François Wald¹, Leo Deperne¹, Luc Bertolotti¹, Samuel Gaucher¹, Clément Colas¹, Sofiane Benhamadouche^1,2

1: EDF R&D, France; 2: CEREA, France

Wall-resolved LES for Predicting Turbulent Flow through Tube Bundles

Sofiane Benhamadouche

EDF R&D, France

Simulation of NACIE Tests with Thermal Hydraulics Systems Codes

Konstantin Mikityuk¹, Anton Moisseytsev², Ossama Abedelhalim³, Luca Ammirabile⁴, Ilya Aulov⁵, Akhmed Baisov⁵, Barbara Calgaro⁶, Gianfranco Caruso⁷, Zhao Chen⁸, Pietro Cioli Puviani⁹, Cristiano Ciurluini⁷, Tommaso Del Moro⁹, Ivan Di Piazza⁹, Daniel Dupleac¹⁰, Anton Ekimov¹¹, Paolo Ferroni¹², Nicola Forgione¹³, Fabio Giannetti⁷, Wenjun Hu¹⁴, Olga Ilyasova¹⁵, Vladimir Kriventsev¹⁶, Sung Jin Lee¹⁷, Jun Liao¹², Pierdomenico Lorusso⁹, Daniele Martelli⁹, Mirea Mladin¹⁸, Evgenii Moiseenko¹⁵, Andrei Moiseev¹¹, Nikoleta Morelova¹⁶, Nastasya Mosunova¹⁵, Roxana-Mihaela Nistor-Vlad¹⁰, Marilena-Luminita Nitulescu¹⁸, Dan Ogrady², Fajar Pangukir¹⁹, Oriol Parera Villacampa⁴, Byung Ha Park²⁰, Andrea Pucciarelli¹³, Lixia Ren¹⁴, Zhihao Ren⁸, Ferry Roelofs¹⁹, Konstantin Sergeenko¹¹, Pietro Stefanini³, Mariano Tarantino⁹, Rachel Thomas², Vyacheslav Vereshchagin¹¹, Vasily Volkov⁵, Dalin Zhang²¹, Lei Zhou¹⁴

1: PSI, Switzerland; 2: ANL, United States of America; 3: UNIPI, Italy; 4: JRC, EC; 5: Gidropress, Russian Federation; 6: newcleo, Italy; 7: La Sapienza, Italy; 8: CNPRI, China, People's Republic of; 9: ENEA, Italy; 10: PUB, Romania; 11: NIKIET, Russian Federation; 12: Westinghouse, United States of America; 13: UNIPI, Italy; 14: CIAE, China, People's Republic of; 15: IBRAE RAN, Russian Federation; 16: IAEA, Austria; 17: Fauske & Associates, United States of America; 18: RATEN ICN, Romania; 19: NRG, Netherlands, The; 20: KAERI, Korea, Republic of; 21: XJTU, China, People's Republic of

Evaluation of CTF and DRACCAR Capabilities for Reflood Modelling with RBHT-I Open Tests

Lucia Rueda Villegas, Adrian Sabater Alcaraz, Jinzhao Zhang

TRACTEBEL, Belgium

Validation of Domain Overlapping Coupling Between SAM and MOOSE-Subchannel Using NACIE Test

Gang Yang, Nicolas Bodet, Anton Moisseytsev, Ling Zou, Rui Hu

Argonne National Laboratory, United States of America

Comparisons of ASTEC and DRACCAR Codes Results Against COAL B0 Experiments under Bundle Reflooding Conditions

Pietro Maccari, Marica Eboli, Alessandro Del Nevo

ENEA – Nuclear Department, Experimental Engineering Division (NUC-ING), Italy

HYBISCUS-II: Numerical Simulation of a Pressurized Thermal Shock on a Reduced Scale Experiment

Alexis Thibault Marie Berny¹, Cecile Demathieu², Joël Nibas², Alain Ungar²

1: Université Paris Saclay, CEA, France; 2: EDF R&D Lab Chatou, France

Numerical Activities of PASI Experiments for Passive Containment Cooling in the European PASTELS Project

Franck David¹, Paul Dene¹, Gian Marco Grippo³, Julie Anne Zambaux⁵, Sebastian Buchholz⁴, Joonas Telkka⁶, Omar Al Yahia⁷, Adam Kecek⁸, Zbynek Parduba⁸, Augustin Brissart², Massimiliano Polidori³, Ignacio Gomez²

1: EDF R&D, France; 2: CEA, France; 3: ENEA , Italy; 4: GRS, Germany; 5: IRSN, France; 6: LUT, Finland; 7: PSI, Switzerland; 8: UJV, Czechia

Experimental Study of Two-phase Flow in a Four-by-four Unheated Rod Bundle for Validation of Thermal-hydraulics Simulation Codes

Muhammad Ilham, Taku Nagatake, Shinichiro Uesawa, Ayako Ono

Japan Atomic Energy Agency, Japan

Assessment of Spacer Grid Effects and Flow Development in a Triangular Rod Bundle: A PIV-DNS Cross-Comparison

Danila Trane¹, Mattia Grespan¹, Enrico Stalio², Diego Angeli¹

1: Department of Sciences and Methods for Engineering, University of Modena and Reggio Emilia, Italy; 2: Department of Engineering Enzo Ferrari, University of Modena and Reggio Emilia, Italy

Evaluation on Two Different Mixing Vanes by PIV Experiment in 5x5 Rod Bundle

Junya Inoue¹, Yoshitaka Mizutani¹, Yasushi Hirai¹, Tatsuya Sasakawa¹, Shigeo Hosokawa²

1: Nuclear Fuel Industries, Ltd., Japan; 2: Kansai University, Japan

Three-dimensional Turbulent Flow Measurements in a 6 × 6 Fuel Rod Bundle

Anis Rezaiguia², Roberto Capanna¹, Charles Fort¹, Emmanuel Lo Pinto², Guillaume Ricciardi², Philippe Bardet¹

1: George Washington University, United States of America; 2: CEA, DES, IRESNE, Nuclear Technology Departement, France

Three-dimensional Void fraction Distribution of Transient Boiling Two-phase Flow in a Heated 5×5 Rod Bundle During Depressurization Process

Shota Ueda¹, Takahiro Arai¹, Masahiro Furuya¹, Riichiro Okawa¹, Atsushi Ui¹, Wataru Sakuma²

1: Central Research Institute of Electric Power Industry, Japan; 2: Mitsubishi Heavy Industries, Ltd., Japan

Investigation of the Post-CHF Heat Transfer Modeling based on the Large Scale RBHT Reflood Data Compilation and Assessment

Yue Jin¹, Andrew Ireland², Stephen Bajorek², Kirk Tien², Chris Hoxie², Fab-Bill Cheung³

1: University of Missouri, United States of America; 2: U.S. Nuclear Regulatory Commission, United States of America; 3: The Pennsylvania State University, United States of America

How Good is “Good Agreement”? Considerations when Comparing Experiments with Simulations for Rod Bundles with Wire Spacers

Julio Pacio¹, Ferry Roelofs², Heleen Uitslag-Doolaard², Akshat Mathur², Dillon Shaver³, Elia Merzari⁴

1: Belgian Nuclear Research Centre (SCK CEN), Belgium; 2: Nuclear Research and Consultancy Group (NRG), The Netherlands; 3: Argonne National Laboratory (ANL), United States of America; 4: Pennsylvania State University (PSU), United States of America

Analyzing the Effect of Deformed Pins in LMFR Rod Bundles

Ferry Roelofs¹, Shubham Dalvi¹, Heleen Uitslag-Doolaard¹, Markus Daubner², Karsten Litfin², Manuela Profir³, Vincent Moreau³, Pietro Cioli Puviani⁴, Daniele Martelli⁴, Ivan Di Piazza⁴, Sara Gonzalez Ruiz⁵, Luis Carbonelle⁵, Matilde Fiore⁵, Mushin Can Akkurt⁵, Philippe Planquart⁵, Julio Pacio⁶

1: NRG, Netherlands, The; 2: KIT, Germany; 3: CRS4, Italy; 4: ENEA, Italy; 5: VKI, Belgium; 6: SCK CEN, Belgium

Numerical Simulation of Flow and Heat Transfer Characteristics for LBE in Plate-type Bundle Fuel Assembly with LES-Smagorinsky Lilly Model

Feiran Wang^1,2, Zhengjie Yang^1,2, Jiming Wen^1,2, Bangyang Xia^3,4, Chong Chen^3,4, Xue Zhang^3,4, Gongjian Xie¹, Ruifeng Tian^1,2, Sichao Tan^1,2

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: National Key Laboratory of Nuclear Reaction Technology, China, People's Republic of; 4: State Key Laboratory of Advanced Nuclear Energy Technology, China, People's Republic of

Thermal-hydraulic and Safety Analysis of a Horizontal Assembly in the LFR during Flow Blockage Accident

Di Wu^1,2, Zhibo Gao^1,2, Minyang Gui^1,2, Jiaxin Zhang³, Jie Cheng^1,2, Xu Han^1,2, Xiaobo Zeng^1,2, Jianjun Wang^1,2

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

Local Hot Spots of the Wire-wrapped Pin Bundle in the Liquid Metal Cooled Reactors

Taehwan Ahn^1,2, Wilhelm-Martin Bissels², Victor Petrov^1,2, Annalisa Manera^1,2

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

Uncertainty Analysis of Rod Bundle Channel for Lead-Bismuth Reactors Based on Sub-channel Code

