Conference Agenda

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

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

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

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

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

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

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

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

Benchmark of System Thermal-hydraulic Codes for the Dip Cooler Instability Facility Test Section

Danilo Caterino¹, Giuseppe Francesco Nallo¹, Antonino Lombardo², Matteo Rostagno², Valeria Parrinello², Giuseppe Spina², Roberto Bonifetto¹

1: Politecnico di Torino, Italy; 2: newcleo SpA, Italy

Experimental Investigation on Convection Heat Transfer Characteristics of Lead–bismuth Eutectic in Circular Tubes Under Natural Circulation

Jieming Hou¹, Kuang Bo¹, Wei Chen², Shirui Li³

1: Shanghai Jiao Tong University, China, People's Republic of; 2: Wuhan University of Science and Technology, China, People's Republic of; 3: China Institute of Atomic Energy, China, People's Republic of

A Thermal-hydraulic Simulation Model and Control Modeling of a Lead-bismuth Reactor based on Numap Software

Xinheng Zhao, Minjun Peng, Chunyuan Niu

Harbin Engineering University, China, People's Republic of

Simulation of NACIE Benchmark Tests with the SAS4A/SASSYS-1 Code

Anton Moisseytsev

Argonne National Laboratory, United States of America

Numerical Study on the Influence of Rough Surfaces on Flow and Heat Transfer Characteristics in Narrow Rectangular Channels Cooled by Liquid Metals

Xue Zhang¹, Wenpei Feng^1,2, Xiaolong Yang², Shangdong Yang²

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

Computational Study of Helium Bubble Dynamics in Molten Salt via Coupled VOF and Neutronics Multi-physics Framework

Seungsu Han¹, Antonio Cammi², Hyungdae Kim¹

1: Kyung Hee University, Korea, Republic of; 2: Politecnico di Milano, Italy

Simulation of Noble Metal Behavior in Molten Salt from a Molecular Dynamics Perspective

Jiaqi Chen¹, Caleb S. Brooks²

1: University of Shanghai for Science and Technology, China, People's Republic of; 2: University of Illinois Urbana Champaign, United States of America

Coupled Two-Phase Flow and Thermochemistry Modeling in Pronghorn for Molten Salt Tritium Breeding Blanket Analysis

Samuel Walker, Mauricio Tano, Parikshit Bajpai, David Reger

Idaho National Laboratory, United States of America

Benchmark of Eulerian-Lagrangian Methods for Solid Fission Product Tracking inside Molten Salt Reactor

Tommaso Pernatsch¹, Stefano Lorenzi¹, Rocco Moretti²

1: Politecnico di Milano, Italy; 2: NAAREA, Nanterre, France

The Investigation of Noble Metal Mass Transfer Efficiency to Circulating Bubbles with Surfactants

Huihua Yang¹, Qiyun Cheng¹, Ling Zou², Rui Hu², Wei Ji¹

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

Verification, Validation, and Uncertainty Quantification Study of Mo-99 Deposition onto a Cylinder in a Molten Salt Reactor

Logan Stewart Brown, Mark Kimber

Texas A&M University, United States of America

Validation of Scaling Laws for Investigating the Thermal Behavior of Sodium Heat Pipes in Microreactors

Dong Hun Lee, In Cheol Bang

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

Numerical Investigation of Filling Ratio Effects on Heat Pipe Performance: Modified Conduction Model

Faruk Celik, In Cheol Bang

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

Numerical Study of Heat Transfer Characteristics of High Temperature Heat Pipe with Wire Mesh Wick

Qishi Sun¹, ZhiPeng Zhang¹, Kailun Guo¹, Hao Sun², Chenglong Wang¹, Kui Zhang¹, Xiaoqiang He³, Wenxi Tian¹, Suizheng Qiu¹, Guanghui Su¹

1: Xi'an Jiaotong Unversity, China, People's Republic of; 2: China Nuclear Power Technology Research Institute, China, People's Republic of; 3: Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, China, People's Republic of

Development of the Modular Passive-cooling Tower Equipped with a Wickless Heat Pipe for Long-term Cooling of a SMR

Yonadan Choi¹, Min Suk Lee², Yong Hoon Jeong¹

1: KAIST, Korea, Republic of; 2: Texas A&M University, United States of America

Design and Analysis of a 20-kW Free-Piston Stirling Generator for Microreactors

Phat Doan, Minghui Chen

The University of New Mexico, United States of America

Study on Intermittent Boiling Phenomena During Sodium Heat Pipe Startup

Yinghuan Cui, Jianjun Xu, Tianzhou Xie, Zhangrui Yan

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

Grid Independence Test on the Lower Plenum Mixing Test of the High Temperature Test Facility Benchmark

Mohammed Muaaz Mohamed Dilshad Hussain, Ferry Roelofs

NRG PALLAS, Netherlands, The

Multiscale Thermal-hydraulic Analysis of the MARVEL Micro-reactor Using Coupled MOOSE Subchannel (SCM) and SAM

Vasileios Kyriakopoulos, Lise Charlot, Mauricio Tano, Guillaume Giudicelli, Joshua E. Hansel

INL, United States of America

Evaluation of Air and sCO₂ Brayton Cycle for Heat Pipe-Cooled Micro Modular Reactor

Matthias Peiretti, Michael Buck, Jörg Starflinger

University of Stuttgart, Germany