Deyan Kong, Di Wu, Jie Cheng, Xu Han, Xiaobo Zeng, Jianjun Wang

Harbin Engineering University, China, People's Republic of

Uncertainty propagation and Global Sensitivity Analysis based on the Hilbert-Schmidt Independence Criterion measures. Application to a load rejection transient in a Pressurized Water Reactor

Paolo Olita, Amandine Marrel

French Atomic Energy and Alternative Energies Commission, France

Verification of Multi-scale Post-process Method on PSBT Subchannel Experiment and Application to Rowe and Angle Experiment

Filippo Arlotti¹, Xavier Manchon¹, Jorge Perez-Manes¹, Nathalie Seiler¹, Nicolas Rimbert², Jean-Marc Labit³, Stéphane Mimouni⁴

1: CEA, DES, IRESNE, Cadarache, F-13108 Saint-Paul-lès-Durance, France; 2: Université de Lorraine, CNRS LEMTA, F-54000 Nancy, France; 3: CEA Saclay, F-91191 Gif-sur-Yvette, France; 4: Electricité de France, R&D Division, F-78401, Chatou, France

Leveraging Multiple Fidelities for Thermal-hydraulics Uncertaintyanalyses of Fuel Assemblies Subjected to Deformation

Matilde Fiore, Silvania Lopes, Lilla Koloszar, Philippe Planquart

von Karman Institute for Fluid Dynamics, Belgium

Development of a Robust Stochastic Framework for Numerical Error Estimation and Uncertainty Quantification in Unsteady Flow Simulations

Yu-Jou Wang¹, Emilio Baglietto¹, Patrick McGah², Monica Pham²

1: Massachusetts Institute of Technology; 2: TerraPower, LLC

Toward Simplified, Scalable Detonation Modeling: Independent Validation of the Generalized Turbulent Flame Closure Approach for DDT in Hydrogen Mixtures

Mantas Povilaitis, Stephen Oluwabukunmi Adesina, Julius Venckus

Lithuanian Energy Institute, Lithuania

Validation of the Open Source TUAS Using Coupled Natural Circulation Data from CIET

Theodore Kay Chen Ong¹, Sicong Xiao¹, Per F. Peterson²

1: National University of Singapore (NUS), Singapore; 2: University of California, Berkeley, United States of America

Characteristic Features of Fission Product Behavior by CINEMA and MELCOR Codes during Severe Accidents of OPR1000

Se Hee Kwon¹, Juhyeong Lee¹, Joon Young Bae¹, Yujin Kim¹, Joon-Eon Yang¹, JinHo Song¹, Sung Joong Kim^1,2

1: Department of Nuclear Engineering, Hanyang University, Korea, Republic of; 2: Institute of Nano Science and Techonology, Hanyang University, Korea, Republic of

Experimental Investigation of Pressure Drop in Single and Two-Phase Flow Through Sand Packed Beds

Aimad Bouloudenine, Liangxing Li, Shang Shi, Zutao Xiang

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

Using Calibrated Sodium Data for Preliminary Validation of the SRT Code for Advanced Reactors

Dong Hoon Kam¹, Dave Grabaskas¹, Kyle Becker², Mark Anderson², Tyler Starkus¹

1: Argonne National Laboratory, United States of America; 2: University of Wisconsin-Madison, United States of America

An Update of the Models Related to the In-Vessel Retention Strategy in the MAAP6 Code

Nikolai Bakouta, Jeremy Bittan

Electricité de France, France

Progress in the Development of the ContainmentFOAM CFD Package for Analysis of Current and Future LWR Containment Phenomena

Stephan Kelm¹, Allen George², Ruiyun Ji³, Carlos Vazquez-Rodriguez¹, Khaled Yassin¹, G. Vijaya Kumar⁴, Xiongguo Liu¹, Daniel Schumacher¹, Karl Sturm⁵, Leon Thelen¹, Martin Lövenich¹, Claudia Druska¹, Astrid Kuhr¹

1: Forschungszentrum Juelich GmbH, Germany; 2: Forschungszentrum Juelich GmbH and Karlsruhe Institute of Technology, Germany; 3: Forschungszentrum Juelich GmbH and Universität der Bundeswehr Muenchen, Germany; 4: Forschungszentrum Juelich GmbH, Germany and Indian Institute of Technology Madras, India; 5: Forschungszentrum Juelich GmbH and RWTH Aachen University, Germany

Uncertainty and Sensitivity Analysis of MELCOR-Based Source Term Predictions for the PHEBUS FPT-1 Experiment

Han Sol Park, Yeon-Gun Lee

Sejong University, Korea, Republic of

Use of PINNs to Improve CHF Model Behaviour

Yi Meng Chan

KTH, Sweden

Probabilistic/Interpretable Neural Network Frameworks for Flow Boiling CHF Prediction in Circular Tubes

Kyung Mo Kim¹, Juhyung Lee², Mooneon Lee², Hyuk Kwon², Han Seo²

1: Korea Institute of Energy Technology (KENTCH), Korea, Republic of; 2: Korea Atomic Energy Research Institute (KAERI), Korea, Republic of

Critical Heat Flux Prediction in Round Tubes Using AI/ML: A Comparison of XGBoost and MLP Models

Mooneon Lee¹, Juhyung Lee¹, Hyuk Kwon¹, Dae-Hyun Hwang¹, Kyung Mo Kim²

1: Korea Atomic Energy Research Institute, Korea, Republic of; 2: Korea Institute of Energy Technology, Korea, Republic of

Evaluating the Performance of Diffusion Models for Scientific Data Augmentation - a Case Study with Critical Heat Flux

Farah Raed Hussein Alsafadi, Xu Wu

North Carolina State University, United States of America

Investigation of Pool and Containment Thermalhydraulic Behavior Using the 3D Module of CATHARE-3

Ignacio Gomez-Garcia-Torano¹, Christophe Bourcier¹, Julie-Anne Zambaux², Jeremy Bittan³

1: Commissariat à l'énergie atomique et aux énergies alternatives (CEA), France; 2: Institut de Radioprotection et de Sûreté Nucléaire (IRSN), France; 3: Electricité de France (EdF), France

Study of Flow Distribution of High Power Density Plate-Type PWR by Two-Phase Neutronics-Thermohydraulics Coupling Code

Guodong Liu¹, Ying He¹, Junren Hou¹, Meiming Qiu¹, Xingyu Zhao¹, Shanfang Huang¹, Kan Wang¹, Zhigang Li^2,3, Chuan Lu^2,3

1: Tsinghua University,China, People's Republic of; 2: Nuclear Power Institute of China, China, People's Republic of; 3: National Key Laboratory of Nuclear Reactor Technology, China, People's Republic of

Prediction of Multiphase Flow and Flow-induced Forces in a Cross-flow Tube Bundle with a Morphology-adaptive CFD Model

Marco Colombo¹, Daniele Vivaldi²

1: University of Sheffield, United Kingdom; 2: Autorité de Sûreté Nucléaire et de Radioprotection (ASNR), France

Modeling a Flexible Operation Scenario in a KWU-PWR Reactor using RELAP5/PARCS

Bárbara Navarro¹, Teresa Barrachina¹, Rafael Miró¹, Gumersindo Verdú¹, Marina García-Fenoll², José María Posada²

1: Universitat Politècnica de València, Spain; 2: Centrales Nucleares Almaraz-Trillo (CNAT), Spain

Multiscale Modelling of Forced-to-natural Circulation Experiments in Heavy Liquid Metal Test Loop NACIE

Andrea Pucciarelli¹, Pietro Stefanini¹, Ossama Halim¹, Nicola Forgione¹, Rajendrakumar Murugesan², Natesan Kumaresan², Amit Kumar Chauhan², Dhrumil Ganatra², Mohammed Muaaz Mohamed Dilshad Hussain³, Fajar Pangukir³, Akshat Mathur³, Pietro Cioli Puviani⁴, Tommaso Del Moro⁵, Dalin Zhang⁶, Yixin Zhang⁶, Vladimir Kriventsev⁷, Ivan di Piazza⁸

1: University of Pisa, Italy; 2: IGCAR, India; 3: NRG, Netherlands, The; 4: Politecnico di Torino / ENEA, Italy; 5: Sapienza University of Rome / ENEA, Italy; 6: Xi'an Jiaotong University, China, People's Republic of; 7: IAEA; 8: ENEA Brasimone Research Centre, Italy

Visualization Analysis of Performance of Boiling Heat Transfer on Micro-cavity Surface According to Cavity Diameter

Do Yeon Kim¹, Seon Ho Choi³, Su Cheong Park², Dong In Yu¹

1: Pukyong National University, Korea, Republic of; 2: Dong-a University, Korea, Republic of; 3: Pohang Accelerator Laboratory, Korea, Republic of

Experimental Research of Wicking Degradation Effect on Boiling Heat Transfer Characteristics with Nanostructured Surface

Jian Zhou, Longchang Tang, Wei Xu, Xiaojing Liu

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

Visual Study of Subcooled Boiling based on Femtosecond Laser Modified Surface

Mingze Cao^1,2, XIao Yan¹, Dianchuan Xing¹, jianjun Xu¹, Yanlin Wang¹, Tianzhou Xie²

1: Nuclear Power Institute of China, China, People's Republic of; 2: State Key Laboratory of Advanced Nuclear Energy Technology, China, People's Republic of

Heat Transfer Enhancement for Nucleate Boiling via Microlayer Evaporation on Micro-pillar Arrayed Surface

Wei Ding¹, Jinming Zhang¹, Uwe Hampel^1,2

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

Hydrodynamic and Thermal Investigation of Oxide Layer Microscale Surface Features on Nucleate Boiling

Matthew Val Meyers¹, Patrick Lewis², Ryan Tunstall², Edward Smith³, Giovanni Giustini¹, Alex Skillen¹

1: University of Manchester, United Kingdom; 2: Rolls Royce, United Kingdom; 3: Brunel University, United Kingdom

Assessment of Correlations for Point of Net Vapor Generation using Direct Visual Observation

Taiyang Zhang¹, Joseph L. Bottini²

1: The Ohio State University, United States of America; 2: United States Military Academy, United States of America

Measurement of Interfacial Phenomena Near CHF in High Pressure Water Flows Using High-Speed X-Ray Radiography

Yang Liu, David Novog

McMaster University, Canada

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

Parametric TRACE Code Survey of Fusion DEMO Reactor on Three Representative LOCA Scenarios

Shun Nukaga¹, Masahiro Furuya¹, Yoji Someya², Mitsuya Kato², Yoshiteru Sakamoto²

1: Waseda University, Japan; 2: National Institutes for Quantum Science and Technology, Japan

A Novel Parametric Dynamic Mode Decomposition Formulation: Application to Magnetohydrodynamic Liquid Metal Flows

Matteo Lo Verso¹, Stefano Riva¹, Carolina Introini¹, Eric Cervi¹, Francesca Giacobbo¹, Matteo Di Prinzio², Marco Caramello², Laura Savoldi³, Antonio Cammi^4,1

1: Politecnico di Milano, Italy; 2: Ansaldo Nucleare SpA, Italy; 3: Politecnico di Torino, Italy; 4: Khalifa University, United Arab Emirates

Direct Numerical Simulation of Magneto-Convection at Low Magnetic Reynolds Number

Jake Ineson¹, Alex Skillen¹, Aleksander Dubas²

1: University of Manchester, United Kingdom; 2: United Kingdom Atomic Energy Authority, United Kingdom

Coupled CFD and Neutronics for Accidents and Transients in Fusion Blankets

Arpan Sircar, Jin Whan Bae, Katarzyna Borowiec, Vittorio Badalassi, Cami Collins

Oak Ridge National Laboratory, United States of America

Thermal-hydraulic Analysis in Support of the Design of Water Loop Experimental Facility for Testing Mock-ups in Fusion-like Environment

Alessandra Vannoni¹, Giorgio Mongiardini², Cristiano Ciurluini², Bruno Gonfiotti¹, Pietro Arena¹, Marica Eboli¹, Ranieri Marinari¹, Fabio Giannetti², Alessandro Del Nevo¹

1: ENEA, Italy; 2: Sapienza University of Rome, Italy

On the Development of Tokamak-based Conventional Power Plants

Sara Perez-Martin

Karlsruhe Institute of Technology, Germany

Numerical Analysis of a WCLL BB TBM Mock-up to be Installed in Water Loop Facility

Giorgio Mongiardini¹, Alessandra Vannoni², Cristiano Ciurluini¹, Bruno Gonfiotti², Fabio Giannetti¹, Marica Eboli², Pietro Arena², Pietro Maccari², Alessandro Del Nevo²

1: Sapienza University of Rome, Italy; 2: ENEA – Nuclear Department, Italy

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

Spacer Grid Rewet during Reflood andModeling Challenges for Thermal-Hydraulic Codes

Stephen M. Bajorek¹, Yue Jin², Fan-Bill Cheung³

1: United States Nuclear Regulatory Commission, United States of America; 2: University of Missouri, United States of America; 3: Pennsylvania State University, United States of America

From Data to Trust: High-Resolution Thermohydraulic Experiments for Future Reactors

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

1: Paul Scherrer Institute, Switzerland; 2: ETH Zurich, Department of Mechanical and Process Engineering, Switzerland; 3: University of Michigan– Ann Arbor, United States of America

Particle-Based Approaches to Multiphysics Simulation in Nuclear Safety

Eung Soo Kim¹, Yelyn Ahn¹, Jin Woo Kim¹, Jin Hyun Kim¹, Juryong Park¹, Young Beom Jo², Hee Sang Yoo³

1: Seoul National University, Korea, Republic of; 2: Kyung Hee University, Korea, Republic of; 3: CEA/DES/IRESNE, France

Development of Wall Convective Heat Transfer Model in the Dispersed Flow Film Boiling Regime based on Quasi-steady-state Experiments

Kyung Mo Kim¹, Joseph Kelly², Stephen Bajorek², Xiaodong Sun³

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

Evaluation of RELAP’s Post Critical Heat Flux Behaviour for Uncertain Parameter Quantification and Analysis

Pablo Osés, Jordi Freixa, Germán López

Technical University of Catalonia (UPC), Spain

Effect of the Initial Cladding Temperature and Subcooling on the Quench Temperature during Single Rod Reflood Experiment

WooHyun Jung¹, Nicolas Fox¹, Cole Dunbar¹, Kumar Sridharan¹, Michael Corradini¹, Hwasung Yeom²

1: University of Wisconsin, United States of America; 2: Pohang University of Science and Technology, Korea, Republic of

Measurement and Modeling of the HTC Profile during Quenching of Hot Wall with a Falling Liquid Film

Tomio Okawa, Hiroyuki Umebayashi

The University of Electro-Communications, Japan

Optical Kerr Effect Gated Ultrafast Imaging of Bubbly Flows

Daniel Andrew Hunter, Philippe Matthieu Bardet

The George Washington University, United States of America

A Characterization of the Two-phase Frictional Pressure Drop within a Cylindrical Mini-channel in the Laminar and Turbulent Regimes

Jimmy Martin, Romuald Mallet, Yann Kervegant, Sylvain Chareyre

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

High-Resolution Miniaturized Impedance Sensor for Two-Phase Flow Measurement

Sergio Chiva, Oscar Prades, Guillem Monros, Salvador Torro, Raul Martinez, Jaume Luis

Universitat Jaume I, Spain

Measurement of Aerosol Temperature Field based on Background Oriented Schlieren

Bifan Liu, Danling Zhuang, Qian Wang, Junlian Yin

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

Validation of CATHARE Code Against PERICLES High Pressure Reflooding Experiments

Nicolas Piette, Raphaël Préa, Lopez Thibaut

CEA, France

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

The Implementation of the P1 Approximation in SpectralElement Code for Conjugate Heat Transfer

Haomin Yuan¹, Dezhi Dai¹, Dillon Shaver¹, Yu-hsiang Lan², Michael Seneca³, Tri Nguyen³, Elia Merzari³

1: Argonne National Laboratory, United States of America; 2: University of Illinois at Urbana-Champaign, United States of America; 3: Pennsylvania State University, United States of America

A Wall-Function Type Interface Turbulence Damping Method for Multiple Flow Regimes with Large Scale Interfaces

Vinesh H Gada, Mohit P Tandon

Siemens Industry Software, India

Implementing Improved Relative Velocity Models for Horizontal Bubbly Flow in Computational Multiphase Fluid Dynamics Simulations

Adam John Dix, Seungjin Kim

Purdue University, United States of America

Two-way Coupling Simulations of a Cantilever Rod under a Turbulent Axial Flow between a Beam Equation Solver and Wall-Resolved Large Eddy Simulation at a Moderate Reynolds Number

Thomas Norddine, Sofiane Benhamadouche

EDF R&D, France

Lattice Boltzmann Simulations of Natural Convection in Compressible Fluids

Junren Hou¹, Minyun Liu², Yugao Ma³, Yanping Huang², ShanFang Huang¹, Kan Wang¹

1: Tsinghua University, China, People's Republic of; 2: CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, China, People's Republic of; 3: Nuclear Power Institute of China, China, People's Republic of

Application of Apros and TRACE Codes for Turbine Trip and Inadvertent Operation of ECCS Transient Simulation of Small Modular Reactor

Matti Paananen¹, Tomi Pakarinen², Eerikki Raiko¹, Janne-Pekka Mäkinen²

1: Fortum Power and Heat Oy, Finland; 2: Platom Oy, Finland

Experimental Investigation of Flow Regimes and Friction Factor in a 9×9 Helical Cruciform Fuel Rod Bundle

Hansol Kim, Matthew Kinsky, Dalton Pyle, Joseph Seo, Yassin Hassan

Texas A&M University, United States of America

Analysis of Available Times during LOCA Sequences in NuScale Reactor Design Using the TRACE Code

Yago Martinez-Gonzalez¹, Cesar Queral¹, Kevin Fernandez-Cosials¹, Jorge Sanchez-Torrijos²

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

Investigation of Multi-Dimensional Phenomena in Novel Passive Safety Systems for i-SMR Using MARS-KS

Jan Hruskovic¹, Sang Gyun Nam¹, Youngjae Park¹, Bub Dong Chung¹, Seong-Su Jeon¹, Seung-Jong Oh²

1: FNC Technology Co., Ltd., Korea, Republic of; 2: KEPCO International Nuclear Graduate School, Korea, Republic of

A Review of AP1000/AP600 Experimental Program and Its Applicability to AP300 SMR

Jun Liao, Megan Durse, Oriol Oliva, Richard Wright

Westinghouse Electric Company, United States of America

Uncertainty Quantification of a Postulated Severe Accident Scenario in a Generic German PWR

Livia Tiborcz, Thorsten Hollands

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

ENSO Contribution to the HORIZON 2020 MUSA Project: In-Vessel Uncertainty Analysis with RELAP/SCDAPSIM/MOD3.4 and IUA2.0 of a Long-term SBO Scenario in a Gen-II PWR

Victor Martinez-Quiroga, Marina Perez-Ferragut

Energy Software Ltd., Spain

ENSO Contribution to the IAEA CRP I31033: Uncertainty Analysis with RELAP/SCDAPSIM/MOD3.5 of a Long-term SBO Scenario in a CAREM-like iPWR

Victor Martinez-Quiroga, Marina Perez-Ferragut

Energy Software Ltd., Spain

Stepwise Uncertainty Analysis Methodology in Severe Accidents

Rafael Bocanegra, Victor Martínez Quiroga, Marina Perez Ferragut

ENSO, Spain

Probability Analysis of Steam Generator Heat Transfer Tube Rupture under Severe Accident

ZhanXiang Huang

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

Acceleration of the Convergence of a Core Thermo-hydraulic Code Using Initialization from a Neural Network

Gauthier Lazare^1,2, Corentin Coudroy¹, Philippe Helluy², Erwan Le Coupanec¹, Qingqing Feng¹, Frank Hulsemann¹, Stéphane Pujet¹, Pauline Rotach¹

1: EDF, France; 2: IRMA, France

Model Calibration and Acceleration Techniques Based on Artificial Bee Colony Algorithm

Shubiao Dong, Xiaoyu Zhang, Qianping Zhang

China Nuclear Power Operation Technology Corporation, Ltd, China, People's Republic of

Thermal-Hydraulic Model Development for Semi-Circular Zigzag Channel PCHEs Based on Machine Learning

Shuai Che, Adam Burak, Xiaodong Sun

University of Michigan, United States of America

ANN-based Approach for Two-phase Model Development and Implementation in Thermal-hydraulic Codes

Van Thai Nguyen¹, Ngoc Dat Nguyen²

1: Hanoi University of Science and Technology, Vietnam; 2: Chungnam National University, Korea, Republic of

Thermal Stratification Prediction in Reactor System Based on CFD Simulations Accelerated by A Data-driven Coarse-grid Turbulence Model

Zijing Liu, Pengcheng Zhao, Yanan Zhao, Wei Deng, Tao Yu, Jinsen Xie

University of South China, China, People's Republic of

Towards a new NEA Framework for Advanced Thermalhydraulic Experiments for Nuclear Analysis and Safety Application (ATHENA)

Didier Jacquemain¹, Martina Adorni¹, Joelle Fleurot²

1: OECD NEA; 2: ASNR

Addressing Future Challenges on Analysis and Management of Accidents by International Cooperation: The Working Group on the Analysis and Management of Accidents (WGAMA)

Hideo Nakamura¹, Ahmed Bentaib², Martina Adorni³, Fulvio Mascari⁴, Pierre Ruyer⁵, Jinzhao Zhang⁶

1: Japan Atomic Energy Agency (JAEA), Japan; 2: Institut de Radioprotection et de Sureté Nucléaire (IRSN), France; 3: OECD Nuclear Energy (NEA); 4: National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy; 5: Autorité de sûreté nucléaire et de radioprotection (ASNR), France; 6: Tractebel (ENGIE), Belgium

NEA ETHARINUS Project: A Flagship Project Relevant to Thermal-Hydraulic Safety Analysis Issues

David Pialla¹, Joakim Holmstrom², Anis Bousbia Salah³, Martina Adorni⁴, Amina Younsi⁵, Sofia Carnevali⁶, Vesa Riikonen⁷, Simon Schollenberger⁸

1: EDF, France; 2: VATTENFALL-Ringhals AB, Sweden; 3: BEL V - Nuclear Safety and Analysis, Belgium; 4: OECD Nuclear Energy Agency, France; 5: ASNR, France; 6: Université Paris-Saclay, CEA, France; 7: LUT University, Finland; 8: Framatome, Germany

Key Outcomes of Recent NEA Nuclear Safety Joint Projects in Thermal-Hydraulics

Didier Jacquemain¹, Martina Adorni¹, Sangmin Lee¹, Markus Beilmann¹, Simon Philipp Shollenberger², Vesa Riikonen³, Jordi Freixa⁴, Kyoung-Ho Kang⁵, Miguel Sanchez Perea⁶, Sanjeev Gupta⁷, Martin Freitag⁷, Domenico Paladino⁸

1: OECD Nuclear Energy (NEA), France; 2: Framatome, Germany; 3: LUT University, Finland; 4: Universitat Politecnica de Catalunya, Spain; 5: Korea Atomic Energy Research Institute, Korea, Republic of; 6: Consejo de Seguridad Nuclear (CSN), Spain; 7: Becker Technologies GmbH, Germany; 8: PSI, Switwerland

Bal-Raj Sehgal Memorial Award Lecture: Interface Capturing Simulations for Nuclear Thermal Hydraulics

Igor Bolotnov

North Carolina State University, United States of America

Preventing Localized Hot Spots in Flow Boiling: CHF Enhancement Method for Non-Uniform Heat Conductance Surface

Nairamdakh Bold¹, Shuichiro Miwa¹, Shixian Wang¹, Koji Okamoto²

1: Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, Japan; 2: Nuclear Professional School, School of Engineering, The University of Tokyo, Japan

Effects of Oxidation on CHF on Bare and Grooved Copper Surface in Vertical-face Pool Boiling based on Bubble Behavior

Yanxu Huang¹, Zihao Wang¹, Kai Wang^1,2, Songbai Cheng³

1: Sun Yat-sen University, China, People's Republic of; 2: The University of Tokyo, Japan; 3: Harbin Engineering University, China, People's Republic of

Evaluating the Impact of Physical Models on Physics-Informed Machine Learning for Critical Heat Flux

Abdullah Al Mahmud^1,2, Asao Kanta¹, Koji Morita¹, Wei Liu¹

1: Kyushu University, Japan; 2: Bangladesh Atomic Energy Commission, Bangladesh

Dean Number Influence on CHF Occurrence in U-bend Tubes of Steam Generators

Tanuj Chauhan, Harish Pothukuchi

Indian Institute of Technology, India

Development of an Upgraded Critical Heat Flux Look-Up Table

Jun Yang, Aleksandar Vasić, Lanqin Yuan, Dé Groeneveld

Canadian Nuclear Laboratories, Canada

Numerical Study on Critical Heat Flux and the Influence of Eccentricity in Helical Cruciform Fuel under Non-Uniform Heating

Yuanjie Zhang¹, Bin Han¹, Xiaoliang Zhu¹, Baowen 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

Experimental Investigation on the Uncertainty of Three-sensor Conductivity Probe for Droplet Measurement

Kang Li^1,2, Wenxiong Zhou^1,2, Xubin Tan^1,2, Runzhi Huang^1,2, Mingyue He^1,2, Kexin Qu^1,2, Qiang Lian^1,2, Wan Sun^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

Preliminary Investigation of a Fiber Optic Technique for Flow Rate Measurement in Horizontal Air-Water Stratified Flow

Hanna Kim¹, Hyewon Lim², Hyungmo Kim¹

1: Mechanical Engineering, Gyeongsang National University, Korea, Republic of; 2: Graduate School of Mechanical and Aerospace Engineering, Gyeongsang National University, Korea, Republic of

Infrared Thermometry Investigation of Flow Boiling Transient Critical Heat Flux under Exponentially Escalating Heat Input on Surfaces with Different Finish and Wettability

Yang Liu, Sebastian Rodriguez Rosero, Florian Chavagnat, Matteo Bucci

Massachusetts Institute of Technology, United States of America

Borescopic Molecular Tagging Velocimetry in PWR Surrogate Bundle

Roberto Capanna¹, Charles Fort¹, Guillaume Ricciardi², Philippe Bardet¹

1: The George Washington University, United States of America; 2: CEA, DES, IRESNE, Nuclear Technology Departement, France

CFD-Grade Measurements of Flow-Debris Interaction and PWR Filter Clogging Behavior using MRI Scanner

Vincent Klempau¹, Markus Rehm², Kristine John¹, Elmar Rieder², Sven Grundmann¹, Martin Bruschewski¹

1: University of Rostock, Germany; 2: Framatome GmbH, Germany

Challenges and Non-Conservatism in Indirect Simulation Heaters for Thermal-Hydraulic Experiments

Bao-Wen Yang¹, Jeremy Silber¹, Eric Yang², Stephanie H. Yang¹, Ray Smith², Yash Yadav², Bin Han³

1: Delta Energy Group New York (DEGNY/GDES), United States of America; 2: CARP Associates USA, LLC, United States of America; 3: Southeast University, China, People's Republic of

Solidification-Melting Behaviors and Mushy Characteristics of Molten Salt in Filling Horizontal Cold Pipe

HongXiang Yu^1,2, Chong Zhou^1,2, Xiaofeng Yuan¹, Yushuang Chen^1,2, Shanwu Wang¹, Hui Deng^1,2

1: Shanghai Institute of Applied Physics, China, People's Republic of; 2: University of Chinese Academy of Sciences, China, People's Republic of

Optimizing Core Stability and Flow in Passive Molten Salt Fast Reactors Using GeN-Foam

Juhyeong Lee¹, Do won Lee¹, Yonghee Kim², Sangtae Kim¹, Sung Joong Kim¹

1: Hanyang University, Korea, Republic of; 2: Korea Advanced Institute of Science and Technology, Korea, Republic of

Thermal-hydraulic Research Progress in Fluoride Salt Cooled Nuclear Reactor Applied in Commercial Ship

Limin Liu, Jiaming Zhang, Hanyang Gu

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

A Coupled OpenFOAM-Modelica Modelling Framework for Analysing MSR Safety-Related Transients

Sophie Deanesi, Stefano Lorenzi

Politecnico di Milano, Italy

Enhanced Multi-Physics Modelling of the MSRE Using GeN-Foam

Shervineh Amirkhosravi¹, Frik van Niekerk¹, Marina du Toit¹, Alessandro Scolaro², Andreas Pautz²

1: North-West University, South Africa; 2: EPFL, Switzerland

Implementation of Molten Salt – Concrete Interactions into a System Thermal-Hydraulic Code SPECTRA

Marek Stempniewicz, Ferry Roelofs, Eric de Geus

NRG, Netherlands, The

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

Numerical Prediction of Two-Phase Flow in Tube Bundles with the 3D Mixture Multiphase Model

Hamza Ahfidi, Nicolas Goreaud, Ferdinand Louis

Framatome SAS, France

OpenSTREAM: An Open-Source Platform for Two-Phase Flow Modeling and Simulation

Jean-Marie Le Corre¹, Jason Chan², Elisabeth Walter³, Evan T. Hurlburt⁴, Roman W. Morse⁵

1: Westinghouse Electric Sweden AB, Sweden; 2: University of Wisconsin-Madison, United States of America; 3: Royal Institute of Technology, Sweden; 4: Naval Nuclear Laboratory, United States of America; 5: Massachusetts Institute of Technology, United States of America

Development of Validation Technology for Detailed Two-Phase Flow Simulation Codes for Innovative Reactor Design

Hiroyuki Yoshida, Naoki Horiguchi, Ayako Ono

Japan Atomie Energy Agency, Japan

Study of Two-Phase Void Fraction in a Rectangular Channel Using Capacitance Sensor

Peterson Ngoc Nguyen, Kazuya Tajiri

Michigan Technological University, United States of America

CFD Analysis of Diffuser Configurations for Enhanced Oxygen Transfer and Flow Mixing in Two-Phase Reactor Systems

Sergio Chiva, Jaume Luis, Raul Martinez-cuenca, Oscar Prades, Guillem Monros, Salvador Torro

Universitat Jaume I, Spain

A Closure Model for Local Vapor Bubble Nucleation and Waiting Time

Aaron Lentner, Sara Youssoufi, Elias Balaras

George Washington University, United States of America

R&D Acitivities of the GIF Safety and Operation Project of Sodium-cooled Fast Reactor Systems

Jewhan Lee¹, Acacia Brunett², Fabienne Delage³, Hidemasa Yamano⁴, Jin Wang⁵, Pierre Sciora^3,6

1: KAERI, Korea, Republic of; 2: ANL, United States of America; 3: CEA, France; 4: JAEA, Japan; 5: CIAE, China, People's Republic of; 6: EURATOM, Europe

Validation of Subchannel Code CTF for Sodium Fast Reactor Modelling

Jonas Vanderwaeren¹, Lucia Rueda Villegas¹, Christophe Schneidesch¹, Elias-Yammir Garcia-Cervantes²

1: TRACTEBEL, Belgium; 2: CEA, France

Analysis of Cliff Effects and Thermal Hydraulic Instabilities in the PLANDTL-2 Sodium Experiment Transient Tests

Sébastien Renaudière de Vaux¹, Simon Li¹, Toshiki Ezure², Masaaki Tanaka²

1: CEA, France; 2: JAEA, Japan

Core Deformation Reactivity with Neutronics-Thermal Hydraulics-Structural Mechanics Coupled Analysis for FFTF LOFWOS Test #13

Norihiro Doda¹, Shinya Kato¹, Tomoyuki Uwaba¹, Yoshiaki Nakamine², Kennichi Igawa³, Masaki Iida³, Masaaki Tanaka¹

1: Japan Atomic Energy Agency, Japan; 2: NDD Corporation, Japan; 3: NESI Inc., Japan

Visualization of Sudden Gas Release Replicating Fuel Pin Failure in LMFR Geometry

Nicholas Gallagher¹, Guillaume Mignot¹, Wade Marcum¹, Aydin Karahan², Samuel Miller³

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

Development of Physical Models to Simulate Disrupted Core in Metal-fuel Sodium-cooled Fast Reactors

Hirotaka Tagami, Yasushi Okano, Hidemasa Yamano

Japan Atomic Energy Agency, Japan

Fuel Coolant Interaction/Ex- vessel Steam Explosion: An Overview of Experimental Tests Performed at CEA with Prototypic Corium in the Frame of ICE Program

Pascal Piluso¹, Jules Delacroix¹, Andrea Quaini¹, Emmanuelle Bracks¹, René Guinebretière², Denis Menut³, Renaud Meignen⁴, Nicolas Rimbert⁵

1: CEA, France; 2: University of Limoges, France; 3: Synchrotron SOLEIL, France; 4: ASNR, France; 5: University of Lorraine, France

Impact of Advanced Technology Fuel Cladding Materials on the Progression of Severe Accidents in a Generic Natural-Circulation iPWR

Mauricio Exequiel Cazado¹, Zaira Itzel Jimenez Balbuena¹, Laure Carénini², Fabrizio Gabrielli¹, Victor Hugo Sánchez-Espinoza¹

1: Karlsruhe Institute of Technology (KIT), Germany; 2: French Authority for Nuclear Safety and Radiation Protection (ASNR), France

Mitigation of Radioactive Release during Underwater Laser-cutting of Corium after a Severe Accident: An Analytical Study

Luis Herranz¹, Rafael Iglesias¹, Emmanuel Porcheron²

1: CIEMAT, Spain; 2: ASNR, France

Investigation of the Effects of Surface Inclination on the Ablation of a Solid by the Impact of Hot Liquid Jet: Implications for Sodium-cooled Fast Reactor Safety

Asma Djermoune^1,2, Alexandre Lecoanet¹, Nathalie Seiler¹, Nicolas Rimbert², Michel Gradeck²

1: CEA, France; 2: Université de Lorraine, France

Development and Qualification of Advanced Core Catcher for SFR

Prabhat Kumar Shukla^1,2, E. Hemanth Rao¹, Anish Kumar^1,2, Sagar Nagpure¹, G. Muthu Kumar¹, Avinash Kumar Acharya¹, Sanjay Kumar Das¹, B.K Sreedhar¹

1: Indira Gandhi Centre for Atomic Research, India; 2: Homi Bhabha National Institute, India

Quantification of the Influence of Air Entrainment on Triggering of Single Molten Droplet

Cheng Peng¹, Kaiwen Shi¹, Dong Li¹, Dmitry Grishchenko², Pavel Kudinov²

1: Shanghai University of Electric Power, China, People's Republic of; 2: Royal Institute of Technology, Sweden

Flashing Induced Instabilities in a Forced Circulation Loop under Low Pressure and Low Power Conditions

Karim Mohammed Abdalwaged Yassin, Carlos Alberto Dorao, Maria Fernandino

Norwegian university of Science and Technology (NTNU), Norway

Theoretical Analysis of Two-Phase Flow Instability in Once-through Steam Generator Using Drift Flow Model

Yang Su, Xiaowei Li, Hongyi Fan, Xinxin Wu, Xiaowei Luo

Tsinghua University, China, People's Republic of

Theoretical Model of Water Jet Instability and Phase Change Steam Expansion during SGTR in Lead-cooled Fast Reactor

Ruiqi Bao, Li Liu, Hanyang Gu

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

The Study of Transient Heat Transfer Mechanisms and Two-Phase Flow during Post Flow Instability Dryout Accident

Yosef Aharon¹, Almog Chakmon^1,2, Gennady Ziskind²

1: NRCN, Israel; 2: Ben Gurion University, Israel

Defining Axial Void Fraction and Flashing Details from Pressure Profiles in Two-phase Critical Flow Discharges Made in CRAFTY Facility

Lauri Pyy, Giteshkumar Patel, Silja Oksanen, Juhani Hyvärinen

LUT University, Finland

A Study on the Phase State Measurement Method for Gas-liquid Two-phase Flow in a Tube during SCO₂ Loss-of-pressure Flash Vaporization

Junjie Yuan, Li Liu, Hanyang Gu

Shanghai Jiaotong University, China, People's Republic of

Sodium Heat Pipe under Pulsed Power near Operation Limits

Pablo Diaz Gomez Maqueo, Reilly MacCoy, Rory McGrath, Chukwudi Azih

Canadian Nuclear Laboratories, Canada

Development of a High-Temperature Heat Pipe Simulation Module for the Thermal Hydraulic System Code AC²/ATHLET: Laminar Vapor Flow Modeling

Daniel Eckert¹, Philipp Schöffel¹, Fabian Weyermann¹, Andreas Schaffrath¹, Jörg Starflinger²

1: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Germany; 2: Institute of Nuclear Technology and Energy Systems (IKE), University of Stuttgart, Germany

Parametric Optimization of Heat Pipe Design for Enhanced Thermal Performance Using Genetic Algorithm

Linjie Xu, Panhe Ge, Gu Hu

China Institute of Atomic Energy, China, People's Republic of

A Flexible Coupling Approach for Heat Pipe Microreactor Analysis

Mario A. Ponce Tovar^1,2, Cynthia Camila Güttlein^1,3, Ivor David Clifford¹, Alexey Cherezov¹, Hakim Ferroukhi¹

1: Paul Scherrer Institut, Switzerland; 2: Eidgenössische Technische Hochschule Zürich (ETH Zurich), Switzerland; 3: École Polytechnique Fédérale de Lausanne (EPFL), Switzerland

Neutronics and Planned Coupled Neutronics-Thermal hydraulics Simulations of a Heat Pipe Cooled MMR Core

Norman Dünne¹, Jeremy Bousquet¹, Andreas Wielenberg¹, Andreas Schaffrath¹, Jörg Starflinger²

1: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Germany; 2: University of Stuttgart, Germany

Design Study to Develop an Experimental Facility for a Microreactor Process Heat Application: Cement Calcination

Scout Bucks¹, Ama Dahanayake¹, Erik Hisahara¹, Martin Nieto-Perez¹, Saya Lee¹, Sola Talabi², Sathish Krishnamoorthy³

1: The Pennsylvania State Univeristy, United States of America; 2: Pittsburgh Technical, United States of America; 3: Nazareth Cement Plant, Heidelberg Materials US Inc., United States of America

Numerical Simulation and Sensitivity Analysis of Condensation-Induced Water Hammer in Steam Generator Blowdown System

Rui Zhang, Jiaming Zhao, Wentao Guo, Shengfei Wang, Li Peng, Jiannan Shen, Danfeng Zhao, Xiaoxia Chen

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

A Priori Modelling of the Temperature Variance Dissipationtransport Equation

Corina Sanz Souhait¹, Jean-François Wald¹, Sofiane Benhamadouche¹, Rémi Manceau²

1: EDF R&D, France; 2: CNRS, Universite de Pau et des Pays de l’Adour, INRIA, France

Analysis of Surface Temperature Distribution on Micro-Pillar Structures Using Fiber Optic Sensors under Pool Boiling Conditions

Minkyu Park¹, Hyeon Taek Nam¹, Seongro Lee², Euijae Kim³, Jae-Ho Lee³, Donghwi Lee^1,4

1: Graduate School of Mechanical-Aerospace-Electric Convergence Engineering, Jeonbuk National University, Korea, Republic of; 2: Department of Mechanical Engineering, Jeonbuk National University, Korea, Republic of; 3: KEPCO Nuclear Fuel Co.,Ltd., Korea, Republic of; 4: Department of Mechanical System Engineering, Jeonbuk National University, Korea, Republic of

Bubble Growth Characteristics on a Moving Wall

Sohyeun Kang, Jaejun Kim, Minseop Lee, Daegyoum Kim

KAIST, Korea, Republic of

Numerical Study of the Dynamic Response of RCS Boron Concentration during Dilution Operation

Xiangyu Chi, Jiazhuo Wang, Pei Yu, Jiazhi Liu, Jiaming Zhao, Yaru Li

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

Reverse Forces on Check Valve Obturators Produced by Geometric Parameter Combinations – a DOE Study

Tsolag Apelian, Genesis Ponce

Curtiss Wright, United States of America

The Cross Section Generation Method based on Deep Learning for Neutronics and Thermal Hydraulics Coupling

Lu Cao¹, Lipeng Wang¹, Duoyu Jiang¹, Tianliang Hu¹, Huaqi Li¹, Hailong Huang¹, Zhouyu Liu², Liangzhi Cao²

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

A Study on PAFS Heat Transfer Performance Prediction Using Korean Thermal-hydraulic System Codes

Kyungho Nam, Hyunjoon Jeong

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

Vertical Mini-Channel Multiphase Flow Model Simulation and Experiment

Witiwat Jiragoontansiri¹, Taeseok Kim², Florian Chavagnat¹, Arunkumar Seshadri¹, Koroush Shirvan¹

1: Massachusetts Institute of Technology, United States of America; 2: Jeju National University, Korea, Republic of

Development and preliminary verification of system analysis models and code for Small Lead-based cooled fast reactor Based on modelica

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

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

Analysis of MSGTR Accident with PAFS at the ATLAS Experimental Facility Using the MARS-KS and SPACE Code

Hyunjoon Jeong, Kyungho Nam

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

Review and Analysis of the Passive Heat Removal System Experimental Facility for Validation of the Robustness Assessment Methodology

Jehee Lee¹, Seong-Su Jeon¹, Nhan Hien Hoang¹, Ju-Yeop Park²

1: FNC Technology co., LTD., Korea, Republic of; 2: Korea Institute of Nuclear Safety, Korea, Republic of

Numerical Simulation of Single Boiling Bubble Generation with Implicit Microlayer Model and Explicit Transition Region under Conjugated Heat Transfer

Zuzhi Li, Jonghwi Choi, Hyungdae Kim

Kyung Hee University, Korea, Republic of

The Numerical Study of the Thermal-hydraulic Characteristics of the In-containment Storage Rack

Yaru Li, Yanyu Sun, Shuliang Huang

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

Investigation of Flow Mixing Characteristics in APR1000 Reactor Flow Model Based on Turbulence Models

Uiju Jeong

KHNP Central Research Institute, Korea, Republic of

Comparative Study on the Effect of Heat Transfer and Steam Fraction on Safety Valve through CFD Analysis

Ji Su Kim, Hui Un Ha, Kwan Geun Lee, Jae Yoon Jeong, Hyun Suk Nam, Kyoung Uk Lee, Seok Jong Yoon, Uiju Jeong, Dae Hun Kim

KHNP, Korea, Republic of

Impact of Fin Height and Side Slot Design on Cooling Performance of Hypervapotron Technique

Iljin Kim¹, Gubin Lee², In-Yeop Kang², Hyungdae Kim²

1: Korea Atomic Energy Research Institute, Korea, Republic of; 2: Kyung Hee University, Korea, Republic of

Numerical Investigation on Phase Separation Characteristics of Bubbly Flow in Horizontal T-Junction

Wang Fupeng^1,2, Zhu Di^1,2, Sun Rulei^1,2, Wang Shixuan^1,2, Tan Sichao^1,2, Tian Ruifeng^1,2

1: State Key Laboratory of Marine Thermal Energy and Power, China, People's Republic of; 2: Harbin Engineering University, China, People's Republic of

Application of SPACE/CUPID/MASTER Coupled Code: Main Steam Line Break Accident

Chiwoong Choi, Ik-kyu Park, Jaeseok Heo, Kwiseok Ha

KAERI, Korea, Republic of

Design and Development of Separate and Integral Effects Test Facilities for Licensing and Deployment of Advanced Water-Cooled Small Modular Reactors

Kenneth Lee Fossum¹, Grant Hendrickson¹, Palash Kumar Bhowmik¹, Justin Thales Johnson¹, Piyush Sabharwall¹, Clay Lietwiler²

1: Idaho National Laboratory, United States of America; 2: Holtec International, United States of America

Lucky Loop - A Thermal Hydraulic Experiment to Create Data for Validation of Modern Codes

Tilmann Schlitt¹, Christian Reiter¹, Mackenzie Wilson²

1: FRM II / TUM, Germany; 2: McMaster University, Canada

Experimental Study on an Intermediate Break Loss of Coolant Accident (IBLOCA) under the OPR1000 Operation Condition

Yusun Park, Byoung Uhn Bae, Jae Bong Lee, Seok Cho, Jong Rok Kim, Jun Yeong Jung, Kyoung Ho Kang, Seung Wook Lee

KAERI, Korea, Republic of

Heating and Cooling System Design of the Separate Effect Test Facility for Flow-Accelerated Corrosion and Erosion (SEFACE) Studies in Liquid Lead

Haipeng Li, Max Persson, Ignas Mickus, Dmitry Grishchenko, Pavel Kudinov

KTH Royal Institute of Technology, Sweden

Integral Effect Test and Code Analysis on the Cooling Performance of the PAFS during a SLB Accident

Jongrok Kim, Seok Cho, Byoung-Uhn Bae, Jae Bong Lee, Yusun Park, Kyoung-Ho Kang

Korea Atomic Energy Research Institute, Korea, Republic of

Experimental Study of Oxidized Cladding Effects on Fuel Cladding Behavior During Reflood Phase Using Thermo-Mechanical and Thermal-Hydraulic Coupled Experiment

Jongrok Kim, Kihwan Kim, Jae-Bong Lee, Jong-Kuk Park, Young-Jung Youn, Hae Seob Choi, Byung-jun Lim, Sang-Ki Moon, Seok Kim

Korea Atomic Energy Research Institute, Korea, Republic of

Study on Effects of Pressurizer Spray and Heater on Pressurizer Level Control System Malfunction Event of APR1000 Using SPACE Code

Ye Eun An, Sang Jin Lee, Jisu Kim, Se Young Ro, Min Shin Jung, Ung Soo Kim

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

Numerical Analysis of Hydrogen Recombination under Natural Convection Condition between Two Vertical Flat Catalytic Plates

Dae-Hyun Kim¹, Stephan Kelm², Ernst-Arndt Reinecke², Youngsu Na³, Dong-Wook Oh¹

1: Chosun University, Korea, Republic of; 2: Forschungszentrum Juelich GmbH, Institute of Energy Technologies (IET-4), Germany; 3: Korea Atomic Energy Research Institute, Korea, Republic of

CFD Modeling for Ring Type Fuel Spacers in Sleeveless HTGRs Core Design

Yusuke Yamazaki, Yosuke Nishimura, Koji Okamoto

The University of Tokyo, Japan

Numerical Analysis of Low-leakage Molten Salt Spreading behavior and Solidification in NaCl-KCl-UCl3

Ye Hwan Chun¹, Jae Hyung Park¹, Sang Mo An², Sung-Il Kim², JinHo Song¹, Jae Min Han¹, Sung Joong Kim^1,3

1: Hanyang University, Korea, Republic of; 2: Korea Atomic Energy Research Inst., Korea, Republic of; 3: Institute of Nano Science and Technology, Hanyang University, Korea, Republic of

Research on Data-Physics Hybrid-driven Core Temperature Reconstruction of Heat Pipe Reactors

Pengzhi Zhang, Zeguang Li, Huifu Wang, Henglong Lin, Zilin Su

Tsinghua University, Department of Engineering Physics, Beijing 10003, China, People's Republic of

Development Status of Standard Design for Passive Safety System in i-SMR

Chang Hun Lee, Seok Ho Lee, Hui Un Ha, Seok Jong Yoon, Ji Su Kim, Dong Won Jeong, Dae Hun Kim

Central Research Institute, Korea Hydro Nuclear Power, Korea, Republic of

Improving Initialization of ASTEC's Thermal-Hydraulic Solver with Machine Learning-Based Methods for Enhanced Convergence in Severe Accident Simulations

Marcello Savini^1,2, Raphael Monod³, Simone Gianfelici², Bastien Poubeau³, Fulvio Mascari²

1: Alma Mater Studiorum - Università di Bologna, Italy; 2: ENEA, Italy; 3: IRSN, France

Comparative Evaluation of Advanced PINN Techniques for Efficient Fluid Dynamics Modeling

Hanseul Kang, Kin Wing Wong

KTH Royal Institute of Technology, Sweden

Enhancing Critical Heat Flux Prediction in Fuel Rod Bundles with Interpretability Transformer Architectures

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

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

Modeling of Sodium Heat Pipes for Microreactors Passive Cooling

Riccardo Luigi Sansone^1,2, JunSoo Yoo³, Victor Petrov^1,2, Annalisa Manera^1,2

1: ETH Zurich, Switzerland; 2: Paul Scherrer Institut, Switzerland; 3: Idaho National Laboratory, United States of America

Experimental Evaluation of the Buckling of a Single Cylindrical Plate

Alexander Hans Westerberg¹, Wade Marcum¹, Aaron Weiss¹, Guanyi Wang², Cezary Bojanowski², Andrew Hebden²

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

Flow Pattern Analysis of Printed Circuit Heat Exchangers with Zigzag Channels Using PIV Visualization and CFD

Yoomyeong Lee¹, Jaejun Hyun¹, Seongmin Lee², Hong Beom Park², Kyoungwoo Seo², Donghwi Lee^1,3

1: Graduate School of Mechanical-Aerospace-Electric Convergence Engineering, Jeonbuk National University, Korea, Republic of; 2: Korea Atomic Energy Research Institute, Korea, Republic of; 3: Department of Mechanical System Engineering, Jeonbuk National University, Korea, Republic of

Dynamic Simulation and Performance Analysis of a Nuclear-solar Energy System with Thermal Energy Storage

Xinyu Li^1,2, Yan Liu^1,2, Xiang Chai^1,2, Xiaojing Liu^1,2

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

FONESYS Benchmark of Core Interfacial Friction Models in System Codes

Dominique Bestion¹, Francesco D'Auria², Qingling Cai², Jaeseok Heo³, Philipp Schoeffel⁴, Hongsheng Yuan⁵, Amandine Capogna⁶, Sofia Carnevali⁶

1: Consultant, France; 2: UNIPI, Italy; 3: KAERI, Korea, Republic of; 4: GRS, Germany; 5: CNPRI, China, People's Republic of; 6: CEA, France

Giving a Major Role to Bifurcating Events in Pirt and Scaling Analysis for Light Water Reactor Thermalhydraulics

Dominique Bestion¹, Sofia Carnevali², Francesco D'Auria³

1: Consultant, France; 2: CEA, France; 3: UNIPI, Italy

Current and Planned Activities of the FONESYS Network of System Code Developers in Collaboration with the SILENCE Network of Experimentalists

Dominique Bestion¹, Francesco D'Auria², Klaus Umminger³, Jaeseok Heo⁴, Philipp Schoeffel⁵, Sofia Carnevali⁶, Jean Luc Vacher⁷, Pierre Gaillard⁸, Huie Sha⁹, Hongsheng Yuan¹⁰, Geoffrey Waddington¹¹, Jean Marie LeCorre¹²

1: Consultant, France; 2: University of Pisa, Italy; 3: Consultant, Germany; 4: KAERI, Korea, Republic of; 5: GRS, Germany; 6: CEA, France; 7: EDF, France; 8: Framatome, France; 9: SPICRI, China, People's Republic of; 10: CNPRI, China, People's Republic of; 11: CNL, Canada; 12: Westinghouse, Sweden

FONESYS Benchmark of Core-3D-Mixing in System Codes

Dominique Bestion¹, Jaeseok Heo², Huie Sha³, Philipp Schoeffel⁴, Adrien Rodriguez⁵, Hongsheng Yuan⁶, Vincenzo Zingales⁷, Raphael Prea⁵, Lin Wang³, Francesco D'Auria⁷

1: Consultant, Grance; 2: KAERI, Korea, Republic of; 3: SPICRI, China, People's Republic of; 4: GRS, Gernany; 5: CEA, France; 6: CNPRI, China, People's Republic of; 7: UNIPI, Italy

Fonesys Benchmark of Singular Pressure Losses in System Codes

Dominique Bestion¹, Philipp Schoeffel², Nicolas Piette³, Jean Luc Vacher⁴, Jae Jun Jeong⁵, Qingling Cai⁶, Vincenzo Zingales⁶, Jaeseok Heo⁷, Francesco D'Auria⁶

1: Consultant, France; 2: GRS, Germany; 3: CEA, France; 4: EDF, France; 5: Pusan National University, Korea, Republic of; 6: UNIPI-GRNSPG, Italy; 7: KAERI, Korea, Republic of

Using System Code for Scaling Analysis: A New Integrated Tool in the CATHARE Code Applied to a SB-LOCA Transient

Valentin Cottarel¹, Sofia Carnevali¹, Marco Montanini¹, Dominique Bestion²

1: French Alternative Energies and Atomic Energy Commission (CEA), France; 2: Scientific Consultant, France

The CATHARE Code Validation on CRIEPI Core Void Fraction Data

Amandine Capogna¹, Jean-Luc Vacher², Thibaut Lopez¹, Dominique Bestion³

1: CEA, France; 2: EDF, France; 3: Consultant, France

Impact of Electroplated Porous Copper Layer Thickness on Critical Heat Flux in Saturated Pool Boiling

Yuya Hayashida¹, Yutaro Umehara^1,2, Atsuro Etoh¹, Shoji Mori^1,2

1: Kyushu University, Japan; 2: International Institute for Carbon-Neutral Energy Research, Japan

Numerical Study on Critical Heat Flux and the Influence of Eccentricity in Annular Pipe under Non-Uniform Heating

Bin Han¹, Yuanjie Zhang¹, Xiaoliang Zhu¹, Baowen 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

Preliminary Pool Boiling Experimental Study on the Impact of Deposition on the Critical Heat Flux of Horizontally Placed Tubes

Xiaowen Wang¹, Maolong Liu², Tenlong Cong¹, Yao Xiao¹, Hanyang Gu¹

1: Shanghai Jiaotong University, China, People's Republic of; 2: Fudan University, China, People's Republic of

Exploring a Link between Void Fraction Profile and Critical Heat Flux in Subcooled Flow Nucleate Boiling

Sebastian Rodriguez Rosero, Alexander Edwards, Matthew Hughes, Florian Chavagnat, Matteo Bucci

Massachusetts Institute of Technology, United States of America

CHF Experiments for Pool Boiling from Inclined Heater Surfaces with Stainless Steel 304 Plates

Young Hee Kim¹, Dong Hun Shin¹, Min Suk Lee², Dong Hoon Kam³, Yong Hoon Jeong¹

1: Korea Advanced Institute of Science and Technology, Korea, Republic of; 2: Texas A&M University, United States of America; 3: Argonne National Laboratory, United States of America

Assessment of Heat Flux Partitioning Approaches for the Prediction of Boiling and the Critical Heat Flux

Omid Bidar, Marco Colombo

University of Sheffield, United Kingdom

Experimental Design and Algorithm Validation of Reactor Chemical and Volume Control System Upper Charging Line Valve Leakage Faults

Jiajun Li, Hengyuan Wei, Dongyang Li, Sichao Tan, Ruifeng Tian

Harbin Engineering University, China, People's Republic of

Basic Design of PCCS Heat Exchanger of Integral Effect Test Facility for i-SMR Validation Test

Jin-Hwa Yang, Byoung-Uhn Bae, Hwang Bae, Kyoung-Ho Kang

Korea Atomic Energy Research Institute, Korea, Republic of

The Improvement and Preliminary Validation of Relap5 Code for Integrated Natural Circulation SMR

Zhanfei QI, Xiumei Zhao, Zijiang Yang, Shengzhe Li, Yue Wu

SNERDI, China, People's Republic of

Integrated Thermo-Mechanics and Thermal-Hydraulics of Zircaloy-4 Cladding Behavior under LBLOCA Conditions Using the ICARUS Facility

Rahmatika Intan Ali^1,2, Jongrok Kim¹, Kihwan Kim¹, Sang-Ki Moon¹, Seok Kim¹, Jae Bong Lee^1,2

1: Korea Atomic Energy Research Institute, Korea, Republic of; 2: KAERI School, University of Science and Technology, Korea, Republic of

A Scaled-Down Approach for Designing a Compact Hydraulic Apparatus for Nuclear Experimental Liquid fuel reactors (CHANEL)

Hyeon Ji Kim, In Cheol Bang

Ulsan National Institute of Science and Technology, Korea, Republic of

Experimental Investigations of Natural Convection in a Differentially-Heated Cavity

Noah LeFrancois, Robert Bowden

Canadian Nuclear Laboratories, Canada

Undercooling Conditions in SFR Low Void Core Designs

Sara Perez-Martin

Karlsruhe Institute of Technology, Germany

Development of a Novel OpenFOAM Solver for Tritium Permeation Modeling in Molten Fluoride Salt Systems

Wooseong Park^1,2, Guanyu Su¹, Yong Hoon Jeong²

1: UC Berkeley, United States of America; 2: KAIST, Korea, Republic of

CFD Simulation of Geyser Boiling and Startup Instabilities of a Two-Phase Closed Water Thermosyphon using OpenFOAM

Almanzo Arjuna, In Cheol Bang

Ulsan National Institute of Science and Technology, Korea, Republic of

CFD Modeling of Thermal-hydraulic Behavior in SCWR: Analysis of Trans-critical Transients

Shubham Rajesh Vaidya, Goutam Dutta, Harish Pothukuchi

Indian Institute of Technology Jammu, India

Modelling of Selected SAFARI-1 Research Reactor Transients using RELAP/SCDAPSIM/MOD3.4

Samukeliswe Nozipho Khoza, Rian Hendrik Prinsloo

South African Nuclear Energy Corporation SOC Limited (Necsa), South Africa

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

A Broyden-Type Newton Method for Two-Fluid Model’s Fully-Implicit Solution Scheme

Hao Zhang, Meng Zhao, Yanhua Yang

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

Simulation of BWR-Relevant Swirling Annular Flow Using the Conservative Level Set Method

Taylor Edwin Grubbs, Nicholas Mecham, Igor Bolotnov

North Carolina State University, United States of America

Assessment of Drag Force Formulations for Two-field All-flow-regime Models

Peio Berrécochéa Machingoréna^1,2, Dominique Legendre², Daniele Vivaldi¹

1: ASNR (Autorité de Sûreté Nucléaire et de Radioprotection), France; 2: IMFT (Institut de Mécanique des Fluides de Toulouse), France

Investigation on Bubble Residence Time of Swarm Flow in Pool Scrubbing Process

Fanli Kong, Xing Chen, Xu Cheng

Karlsruhe Institute of Technology, Germany

Numerical Investigation on the Structural Integrity of an Ex-vessel Core Catcher under Guillotine-type RPV Failure Using Smoothed Particle Hydrodynamics

Jin-Woo Kim¹, Hee Sang Yoo^1,2, Tae Hoon Lee¹, Yelyn Ahn¹, Min Chan Kwon¹, Eung Soo Kim¹

1: Seoul National University, Korea, Republic of; 2: French Alternative Energies and Atomic Energy Commission (CEA), France

Analysis of Corium Stratification Effect on Molten Core-Concrete Interaction observed in the MOCKA 3.1 Test

Jun Sung Choi, Tae Hwan Kim, Jin-Woo Kim, Min Jae Kim, Eung Soo Kim, Hyun Sun Park

Seoul National University, Korea, Republic of

Development of the New Jet Breakup Length Correlation Considering the Effect of Vapor Generation Intensity on the Melt Jet Fragmentation

WooHyun Jung¹, Hyun Sun Park²

1: University of Wisconsin, United States of America; 2: Seoul National University, Korea, Republic of

Empirical Feedback on Layout Design Optimization of Reactor Cavity Water Injection Cooling System

Li Peng, Zhe Jiang

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

URANS Simulation of CANDU Debris Bed Transient Melting in a Severe Accident

Yeshan Sun, Stephen Tullis, Marilyn Lightstone

McMaster University, Canada

Hydrogen Concentration Distribution in the Reactor Building of Fukushima Daiichi NPP Unit-3

Masanori Naitoh, Fumitomo Onishi, Atsuo Takahashi, Toshoharu Mitsuhashi, Achihoro Hamano, Hirotaka Hadachi

Advancesoft Corporation, Japan

Test Results of an HTSE Experimental Facility Integrating with a Helium Loop for Low Carbon Hydrogen Production

In-Woo Son, Sin-Yeob Kim, Sung-Deok Hong, Byung-Ha Park, Chan-Soo Kim

Korea Atomic Energy Research Institute, Korea, Republic of

NuScale Integrated Energy System for Petrochemical Plant Emissions Reduction

Jose Reyes, Luis DePavia, John Swanson

NuScale Power, United States of America

Performance Analysis of the Nuclear-powered IECT Propulsion System

Hangqian Liu^1,2, Ao Zhang¹, Xiang Wang*¹

1: Harbin Engineering University, China, People's Republic of; 2: University of Science and Technology of China, China, People's Republic of

Multi-objective Decisions on Integrated Energy Systems Planning and Operation for Industrial Combined Heat and Power Supply

So-Bin Cho^1,2, Junyung Kim¹, Václav Novotný¹, Rami M. Saeed¹, Todd Allen², Xiaodong Sun²

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

A Realistic Metric for Latent Heat Thermal Energy Storage Systems to be Paired with Advanced Reactors

Beau Walsh, Guanyu Su

UC Berkeley, United States of America

S-Allegro Integral Test Facility Thermal Hydraulic Benchmark: Steady State Qualification of Heat Exchanger Models

Boris Kvizda¹, Miriama Lacková¹, Slavomír Bebjak¹, Bendegúz Kopp², Tamás Varju², Zalán Csekei³, Zsombor Bali³, Gusztáv Mayer³, Tomas Melichar⁴, Daniel Kříž⁴, Jedrzej Walkowiak⁵, Michal Spirzewski⁵

1: VUJE, a.s., Slovak Republic; 2: Budapest University of Technology and Economics, Hungary; 3: HUN-REN Centre for Energy Research, Hungary; 4: Centrum Vyzkumu Rez, s.r.o., Czech Republic; 5: Narodowe Centrum Badan Jadrowych, Poland

Thermal-hydraulics Analysis of ALLEGRO Gas-cooled Fast Reactor with Improved Refractory Core

Gusztáv Mayer¹, Zsombor Bali^1,2, Attila Guba¹

1: HUN-REN Centre for Energy Research, Institue for Atomic Energy Research, Hungary; 2: Budapest University of Technology and Economics, Hungary

Preliminary Validation of Radiation Model Comparison for Radiative Heat Transfer Analysis in MHTGR RCCS

Je-Hyeong Park, Jae-Ho Jeong

Chung-Ang University, Korea, Republic of

Research on the Development of Simulation Models for Stirling Integrated Gas-Cooled Reactors

Tao Zhou¹, Shuaijie Lu², Genglei Xia¹, Yuandong Zhang¹, Guanghui Jiao¹

1: Harbin Engineering University, China, People's Republic of; 2: Wuhan Second Ship Design and Research Institute, China, People's Republic of

Modelling of the S-ALLEGRO Secondary Heat Exchanger in ATHLET

Zsombor Bali^1,2, Attila Guba¹, Gusztáv Mayer¹

1: HUN-REN Centre for Energy Research, Institute for Atomic Energy Research, Hungary; 2: Budapest University of Technology and Economics, Hungary

Analysis of ATWS Scenarios in HTR-10 Operating at Higher Temperatures

Yanhua Zheng¹, Marek Stempniewicz², Xin Wang¹, Ferry Roelofs², Sun Jun¹

1: INET, Tsinghua University, China, People's Republic of; 2: Nuclear Research Group, Netherlands, The

Experimental Investigation of a Compact Diffusion-Bonded Steam Generator for High-Temperature Gas Reactors

Ji Yong Kim¹, Haekyun Park^1,2, Xiaodong Sun¹

1: University of Michigan, United States of America; 2: Kyungpook National University, Korea, Republic of

Experimental Validation of Identified Dimensionless Pitch Parameter of Additively Manufactured Helically Rifled Tubing for Molten Salt Heat Exchangers

Ryan Patrick McGuire, Lane Carasik, Andrew Hutsell, James Vulcanoff, Sierra Tutwiler

Virginia Commonwealth University, United States of America

Experiments on Thermal Stratification at Horizontally Oriented Pipeline-Elbows

Zhongkang Ma, Mi Zhou, Naiqiang Zhang

North China Electric Power University, China, People's Republic of

Flow Pattern Transition and Separation Performance of Swirl-Vane Separator in Small Modular Reactors

Luyun Mao, Nan Gui

Tsinghua University, China, People's Republic of

Comparison of MARS-KS and SPACE Code for Simulating a Helical Steam Generator Two-phase Instability

Junha Hwang, Seunghwan Oh, Jeong Ik Lee, Doh Hyeon Kim, Semin Joo, Taeyeon Min

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

Experimental and Numerical Investigation on the Resistance Characteristics of the New Throttling Device for Steam Generators

Bo Wang^1,2, Yukang Liu¹, Yin Fang², Xianliang Lei², Yanfang Liu¹, Runze Liu², Xinyang Guo²

1: Dongfang Electric Co.Ltd., China, People's Republic of; 2: Xi’an Jiaotong University, China, People's Republic of

Study of Reactor Core Residual Heat Removal Schematic

Jiaming Zhao, Rui Zhang, Xiaoxia Chen, Nianzhe Zhang, Xiangyu Chi

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

Verification of the Reflectivity of the Boundary Surface Regarding the Development of Water Level Measurement Technology Using Ultrasound

Takahiro Ito¹, Kohei Kikuchi¹, Weichen Zhang¹, Hiroshige Kikura¹, Daisuke Sasa², Takeshi Suzuki²

1: Laboratory for Zero-Carbon Energy, Institute of Science Tokyo, Japan; 2: Tokyo Electric Power Company Holdings, Inc., Japan

BEEs: A Multiphysics Simulation Engine for Advanced Nuclear Fuel Innovation

Guanghui Su

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

On Flow-Induced Vibrations of Single and Multiple Cylinders in Cross-Flow and using Medium-Resolution Methods for Numerical Predictions

Kevin Zwijsen

Nuclear Research and Consultancy Group, Netherlands, The

A Plan for Developing a Korean Virtual Reactor Platform for SMR Demonstration and Operation

Yun-Je Cho